Text

Enclosure 4: Attachment 3 - WCAP-18364-NP, Rev. 1, North Anna Units 1 and 2 Time-Limited Aging Analysis on Reactor Vessel Integrity for Subsequent License Renewal (SLR)
	ML20246G701
Person / Time
Site:	North Anna
Issue date:	03/31/2020
From:	Geer J, Lynch D; Virginia Electric & Power Co (VEPCO), Westinghouse
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML20246G703	List: ML20246G697; ML20246G698; ML20246G700; ML20246G701; ML20246G706; ML20246G707; ML20246G709;
References
	20-115
	Download: ML20246G701 (112)
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Serial No.: 20-115 Docket Nos.: 50-338/339 Enclosure 4 Attachment 3 WCAP-18364-NP, REVISION 1 Virginia Electric and Power Company (Dominion Energy Virginia)

North Anna Power Station Units 1 and 2

I Westinghouse Non-Proprietary Class 3 WCAP-18364-N P March 2020 Revision 1 North Anna Units 1 and 2 Time-Limited Aging Analysis on Reactor Vessel Integrity for Subsequent License Renewal (SLR)

@Westinghouse

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Westinghouse Non-Proprietary Class 3 WCAP-18364-NP Revision 1 North Anna Units 1 and 2 Time-Limited Aging Analysis on Reactor Vessel Integrity for Subsequent License Renewal (SLR)

D. Brett Lynch*

RV/CV Design & Analysis Jared L Geer*

Nuclear Operations & Radiation Analysis March 2020 Reviewers: Benjamin E. Mays*

License Renewal, Radiation Analysis, and Nuclear Operations Benjamin W. Amiri*

Nuclear Operations & Radiation Analysis Approved: Lynn A. Patterson*, Manager RV/CV Design & Analysis Laurent P. Houssay*, Manager Nuclear Operations & Radiation Analysis

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

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

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Westinghouse Non-Proprietary Class 3 11 RECORD OF REVISION Revision O: Original Issue Revision 1: This revision updates the reference to WCAP-18363-NP to Revision 1, corrects the footnote numbering in Table 2-1, and corrects the description in Section 4.0 of those materials with PTS values which exceed that of North Anna Unit 2, Lower Shell Forging

03. Appendix D was added to provide justification for the use of PWROG-17090-NP-A, consistent with the NRC Safety Evaluation.

WCAP-18364-NP March 2020 Revision 1

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

SUMMARY

........................................................................................................................... X 1 TIME-LIMITED AGING ANALYSIS ......................................................................................... 1-1 2 CALCULATED FLUENCE ......................................................................................................... 2-1 3 MATERIALPROPERTYINPUT ................................................................................................. 3-1 4 PRESSURIZED THERMAL SHOCK ......................................................................................... 4-1 5 UPPER-SHELF ENERGY ........................................................................................................... 5-1 6 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES ....................... 6-1 6.1 ADJUSTED REFERENCE TEMPERATURES CALCULATION .................................... 6-1 6.2 P-T LIMIT CURVES APPLICABILITY .......................................................................... 6-27 7 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULES .................................................. 7-1 8 REFERENCES ............................................................................................................................. 8-1 APPENDIX A CREDIBILITY EVALUATION OF THE NORTH ANNA UNITS I AND 2 SURVEILLANCE PROGRAMS .................................................................................... A-1 APPENDIX B EMERGENCY RESPONSE GUIDELINES .................................................................. B-1 APPENDIX C NORTH ANNA UNITS I AND 2 LICENSING BASIS FOR DETERMINING CHEMISTRY FACTOR WHEN SURVEILLANCE DATA IS AVAILABLE ............... C-1 APPENDIX D JUSTIFICATION FOR THE USE OF PWROG-17090-NP-A ...................................... D-1 WCAP-18364-NP March 2020 Revision 1

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Westinghouse Non-Proprietary Class 3 iv LIST OF TABLES Table 1-1 Evaluation of Time-Limited Aging Analyses Per the Criteria of 10 CFR 54.3 ............... 1-2 Table 2-1 Calculated Fast Neutron (E > 1.0 MeV) Fluence at the Surveillance Capsule Center for North Anna Unit l ............................................................................................................ 2-4 Table 2-2 Calculated Fast Neutron (E > 1.0 MeV) Fluence at the Surveillance Capsule Center for North Anna Unit 2 ............................................................................................................ 2-5 Table 2-3 North Anna Unit 1 - Maximum Fast Neutron (E > 1.0 Me V) Fluence Experienced by the Pressure Vessel Materials in the Beltline and Extended Beltline Regions ....................... 2-6 Table 2-4 North Anna Unit 2 - Maximum Fast Neutron (E > 1.0 Me V) Fluence Experienced by the Pressure Vessel Materials in the Beltline and Extended Beltline Regions ....................... 2-8 Table 3-1 Best-Estimate Cu and Ni Weight Percent Values, Initial RTNnT Values, and Initial USE Values for the North Anna Unit 1 RPV Beltline and Ext~hded Beltline Materials .......... 3-5 Table 3-2 Best-Estimate Cu and Ni Weight Percent Values, Initial RTNnT Values, and Initial USE Values for the North Anna Unit 2 RPV Beltline and Extended Beltline Materials .......... 3-6 Table 3-3 Initial RTNDT Values for the North Anna Units 1 and 2 Replacement Reactor Vessel Closure Head and Vessel Flange Materials ................................................................................... 3-8 Table 3-4 Calculation of Position 2.1 CF Values for North Anna Unit 1 Surveillance Materials .... 3-8 Table 3-5 Calculation of Position 2.1 CF Values for North Anna Units 1 and 2 Welds with Data from Other Plant Surveillance Programs .................................................................................. 3-9 Table 3-6 Calculation of Position 2.1 CF Values for North Anna Unit 2 Surveillance Materials

.................................. '. .................................................................................................... 3-10 Table 3-7 Summary of the North Anna Unit 1 RPV Beltline, Extended Beltline, and Surveillance Material CF Values based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 .................................................................................................................... 3-11 Table 3-8 Summary of the North Anna Unit 2 RPV Beltline, Extended Beltline, and Surveillance Material CF Values based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 .................................................................................................................... 3-12 Table 4-1 Calculation ofNorthAnna Unit 1 RTPTs Values for 72 EFPY (SLR) at the Clad/Base Metal Interface ........................................................................................................................... 4-3 Table 4-2 Calculation ofNorthAnna Unit 2 RTPTS Values for 72 EFPY (SLR) at the Clad/Base Metal Interface ........................................................................................................................... 4-6 Table 5-1 Predicted USE Values at 72 EFPY (SLR) for North Anna Unit 1 ................................... 5-3 Table 5-2 Predicted USE Values at 72 EFPY (SLR) for North Anna Unit 2 ................................... 5-5 Table 6.1-1 Calculation of the North Anna Unit 1 ART Values at the 1/4T Location for the Reactor Vessel Beltline Materials at 50.3 EFPY ........................................................................... 6-3 WCAP-18364-NP March2020 Revision 1

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Westinghouse Non-Proprietary Class 3 V Table 6.1-2 Calculation of the North Anna Unit 1 ART Values at the 3/4T Location for the Reactor Vessel Beltline Materials at 50.3 EFPY ........................................................................... 6-5 Table 6.1-3 Calculation of the North Anna Unit 1 ART Values for the Reactor Vessel Extended Beltline Materials at 50.3 EFPY .................................................................................................... 6-7 Table 6.1-4 Calculation of the North Anna Unit 2 ART Values at the 1/4T Location for the Reactor Vessel Beltline Materials at 52.3 EFPY ........................................................................... 6-9 Table 6.1-5 Calculation of the North Anna Unit 2 ART Values at the 3/4T Location for the Reactor Vessel Beltline Materials at 52.3 EFPY ......................................................................... 6-11 Table 6.1-6 Calculation of the North Anna Unit 2 ART Values for the Reactor Vessel Extended Beltline Materials at 52.3 EFPY .................................................................................................. 6-13 Table 6.1-7 Calculation of the North Anna Unit 1 ART Values at the 1/4T Location for the Reactor Vessel Beltline Materials at 72 EFPY ............................................................................ 6-15 Table 6.1-8 Calculation of the North Anna Unit 1 ART Values at the 3/4T Location for the Reactor Vessel Beltline Materials at 72 EFPY ............................................................................ 6-17 Table 6.1-9 Calculation of the North Anna Unit 1 ART Values for the Reactor Vessel Extended Beltline Materials at 72 EFPY .................................... :................................................................ 6-19 Table 6.1-10 Calculation of the North Anna Unit 2 ART Values at the 1/4T Location for the Reactor Vessel Beltline Materials at 72 EFPY ............................................ :............................... 6-2 l Table 6.1-11 Calculation of the North Anna Unit 2 ART Values at the 3/4T Location for the Reactor Vessel Beltline Materials at 72 EFPY ............................................................................ 6-23 Table 6.1-12 Calculation of the North Anna Unit 2 ART Values for the Reactor Vessel Extended Beltline Materials at 72 EFPY ..................................................................................................... 6-25 Table 6.2-1 Summary of the Limiting ART Values ........................................................................... 6-27 Table 7-1 North Anna Unit 1 Recommended Surveillance Capsule Withdrawal Schedule ............. 7-3 Table 7-2 North Anna Unit 2 Recommended Surveillance Capsule Withdrawal Schedule ............. 7-5 Table 7-3 Comparison of the North Anna Unit 1 Capsule Withdrawal Schedule withASTM E185-82 and the GALL Reports ..................................................................................................... 7-7 Table 7-4 Comparison of the North Anna Unit 2 Capsule Withdrawal Schedule,withASTM E185-82 and the GALL Reports ..................................................................................................... 7-9 Table A-1 Regulatory Guide 1.99, Revision 2, Credibility Criteria ....................... :........................ A-1 Table A.1-1 North Anna Unit 1 Surveillance Capsule Data Scatter about the Best-Fit Line Using All Available Surveillance Data ............................................................................................ A-3 Table A.2-1 North Anna Unit 2 Surveillance Capsule Data Scatter about the Best-Fit Line Using All Available Surveillance Data ............................................................................................ A-6 Table B-1 Evaluation of North Anna Units 1 and 2 ERG Limit Category ...................................... B-1 Table C-1 North Anna Units 1 and 2 Chemistry Factor Licensing Basis History .......................... C-2 WCAP-18364-NP March 2020 Revision 1

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Westinghouse Non-Proprietary Class 3 vi Table C-2 Conservatism Check for Position 1.1 for Non-Credible North Anna Unit 1 Surveillance Data ................................................................................................................................. C-4 Table C-3 Conservatism Check for Position 1.1 for Non-Credible North Anna Unit 2 Surveillance Data ................................................................................................................................. C-5 Table D-1 North Anna Units 1 and 2 Reactor Vessel Material which Use PWROG-17090-NP-A Generic Values ................................................................................................................ D-2 WCAP-18364-NP March 2020 Revision 1

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Westinghouse Non-Proprietary Class 3 vii LIST OF FIGURES Figure 2-1 North Anna Unit 1 - Axial Boundary of the LOE+ 17 n/cm2 Fast Neutron (E > 1.0 MeV)

Fluence Threshold in the +Z Direction (at 50.3 EFPYand 72 EFPY) ............................ 2-7 Figure2-2 North Anna Unit 2 -Axial Boundary of the l.OE+17 n/cm2 Fast Neutron (E> l.OMeV)

Fluence Threshold in the +Z Direction (at 52.3 EFPY and 72 EFPY) ............................ 2-9 Figure 3-1 RPV Base Metal Material Identifications for North Anna Units 1 and 2 ........................ 3-4 Figure 5-1 Regulatory Guide 1.99, Revision 2, Position 1.2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for North Anna Unit 1 at SLR (72 EFPY)

- - - - 5-7 Figure 5-2 Regulatory Guide 1.99, Revision 2, Position 2.2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for North Anna Unit 1 at SLR (72 EFPY)

- - - - 5-8 Figure 5-3 Regulatory Guide 1.99, Revision 2, Position 1.2 Predicted Decrease in Upper-ShelfEnergy as a Function of Copper and Fluence for North Anna Unit 2 at SLR (72 EFPY)

- - - - 5-9 Figure 5-4 Regulatory Guide 1.99, Revision 2, Position 2.2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for North Anna Unit 2 at SLR (72 EFPY)

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Westinghouse Non-Proprietary Class 3 viii ACRONYMS 1/4T one-quarter thickness 3/4T- three-quarter thickness 3D three dimensional ADAMS - Agencywide Documents Access and Management System ART Adjusted Reference Temperature ASME American Society of Mechanical Engineers ASTM - American Society for Testing and Materials B&PV - Boiler and Pressure Vessel BTP Branch Technical Position CF - Chemistry Factor CFR - Code of Federal Regulations CLB Current Licensing Basis COMS Cold Overpressure Mitigating System Cu-Copper EFPY - Effective Full Power Years EMA - Equivalent Margins Analysis EOC - End of Cycle EOL- End of License ERG Emergency Response Guideline(s)

EOLE End of License Extension FF Fluence Factor ft-lbs-Foot-Pounds ID - Inner Diameter IS - Intennediate Shell LOCA - Loss-of-Coolant Accident LS - Lower Shell LST - Lowest Service Temperature LTOPS - Low Temperature Overpressure Protection System Ni-Nickel NRC Nuclear Regulation Commission NUREG NRC technical report designation (Nyclear Regulatory Commission)

OD Outer Diameter WCAP-18364-NP March 2020 Revision 1

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Westinghouse Non-Proprietary Class 3 ix ACRONYMS - Continued P-T - Pressure-Temperature psi - pounds per square inch PTS - Pressurized Thermal Shock PWR - Pressurized Water Reactor PWROG - Pressurized Water Reactor Owners Group RPV - Reactor Pressure Vessel RV - Reactor Vessel

~RTNoT- Change in Reference Nil-Ductility Transition Temperature RTNDT - Reference Nil-Ductility Transition Temperature RTNoT(u)- Initial (or Unirradiated) Reference Nil-Ductility Transition Temperature RTPTs - Reference Temperature for Pressurized Thermal Shock SE - Safety Evaluation SLR - Subsequent ( or Second) License Renewal SPEO - Subsequent Period of Extended Operation SSC - Systems, Structures, and Components TLAA- Time-Limited Aging Analysis T NDT - Nil-Ductility Transition Temperature US - Upper Shell USE - Upper-Shelf Energy VEPCO - Virginia Electric and Power Company WOG - Westinghouse Owners Group Wt. % - Weight Percent WCAP-18364-NP March 2020 Revision 1

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Westinghouse Non-Proprietary Class 3 X EXECUTIVE

SUMMARY

This report presents the Time-Limited Aging Analyses (TLAA) for the North Anna Units 1 and 2 reactor pressure vessels (RPVs) in accordance with the requirements of the License Renewal Rule, 10 CFR Part

54. TLAAs are calculations that address safety-related aspects of the RPV within the bounds of the current 60-year license. These calculations must also be evaluated to account for an extended period of operation (80 years) also termed Subsequent (or Second) License Renewal (SLR) period or Subsequent Period of Extended Operation (SPEO).

North Anna Units 1 and 2 are currently licensed through 60 years of operation; therefore, with a 20-year license extension, the subsequent license renewal term is applicable through 80 years of operation. The evaluations in this report for 60 years of operation are applicable through 50.3 Effective Full Power Years (EFPY) for Unit 1 and 52.3 EFPY for Unit 2, which is deemed end of license extension (BOLE). Similarly, evaluations in this report performed at 80 years of operation are applicable through 72 EFPY for both Units, which is deemed the end of SLR. Updated neutron fluence evaluations were performed and documented in WCAP-18015-NP, as well as in Section 2 of this report. Updated neutron fluence evaluations were used to identify the North Anna Units 1 and 2 extended beltline materials and as input to the reactor vessel (RV) integrity evaluations in support of current plant operations and subsequent license renewal.

In addition to the RV integrity TLAA evaluations, the North Anna Units 1 and 2 surveillance data credibility evaluation is contained in Appendix A of this report. While not a TLAA, Appendix B provides an Emergency Response Guidelines (ERG) assessment for North Anna Units 1 and 2 for completeness.

Appendix C contains a summary of the North Anna licensing basis related to selection of chemistry factors (CFs) when surveillance data is available. Appendix D contains the justification for the use of PWROG-17090-NP-A per the stipulations in the NRC's Safety Evaluation.

A summary of results for the North Anna Units 1 and 2 TLAA is provided below. Based on the results of this TLAA evaluation, it is concluded that the North Anna Units 1 and 2 RVs will continue to meet regulatory requirements through the SLR period of operation.

Fluence The RV beltline and extended beltline neutron fluence values applicable to a postulated 20-year license renewal period were calculated for the North Anna Units 1 and 2 materials. The analysis methodologies used to calculate the North Anna Units 1 and 2 vessel fluence values satisfy the requirements set forth in Regulatory Guide 1.190. See Section 2 for more detai,ls. Industry efforts to benchmark and qualify calculated neutron fluence values in the extended beltline region are currently ongoing.

SLR Pressurized Thermal Shock All of the beltline and extended beltline materials, in the North Anna Units 1 and 2 RVs are projected to remain below the RTPTS screening criteria values of 270°F for base metal and/or longitudinal welds and 300°F for circumferentially oriented welds (per 10 CFR 50.61) through SLR (72 EFPY). See Section 4 for more details.

WCAP-18364-NP March 2020 Revision 1

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Westinghouse Non-Proprietary Class 3 xi SLR Upper-Shelf Energy Most of the beltline and extended beltline materials in the North Anna Units 1 and 2 RVs are projected to remain above the upper-shelf energy (USE) screening criterion value of 50 ft-lb (per 10 CFR 50, Appendix G), through SLR (72 EFPY). However, the Unit 2 Intermediate Shell Forging 04 material is projected to have a USE value below the screening criterion and requires an Equivalent Margins Analysis (EMA) to be performed. See Section 5 for more details.

EOLE and SLR Adjusted Reference Temperatures and EOLE P-T Limit Curves Applicability Check Adjusted Reference Temperatures (ARTs) are calculated for EOLE at 50.3 EFPY for Unit 1, 52.3 EFPY for Unit 2, and for SLR at 72 EFPY for both Units. The EOLE ART values are used to perform an applicability check on the existing pressure-temperature (P-T) limit curves for North Anna Units 1 and 2. With the consideration of TLAA fluence projections, revised Position 2.1 chemistry factor values, and recalculated initial RTNOT values, the applicability of the North Anna Units 1 and 2 cylindrical shell P-T limit curves currently in the Technical Specifications through 72 EFPY for both Units. Nozzle P-T limit curves were developed in WCAP-18363-NP with consideration of embrittlement for those extended beltline materials with fluence values that exceed 1 x 10 17 n/cm2 (E > 1.0 MeV) through 72 EFPY. As documented in WCAP-18363-NP, the North Anna Units 1 ancr2 cylindrical shell P-T limit curves will remain bounding through SLR.

Surveillance Capsule Withdrawal Schedule With consideration of a 20-year license renewal to 80 years of operation (72 EFPY), additional capsule withdrawals are recommended for both North Anna Units 1 and 2. At each unit, one capsule should be withdrawn at the refueling outage after a capsule reaches a fluence equivalent to the reactor vessel fluence at 80 years in order to meet the requirements of ASTM E185-82. To assist in asset management at each unit, an additional capsule should be withdrawn at the refueling outage after a capsule reaches a fluence equivalent to the reactor vessel fluence at 100 years. The surveillance capsule withdrawal schedules in Tables 7-1 and 7-2 for North Anna Units 1 and 2, respectively, identify when capsules should be withdrawn to meet these criteria. See Section 7 for more details.

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Westinghouse Non-Proprietary Class 3 1-1 1 TIME-LIMITED AGING ANALYSIS Time-Limited Aging Analyses (TLAAs) are those licensee calculations that:

1. Consider the effects of aging

2. Involve time-limited assumptions defined by the current operating term (e.g., 60 years)

3. Involve systems, structures, and components (SSCs) within the scope of license renewal

4. Involve conclusions or provide the basis for conclusions related to the capability of the SSC to perform its intended functions

5. Were determined to be relevant by the licensee in making a safety determination

6. Are contained or incorporated by reference in the current licensing basis (CLB)

The potential TLAAs for the reactor pressure vessel (RPV) are identified in Table 1-1 along with indication of whether or not they meet the six criteria of 10 CFR 54.3 (Reference 1) for TLAAs.

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Westinghouse Non-Proprietary Class 3 1-2 Table 1-1 Evaluation of Time-Limited Aging Analyses Per the Criteria of 10 CFR 54.3 Pressure-Pressurized Temperature Calculated Upper-Shelf Time-Limited Aging Analysis Thermal Limits for Fluence ShockCa) Energy Heatupand Cooldown Considers the Effects of Aging YES YES YES YES Involves Time-Limited Assumptions YES YES YES YES Defined by the Current Operating Term Involves SSC Within the Scope of YES YES YES YES License Renewal Involves Conclusions or Provides the Basis for Conclusions Related to the YES YES YES YES Capability of SSC to Perform Its Intended Function Determined to be Relevant by the Licensee in Making a Safety YES YES YES YES Determination Contained or Incorporated by Reference YES YES YES YES in the CLB Note:

(a) The limiting Pressurized Thermal Shock (PTS) values are used to determine the appropriate Emergency Response Guideline (ERG) Limits category for North Anna Units 1 and 2 through the end of the potential subsequent license extension period.

However, ERG limits are outside the scope of 10 CFR Part 54.3. ERG limits are discussed in Appendix B.

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Westinghouse Non-Proprietary Class 3 .2-1 2 CALCULATEDFLUENCE For the initial 60-year End of License Extension (EOLE) term, the North Anna Units 1 and 2 fracture toughness properties provide adequate margins of safety against vessel failure. However, as the reactor operates, neutron irradiation (fluence) reduces material fracture toughness. RPV integrity is assured by demonstrating that RPV material fracture toughness will remain at levels that resist brittle fracture throughout the period of Subsequent (or Second) License Renewal (SLR) operation. The first step in the analysis of vessel embrittlement is calculation of the neutron fluence that causes increased embrittlement.

Estimated RPV beltline and extended beltline fast neutron (E > 1.0 Me V) fluences at the end of 80 years of operation were calculated for North Anna Units 1 and 2. The analyses methodologies used to calculate the North Anna Units 1 and 2 RPV fluences satisfy the guidance set forth in Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" (Reference 2).

These methodologies have been approved by the U.S. NRC for the beltline region, i.e. materials directly surrounding the core and adjacent materials per 10 CFR 50, Appendix G (Reference 10), which are projected to experience the highest fluence. The methodologies, along with the NRC safety evaluation, are contained in detail in WCAP-14040-A (Reference 3). For North Anna Units 1 and 2, the beltline region has traditionally included the upper, intermediate, and lower shell forgings, and the circumferential welds between these components. Note that while a consistent approach is applied to the extended beltline, there is, at present, no generically-approved methodology for performing neutron fluence evaluations of the reactor vessel extended beltline. The traditional beltline and extended beltline materials are identified by heat numbers in Tables 2-3 and 2-4 and Figures 2-1 and 2-2.

Materials exceeding a fast neutron (E > 1.0 MeV) fluence of 1.0 x 10 17 n/cm2 at the end of the SLR period are evaluated for changes in fracture toughness. RPV materials that are not traditionally plant-limiting because of low levels of neutron radiation must now be evaluated to determine the accumulated fluen_ce at SLR. Therefore, fast neutron (E > 1.0 Me V) fluence calculations were performed for the North Anna Units 1 and 2 RPV circumferential welds (lower shell to lower vessel head, intermediate shell to lower shell, and upper shell to intermediate shell), centerline of the inlet and outlet nozzle forging to vessel shell welds at the lowest extent, 1/4T flaw location in the inlet and outlet nozzle, and forgings (lower shell, intermediate shell, and upper shell), to determine if they will exceed a fast neutron (E > 1.0 MeV) fluence of 1.0 x 10 17 n/cm2 at SLR. The materials that exceed the 1.0 x 10 17 n/cm2 fast neutron (E > 1.0 MeV) fluence threshold, and were not evaluated in past analyses of record as part of the traditional beltline, are referred to as extended beltline materials in this report and are evaluated to determine the effect of neutron irradiation embrittlement during the SLR period. The need to evaluate these extended beltline mat~rials was previously identified during Dominion submittal and NRC review of the P-T limit curves with vacuum refill (Reference 26).

In performing the fast neutron exposure evaluations for the North Anna Units 1 and 2 reactor vessels, a series of fuel-cycle-specific forward transport calculations were carried out using the following two-dimensional/one-dimensional fluence rate synthesis technique:

<p(r,z)

<p(r,0,z) = <p(r,0) x <p(r)

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Westinghouse Non-Proprietary Class 3 2-2 where <p(r, 0, z) is the synthesized 3D neutron fluence rate distribution, <p(r, 0) is the transport solution in r,0 geometry, <p (r, z) is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and <p(r) is the one-dimensional solution for a cylindrical reactor model using the same source per unit height as that used in the r,0 two-dimensional calculation. This synthesis procedure was carried out for each operating cycle at North Anna Units 1 and 2.

All of the transport calculations were carried out using the DORT discrete ordinates code (Reference 4) with the BUGLE-96 cross-section library (Reference 5). The BUGLE-96 library provides a coupled 47-neutron-, 20-gamma-ray-group cross-section data set produced specifically for light water reactor applications. In these analyses, anisotropic scattering was treated with a Ps Legendre expansion and the 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.

The calculations for fuel Cycles 1 through 24 for North Anna Unit 1 and fuel Cycles 1 through 23 for North Anna Unit 2 determine the neutron exposure of the pressure vessel and surveillance capsules based on completed fuel cycles. For North Anna Unit 1, projections for Cycle 25 and beyond, up to and including BOLE (50.3 EFPY) and SLR (conservatively set to 72 EFPY), were based on the uprated core power level of 2940 MWt and the uprated Cycle 24. For North Anna Unit 2, projections for Cycle 24 and beyond, up to and including BOLE (52.3 EFPY) and SLR (conservatively set to 72 EFPY), were based on the uprated core power level of2940 MWt and the uprated Cycle 23. These projections are used to perform the reactor vessel integrity evaluation contained herein. Projected results will remain valid as long as future plant operation is consistent with these conservative inputs.

Table 2-1 gives the North Anna Unit 1 calculated fast neutron (E > 1.0 Me V) fluences at the capsule locations including all withdrawn surveillance capsules (Capsules V, U, and W). Table 2-2 gives the North Anna Unit 2 calculated fast neutron (E > 1.0 Me V) fluences at the capsule locations including all withdrawn surveillance capsules (Capsules V, U, and W). These fast neutron (E > 1.0 Me V) fluences were calculated using methodologies that follow the guidance of Regulatory Guide 1.190.

Selected results for the pressure vessel from the neutron transport analyses are provided in Tables 2-3 and 2-4 for North Anna Units 1 and 2, respectively. Calculated fast neutron (E > 1.0 MeV) fluence results for reactor vessel materials, on the pressure vessel clad/base metal interface, are provided for the nominal end of cycle (EOC) 24 for North Anna Unit 1 (29. 7 EFPY) and nominal EOC 23 for North Anna Unit 2 (28.1 EFPY), as well as projected fluence results up to 72 EFPY, which corresponds to the North Ann;:t Units 1 and 2 80-year plant life.

From Table 2-3 it is observed that one outlet nozzle and two inlet nozzles have fast neutron (E > 1.0 MeV) fluence greater than 1.0 x 10 17 n/cm2 at the nozzle forging to vessel shell weld centerline and one inlet nozzle has a fast neutron (E > 1.0 MeV) fluence greater than 1.0 x 10 17 n/cm2 at the postulated 1/4T nozzle flaw location at 72 EFPY for North Anna Unit 1. From Table 2-4, it is observed that one outlet nozzle and two inlet nozzles have fast neutron (E > 1.0 Me V) fluence greater than 1.0 x 10 17 n/cm2 at the nozzle forging to vessel shell weld centerline and one inlet nozzle has a fast neutron (E > 1.0 MeV) fluence greater than 1.0 x 10 17 n/cm 2 at the postulated 1/4.T nozzle flaw location at 72 EFPY for North Anna Unit 2. Tables 2-3 and 2-4 indicate that the lower shell to lower vessel head circumferential weld will remain below 1.0 x 10 17 n/cm2 through SLR for both North Anna Units 1 and 2.

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Westinghouse Non-Proprietary Class 3 2-3 Figure 2-1 shows the axial boundary of the 1.0 x 10 17 n/cm 2 fluence threshold (at 50.3 EFPYand 72 EFPY) as a function of azimuthal position (Z versus 0) for North Anna Unit 1, whereas Figure 2-2 shows the same information (at 52.3 EFPY and 72 EFPY) for North Anna Unit 2. It is noted that the nozzle materials located above the nozzle centerline remain below 1.0 x 10 17 n/cm2 through 72 EFPY. Likewise, the lower shell to lower head circumferential weld remains out of the beltline region through 72 EFPY. The data used to generate Figures 2-1 and 2-2 is tabulated in Appendices A and B of WCAP-18015-NP (Reference 6),

respectively.

' WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-4 Table 2-1 Calculated Fast Neutron (E > 1.0 MeV) Fluence at the Surveillance Capsule Center for North Anna Unit 1<a)

Cumulative Surveillance Capsules [n/cm 2] Clad/Base Time Metal Cycle [EFPY] 15° 25° 35° 45° 35°/15°(*) 35°/25°(1) Interface 1 1.1 3.06E+ 18(b) 2.01E+l8 l.37E+18 l.07E+18 l.37E+18 l.37E+18 1.90E+18 2 1.9 5.45E+18 3.54E+l8 2.38E+18 l.84E+18 2.38E+ 18 2.38E+18 3.39E+l8 3 2.9 7.65E+18 5.0IE+18 3.34E+18 2.58E+18 3.34E+18 3.34E+l8 4.78E+18 4 3.8 l.00E+19 6.39E+18 4.20E+18 3.24E+18 4.20E+l8 4.20E+18 6.31E+18 5 4.8 1.18E+ 19 7.59E+18 5.03E+l8 3.91E+18 5.03E+18 5.03E+l8 7.42E+18 6 5.9 l.40E+ 19 9.14E+18Cc) 6.06E+18 4.70E+18 6.06E+18 6.06E+18 8.75E+18 7 7.1 l.64E+l9 l.08E+19 7.19E+18 5.58E+18 7.19E+18 7.19E+18 l.0IE+19 8 8.4 l.89E+ 19 l.25E+19 8.41E+18 6.54E+18 8.41E+l8 8.41E+18 l.14E+19 9 9.8 2.14E+19 l.42E+l9 9.59E+18 7.46E+l8 9.59E+18 9.59E+18 l.29E+19 IO 11.1 2.37E+19 l.59E+19 l.08E+l9 8.40E+ 18 l.08E+19 l.08E+19 l.41E+19 11 12.4 2.59E+l9 l.75E+19 1.19E+19 9.33E+18 l.19E+ 19 1.19E+ 19 l.54E+19 12 13.5 2.79E+19 l.90E+ 19 l.29E+ 19 l.0IE+19 l.29E+ 19 1.29E+l9 l.65E+19 13 14.8 3.02E+19 2.05E+I9(d) l.40E+l9 l.09E+19 1.40E+19 l.40E+19 l.77E+l9 14 16.2 3.26E+19 2.23E+19 l.52E+19 1.19E+l9 l.52E+19 l.52E+l9 l.90E+l9 15 17.5 3.50E+19 2.41E+l9 l.65E+l9 l.29E+l9 1.76E+19 l.70E+19 2.03E+l9 16 18.9 3.74E+19 2.58E+l9 l.77E+ 19 l.38E+ 19 2.01E+l9 l.8.8E+19 2.16E+19 17 20.2 3.98E+19 2.76E+l9 l.89E+ 19 l.48E+ 19 2.24E+19 2.05E+19 2.28E+19 18 21.6 4.22E+19 2.94E+l9 2.02E+l9 l.58E+l9 2.49E+ 19 2.23E+19 2.41E+l9 19 23.0 4.47E+l9 3.11E+19 2.14E+ 19 l.68E+19 2.73E+19 2.41E+l9 2.55E+l9 20 24.4 4.71E+19 3.29E+l9 2.27E+19 l.78E+ 19 2.98E+19 2.58E+19 2.68E+19 21 25.8 4.95E+19 3.47E+19 2.39E+19 l.88E+19 3.21E+19 2.76E+19 2.81E+l9 22 26.9 5.13E+19 3.60E+19 2.49E+ 19 l.96E+ 19 3.39E+ 19 2.89E+l9 2.90E+19 23 28.3 5.36E+19 3.75E+19 2.60E+19 2.05E+19 3.62E+19 3.04E+19 3.03E+19 24 29.7 5.59E+19 3.92E+19 2.73E+19 2.16E+ 19 3.86E+19 3.22E+19 3.16E+19 Projected 50.3 9.1 IE+ 19 6.48E+19 4.64E+19 3.74E+19 7.37E+19 5.77E+19 5.13E+19 Projected 54.0 9.74E+19 6.94E+l9 4.98E+19 4.03E+19 8.00E+19 6.23E+19 5.48E+19 Projected 72.0 l.28E+20 9.17E+19 6.65E+19 5.41E+19 l.11E+20 8.46E+ 19 7.20E+19 Notes:

(a) Information is taken from WCAP-18015-NP (Reference 6).

(b) Capsule V was withdrawn at the end-of-cycle 1.

(c) Capsule U was withdra~ at the end-of-cycle 6.

(d) Capsule W was withdrawn at the end-of-cycle 13.

(e) Capsule Z was moved at the end-of-cycle 14.

(f) Capsule Twas moved at the end-of-cycle 14.

WCAP-18364-NP March2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-5 Table 2-2 Calculated Fast Neutron (E > 1.0 MeV) Fluence at the Surveillance Capsule Center for North Anna Unit 2<a)

Cumulative Surveillance Capsules [n/cm 2) Clad/Base Time Metal Cycle [EFPY] 15° 25° 35° 45° 35°/15°(e) 35°/25°(t) Interface 1 1.0 2.86E+ 18(b) 1.87E+ 18 1.27E+l8 9.96E+l7 1.27E+18 1.27E+ 18 1.78E+18 2 1.6 4.68E+18 3.05E+l8 2.06E+18 1.61E+ 18 2.06E+18 2.06E+18 2.92E+18 3 2.7 6.99E+18 4.86E+18 3.34E+l8 2.62E+18 3.34E+18 3.34E+ 18 4.20E+18 4 3.8 9.44E+18 6.53E+18 4.51E+18 3.57E+18 4.51E+18 4.51E+18 5.68E+18 5 5.0 l.19E+19 8.20E+18 5.68E+l8 4.48E+18 5.68E+18 5.68E+18 7.07E+18 6 6.2 1.42E+19 9.85E+I8(c) 6.80E+18 5.35E+18 6.80E+18 6.80E+l8 8.43E+18 7 7.5 1.65E+19 l.14E+19 7.88E+18 6.21E+18 7.88E+18 7.88E+18 9.71E+18 8 8.7 l.87E+ 19 l.30E+ 19 8.97E+18 7.06E+18 8.97E+18 8.97E+18 l.10E+19 9 9.9 2.09E+19 1.45E+19 1.01E+19 8.00E+18 1.01E+l9 1.01E+l9 1.22E+ 19 10 11.3 2.30E+l9 1.60E+l9 l.12E+l9 8.91E+18 l.12E+ 19 l.12E+19 l.34E+19 11 12.5 2.52E+19 l.75E+19 1.23E+19 9.71E+l8 l.23E+l9 l.23E+19 1.46E+ 19 12 13.8 2.75E+19 1.91E+19 l.33E+ 19 1.06E+19 l.33E+l9 l.33E+19 1.59E+ 19 13 15.1 2.99E+19 2.08E+ 19(d) 1.45E+ 19 l.15E+19 1.45E+ 19 1.45E+l9 1.71E+ 19 14 16.5 3.22E+19 2.25E+19 l.57E+19 l.24E+19 1.68E+ 19 1.62E+19 1.84E+19 15 17.7 3.44E+l9 2.41E+l9 1.68E+19 l.32E+19 1.91E+19 1.78E+19 1.96E+19 16 19.0 3.65E+l9 2.56E+19 1.78E+19 1.41E+19 2.35E+19 l.93E+ 19 2.07E+19 17 20.3 3.90E+19 2.75E+19 l.91E+l9 l.50E+l9 2.60E+19 2.12E+19 2.20E+l9 18 21.6 4.15E+19 2.93E+19 2.03E+19 l.59E+19 2.85E+l9 2.30E+l9 2.33E+ 19 19 22.9 4.37E+19 3.09E+19 2.14E+19 l.68E+19 3.06E+l9 2.46E+19 2.44E+19 20 24.3 4.59E+19 3.25E+l9 2.26E+l9 l.77E+ 19 3.29E+19 2.62E+19 2.56E+19 21 25.5 4.80E+19 3.41E+19 2.36E+19 l.86E+19 3.50E+l9 2.78E+19 2.67E+19 22 26.8 5.03E+19 3.57E+ 19 2.48E+ 19 l.95E+ 19 3.72E+19 2.94E+19 2.80E+19 23 28.1 5.26E+19 3.73E+19 2.59E+19 2.03E+19 3.96E+19 3.10E+l9 2.92E+19 Projected 52.3 9.75E+19 6.82E+l9 4.63E+19 3.59E+l9 8.44E+19 6.19E+19 5.36E+19 Projected 54.0 l.01E+20 7.04E+19 4.78E+19 3.70E+19 8.76E+l9 6.41E+19 5.53E+19 Projected 72.0 l.34E+20 9.33E+19 6.30E+ 19 4.85E+19 l.21E+20 8.70E+19 7.34E+19 Notes:

(a) Information is taken from WCAP-18015-NP (Reference 6).

(b) Capsule V was withdrawn at the end-of-cycle 1.

(c) Capsule U was withdrawn at the end-of-cycle 6.

(d) Capsule W was withdrawn at the end-of-cycle 13.

(e) Capsule Z was moved at the end-of-cycle 13.

(f) Capsule Twas moved at the end-of-cycle 13.

\

WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-6 Table 2-3 North Anna Unit I -Maximum Fast Neutron (E > 1.0 MeV) Fluence Experienced by the Pressure Vessel Materials in the Beltline and Extended Beltline Regions Neutron Fluence [n/cm 2]

Material Region 29.7 EFPY 50.3 EFPY 54EFPY 72 EFPY<a)

Postulated 1/4T Flaw in Inlet Nozzle Nozzle 09 (Ht. # 990290-11) Extended BeltlineCh) 1.65E+ 16 2.92E+l6 3.15E+l6 4.25E+l6 Nozzle IQ(b) (Ht. # 990290-12) Extended Beltline 6.13E+16 1.04E+17 1.11E+17 l.48E+l7 Nozzle 11 (Ht.# 990268-11) Extended Beltline<h) 2.29E+16 3.94E+16 4.24E+l6 5.68E+l6 Postulated 1/4T Flaw in Outlet Nozzle Nozzle 12 (Ht.# 990290-31) Extended BeltlineCh) 3.62E+l6 6.12E+l6 6.57E+l6 8.75E+16 Nozzle 13 (Ht. # 990290-22) Extended Beltline(h) 9.74E+15 1.72E+ 16 1.86E+16 2.51E+16 Nozzle 14 (Ht.# 990290-21) Extended BeltlineCh) l.35E+16 2.33E+l6 2.50E+l6 3.35E+16 Centerline of the Inlet Nozzle Forging to Vessel Shell Welds - Lowest Extent(il I Nozzle 09 Extended BeltlineCh) 3.50E+ 16 6.17E+l6 6.65E+l6 8.98E+16 Nozzle l0(d) Extended Beltline l.30E+ 17 2.19E+17 2.35E+17 3.13E+ 17 Nozzle 11 (e) Extended Beltline 4.85E+l6 8.33E+16 8.95E+l6 1.20E+ 17 Centerline of the Outlet Nozzle Forging to Vessel Shell Welds - Lowest ExtentCi)

Nozzle 12<c) Extended Beltline 7.53E+16 1.27E+17 l.37E+17 1.82E+ 17 Nozzle 13 Extended BeltlineCh) 2.03E+16 3.59E+16 3.87E+16 5.22E+16 Nozzle 14 Extended BeltlineCh) 2.82E+16 4.84E+16 5.20E+16 6.97E+16 Upper Shell (Ht.# 990286 / 295213) Beltline 1.30E+18 2.15E+18 2.30E+l8 3.04E+18 Upper Shell to Intermediate Shell Circumferential Weld Beltline 1.50E+18 2.48E+18 2.66E+18 3.51E+18 (Ht. # 25295 & 4278)

Intermediate Shell Beltline 3.11E+19 5.03E+19 5.39E+l9 7.07E+19

<Ht.# 990311 I 298244)

Intermediate Shell to Lower Shell Beltline 3.09E+19 5.02E+19 5.36E+l9 7.04E+19 Circumferential Weld (Ht.# 25531)

Lower Shell (Ht. # 990400 / 292332) Beltline 3.16E+19 5.13E+19 5.48E+l9 7.20E+19 Lower Shell to Lower Vessel Head Outside Beltline < 1.00E+17 < 1.00E+17 < 1.00E+17 < 1.00E+l7 Circumferential W eld(g)(i)

Notes:

(a) Corresponds to 80 years of life.

(b) 1/4 T Flaw in Inlet Nozzle Inlet 10 is projected to reach I.OE+l7 n/cm2 at approximately 48.5 EFPY; which corresponds to December 26, 2034(!)_

(c) Outlet Nozzle 12 is projected to reach I.OE+ 17 n/cm2 at approximately 39.5 EFPY; which corresponds to June 6, 2025(0.

(d) Inlet Nozzle 10 reached I.OE+ 17 n/cm2 at approximately 22.4 EFPY, which occurred during Cycle 19.

(e) Inlet Nozzle 11 is projected to reach I.OE+l7 n/cm2 at approximately 60.3 EFPY; which corresponds to May I, 2047(0_

(f) Note, the dates provided in notes b, c, and e are approximations based on an 18 month cycle and average outage time of25 days.

(g) The lower shell to lower vessel head circumferential weld is not modeled, it is known to be below the IE+ 17 n/cm2 fast neutron fluence threshold due to the fact that: it is 32 cm further from the core midplane than the above-core threshold location at 72 EFPY, and that the coolant below the core is cooler than the coolant above the core, which increases the density and shielding effects, reducing the fluence below the core relative to above the core.

(h) Component is conservatively included in the "Extended Beltline" even though its projected SLR fluence is less than IE+ 17 n/cm2 (E > 1.0 Me V) because, either a component at the same axial elevation meets the "Extended Beltline" fluence criterion, or the same component meets the fluence criterion at a lower elevation.

(i) The specific heat numbers of these welds could not be identified in the available information.

WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-7 400.000 ,

Inlet/Outlet Nozzle Postulated 1/4T Flaw Location 1E+17 n/cm2 threshold at 50.3 EFPV X Lowest Extent of the Nozzle to Shell Weld - 1E+17 n/cm2 threshold at 72.0 EFPY 326.7 (nozzle centerline)

Nozzle Shell Forging 300.000 (Heat No 990286/295213 )

276.s*

268.3**

263.8***

2s3.s****

- - - - - - - - -Hotzle5114! tolntemM!diateShelOfdli'llfl!!81 !Weld HeatNo~IIICl42711'____ _ _ _ _ _ _ __

200.000 182 .88 (top of core) 100.000 i Intermediate Shell (Heat No.990311/298244)

E 8

+I

~ 0.000 GI (core midplane)

Q ii

~

-100.000 J Lower Shell (Heat No. 990400/292332)

-200.000

182.88 (bottom of core)

-300.000 er.

45 90 135 180 225 270 315 360 Azimuthal Location (degree)

Outlet Nozzle 1/4 T flaw

-400.000 ** Inlet Nozzle 1/4 T flaw

- - Outlet Nozzle Forging to Vessel Shell Welds

Lowest Extent

- - - inlet Nozzle Forging to Vessel Shell Welds - Lowest Extent Figure 2-1 North Anna Unit 1-Axial Boundary of the 1.0E+17 n/cm 2 Fast Neutron (E > 1.0 MeV) Fluence Threshold in the +Z Direction (at 50.3 EFPY and 72 EFPY)

WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-8 Table 2-4 North Anna Unit 2-Maximum Fast Neutron (E > 1.0 MeV) Fluence Experienced b,y thPe ressure Viesse IMateria s m thBir

1* e et me an dExten ddBir e et me R. egions Neutron Fluence [n/cm 2]

Material Region 72 EFPY<a) 28.1 EFPY 52.3 EFPY 54 EFPY Postulated 1/4T Flaw in Inlet Nozzle Nozzle 09 (Ht. # 990426) Extended Beltline<hl 1.56E+16 2.85E+16 2.94E+16 3.90E+16 Nozzle IQ(bl(Ht. # 54567-2) Extended Beltline 5.69E+16 1.07E+17 l.11E+17 1.48E+ 17 Nozzle 11 (Ht. # 54590-2) Extended Beltline<hl 2.17E+16 4.05E+16 4.19E+16 5.59E+16 Postulated 1/4T Flaw in Outlet Nozzle Nozzle 12 (Ht.# 990426-22) Extended Beltline<h) 3.36E+16 6.33E+16 6.54E+16 8.75E+16 Nozzle 13 (Ht.# 990426-31) Extended Beltline<hl 9.18E+l5 1.68E+16 1.73E+l6 2.30E+ 16 Nozzle 14 (Ht.# 791291) Extended Beltline<hl 1.28E+ 16 2.39E+16 2.47E+l6 3.30E+ 16 Centerline of the Inlet Nozzle Forging to Vessel Shell Welds - Lowest Extent (Ht.# 8816, 20459, & 27622)

Nozzle 09 Extended Beltline<h) 3.29E+16 6.03E+l6 6.22E+16 8.26E+16 Nozzle IQ(d) Extended Beltline l.21E+l 7 2.27E+17 2.35E+17 3.14E+17 Nozzle 11<*l Extended Beltline 4.60E+16 8.58E+ 16 8.86E+16 l.18E+ 17 Centerline of the Outlet Nozzle Forging to Vessel Shell Welds - Lowest Extent (Ht.# 8816, 20459, & 27622)

Nozzle 12<cl Extended Beltline 6.99E+l6 l.32E+17 l.36E+17 1.82E+ 17 Nozzle 13 Extended Beltline<h) 1.91E+ 16 3.50E+16 3.61E+l6 4.79E+16 Nozzle 14 Extended Beltline<hl 2.67E+16 4.98E+16 5.14E+16 6.87E+16 Upper Shell (Ht.# 990598 / 291396) Beltline 1.20E+18 2.23E+18 2.30E+18 3.07E+l8 Upper Shell to Intermediate Shell Circumferential Weld Beltline l.38E+ 18 2.58E+l8 2.66E+l8 3.55E+18 (Ht.# 4278 & 801)

Intermediate Shell Beltline 2.87E+19 5.25E+ 19 5.42E+19 7.20E+19 (Ht. # 990496 I 292424)

Intermediate Shell to Lower Shell Circumferential Weld Beltline 2.86E+l9 5.24E+19 5.41E+l9 7.18E+l9 (Ht. #716126)

Lower Shell (Ht. # 990533 / 297355) Beltline 2.92E+19 5.36E+19 5.53E+19 7.34E+19 Lower Shell to Lower Vessel Head Outside Beltline < 1.00E+17 < 1.00E+17 < l.00E+l7 < l.00E+17 Circumferential Weld(glffit. # 716126)

Notes:

(a) Corresponds to 80 years oflife.

(b) Postulated 1/4T Flaw in Inlet Nozzle Inlet 10 is projected to reach I.OE+ 17 n/cm2 at approximately 48.8 EFPY; which corresponds to May 27, 2036(!)_

(c)' Outlet Nozzle 12 is projected to reach I.OE+ 17 n/cm2 at approximately 39.8 EFPY; which corresponds to February 4, 2027<!)_

(d) Inlet Nozzle 10 reached I.OE+ 17 n/cm2 at approximately 23.1 EFPY, which occurred during Cycle 20.

(e) Inlet Nozzle 11 is projected to reach I.OE+ 17 n/cm2 at approximately 60.9 EFPY; which corresponds to February 12, 2049(!)_

(+/-) }!~te, the dates provided in notes b, c, and e are approximations based on an 18 month cycle and average outage time of25 days.

(g) The lower shell to lower vessel head circumferential weld is not modeled, it is known to be below the IE+ 17 n/cm2 fast neutron fluence threshold due to the fact that: it is 32 cm further from the core midplane than the above-core threshold location at 72 EFPY, and that the coolant below the core is cooler than the coolant above the core, which increases the density and shielding effects, reducing the fluence below the core relative to above the core.

(h) Component is conservatively included in the "Extended Beltline" even though its projected SLR fluence is less than IE+ 17 n/cm2 (E > 1.0 Me V) because, either a component at the same axial elevation meets the "Extended Beltline" fluence criterion, or the same component meets the fluence criterion at a lower elevation.

WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-9 400.000

Inlet/Outlet Nozzle Postulated 1/4T Flaw Location 1E+17 n/cm2 threshold at 52.3 EFPY X Lowest Extent of the Nozzle to Shell Weld - 1E+17 n/cm2 threshold at 72.0 EFPY 326. 7 (nozzle centerr.ne)

Nozzle Shell Forging 300.000 (Heat No. 990598/291396}

276.s*

268.3**

263.8***

253.s****

t::========.?mE ZZJ!

le~sne1101ntenneo1,re 200.000 l 182.88 {top of core) 100.000 J Intermediate Shell

{Heat No. 990496/292424) s

+=

!: 0.000 QI (core midplane) ii:i ii

-100.000 Lower Shell (Heat No. 990533/297355)

-200.000

182.88(bottomofcore1

-300.000 45 90 135 180 225 270 315 360 Azimuthal Location (degree}

Outlet Nozzle 1/4 T flaw

-400.000 .. Inlet Nozzle 1/4 Tflaw

- - Outlet Nozzle Forging to Vessel Shell Welds - Lowest Extent

*** 1nlet Nozzle Forging to Vessel Shell Welds

Lowest Extent Figure 2-2 North Anna Unit 2 -Axial Boundary of the 1.0E+17 n/cm 2 Fast Neutron (E > 1.0 MeV) Fluence Threshold in the +Z Direction (at 52.3 EFPY and 72 EFPY)

WCAP-18364-NP March2020 Revision I

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 3-1 3 MATERIAL PROPERTY INPUT The North Anna Units 1 and 2 beltline materials consist ofone (1) Intermediate Shell (IS) Forging, one (1)

Lower Shell (LS) Forging, one (1) Upper Shell (US) Forging (also termed nozzle shell forging), and two (2) circumferential welds: the IS to LS Circumferential Weld and the US to IS Circumferential Weld. The reactor vessel (RV) forgings and weld materials are shown in Figure 3-1 for North Anna Units 1 and 2.

Used in conjunction with the fluence data in Tables 2-3 and 2-4, and Figures 2-1 and 2-2, the beltline and extended beltline materials are identified as shown in Tables 3-1 and 3-2.

The North Anna Unit 1 surveillance forging material was made from reactor vessel Lower Shell Forging 03, Heat# 990400 / 292332. The North Anna Unit I reactor vessel beltline IS to LS Circumferential Weld was fabricated using weld wire Heat# 25531, Flux Type SMIT 89, Flux Lot Number 1211. The weld material in the North Anna Unit 1 surveillance program was fabricated with the same material heat, flux type, and flux lot number as reactor vessel beltline IS to LS Circumferential Weld. The outer diameter (OD) 94% of the US to IS Circumferential Weld was fabricated with weld wire Heat# 25295, Flux Type SMIT 89, Flux Lot Number 1170 and the inner diameter (ID) 6% was fabricated with weld wire Heat #

4278, Flux Type SMIT 89, Flux Lot Number 1211. Surveillance data does not exist for Heat# 25295 or Heat# 4278 in the North Anna Unit 1 reactor vessel surveillance program; however weld wire Heat# 25295 or Heat# 4278 were included in the surveillance programs of other plants as summarized in Table 3-5.

The North Anna Unit 2 surveillance forging material was made from reactor vessel Intermediate Shell Forging 04, Heat# 990496 / 292424. The North Anna Unit 2 reactor vessel beltline IS to LS Circumferential Weld was fabricated using weld wire Heat# 716126, LW320 Flux Type, Flux Lot Number 26. The weld material in the North Anna Unit 2 surveillance program was fabricated with the same material heat, flux type, and flux lot number as reactor vessel beltline IS to LS Circumferential Weld. The OD 94% of US to IS Circumferential Weld was fabricated with weld wire Heat# 4278, Flux Type SMIT 89, Flux Lot Number 1211. Surveillance data does not exist for Heat# 4278 in the North Anna Unit 2 reactor vessel surveillance program; however, as previously stated, it was included in the surveillance programs of other plants, as summarized in Table 3-5. The remaining 6% of the US to IS Circumferential Weld was fabricated from weld wire Heat# 801, SMIT 89 Flux Type, Flux Lot Number 1211. Surveillance data does not exist for Heat# 801.

Based on the results of Section 2 of this report, the materials that exceeded the Ix 10 17 n/cm2 (E > 1.0 MeV) threshold at 72 EFPY are considered to be the North Anna Units 1 and 2 extended beltline materials and are evaluated to determine their impact on the proposeq SLR period of operation. The North Anna Units 1 and 2 reactor vessels contain three (3) Inlet Nozzles, three (3) Outlet Nozzles, three (3) Inlet Nozzle to US Welds, and three (3) Outlet Nozzle to US Welds per Unit. Only the forgings and welds corresponding to the North Anna Units 1 and 2 Inlet Nozzles 10, Inlet Nozzles 11, and Outlet Nozzles 12 are predicted to experience neutron fluence greater than 1.0 x 10 17 n/cm 2 at SLR. Only those materials with a fluence greater than 1 x 10 17 n/cm2 (E > 1.0 MeV) at SLR require the effects of embrittlement to be included ~hen evaluating the reactor vessel integrity.

For the Unit I Inlet/Outlet Nozzle to US Welds, the heat numbers, flux type, and flux lot numbers of these welds could not be identified in the available information; however, these welds were fabricated at the Rotterdam Dockyard Company (Rotterdam). Therefore, conservative generic/bounding properties from WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 3-2 PWROG-17090-NP-A (Reference 12) are used. The Unit 2 Inlet/Outlet Nozzle to US Welds were fabricated using Heat# 8816, Flux Type LW320, Lot Number 28; Heat# 20459, Flux Type LW320, Lot Number 26; and Heat# 27622, Flux Type LW320, Lot Numbers 26 & 28. The records do not identify which weld heats are associated with which specific nozzles. Therefore, the bounding material properties (which consider all available data, as documented in PWROG-18005-NP [Reference 7]) will be conservatively associated with all North Anna Unit 2 nozzle welds. Justification for the use of PWROG-17090-NP-A, consistent with the NRC Safety Evaluation, is presented in Appendix D.

The unirradiated material property inputs used for the RV integrity evaluations herein are contained in PWROG-18005-NP (Reference 7). PWROG-18005-NP defined or redefined many of the material properties and chemistry values using the most up-to-date methodologies and all available data; therefore, the values utilized herein supersede previously documented values. The sources and methods used in the determination of the chemistry factors and the fracture toughness properties are summarized below.

Chemical Compositions The best-estimate copper (Cu) and nickel (Ni) chemical compositipns for the North Anna Units 1 and 2 beltline and extended beltline materials are presented in Tables 3-1 and 3-2. The best-estimate weight percent copper and nickel values for the beltline and extended beltline materials were previously reported in PWROG-18005-NP and were included in RV integrity evaluations as part of this TLAA effort.

Fracture Toughness Properties The initial fracture toughness properties (initial RTNoT and initial USE) of most of the RV forging materials were originally determined using NUREG-0800, Branch Technical Position (BTP) 5-3 Position 1.1 (Reference 8) methodology. The exceptions are the North Anna Units 1 and 2 IS Forging 04, and LS Forging 03 which were determined using the ASME Code,Section III (Reference 9) methods. Many of the beltline and extended beltline fracture toughness properties were updated per ASME Section III and NUREG-0800, BTP 5-3 Position 1.1 methodologies, as described in PWROG-18005-NP. The most up-to-date initial RTNoT and initial USE values are documented in PWROG-18005-NP for North Anna Units 1 and 2. The beltline and extended beltline material properties of the North Anna Units 1 and 2 reactor vessels are presented in Tables 3-1 and 3-2 herein.

The initial RTNDT values of the reactor vessel flange and closure head serve as input to the P-T limit curves "flange-notch" per 10 CFR 50, Appendix G (Reference 10). Since North Anna Units 1 and 2 share P-T Limit curves for operation, materials for both plants must be considered. The closure heads at both North Anna Units 1 and 2 have been replaced, and the initial RTNoT values of the North Anna Units 1 and 2 flange materials were confirmed in PWROG-18005-NP (Reference 7). The North Anna Unit 1 replacement closure head flange has an initial RTNDT value of -76°F, determined per ASME Code Section III, NB-2300.

The North Anna Unit 1 reactor vessel flange has an initial RTNDT of -22°F, calculated using the BTP 5-3 methodology. The North Anna Unit 2 replacement head flange has an initial RTNDT value of -49°F, determined per ASME Code Section III, NB-2300. The North Anna Unit 2 reactor vessel flange has an initial RTNDT of -22°F, calculated using the BTP 5-3 methodology. See Table 3-3 for a summary of the initial RTNDT values for these two components at each plant.

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Westinghouse Non-Proprietary Class 3 3-3 Chemistry Factor Values The chemistry factor (CF) values were calculated using Positions 1. 1 and 2.1 of Regulatory Guide 1.99, Revision 2 (Reference 11 ). Position 1.1 uses Tables 1 and 2 from the Regulatory Guide along with the best-estimate copper and nickel weight percent values (contained in Tables 3-1 and 3-2). Position 2.1 uses the surveillance capsule data from all capsules tested to date and surveillance data from other plants, as applicable. Credibility evaluations of the North Anna Units 1 and 2 surveillance data are provided in Appendix A of this report. The calculated capsule fluence values are provided in Tables 2-1 and 2-2 and are used to determine the Position 2.1 CFs as shown in Tables 3-4 and 3-6 for North Anna Units 1 and 2, respectively. In addition, North Anna Units 1 and 2 utilize weld materials which are included in the Sequoyah Units 1 and 2 surveillance programs. Table 3-5 calculates the Position 2.1 CFs from the Sequoyah Units 1 and 2 surveillance weld materials for use in North Anna Units 1 and 2 calculations. The credibility evaluations of the Sequoyah Units 1 and 2 surveillance data are contained in WCAP-17539-NP (Reference 17). Tables 3-7 and 3-8 summarize the Positions 1.1 and 2.1 CF values determined for the North Anna Units 1 and 2 RPV beltline and extended beltline materials, respectively. Appendix C contains a description of the North Anna licensing basis relative to selection of CFs when surveillance data is available.

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Westinghouse Non-Proprietary Class 3 3-4 VENT: PIPE (Mi< 23 AND 24)

CCNnict ROD ~HAN!SM-.- - ~

HOUSING I MK 21 ANO 22]

SHROLO SUPPORT - - - CLOS~E STIJO ( Mi( 271 LIFTING LUG ----A CLOSURE HEAD fll:f-......-..=c-;:;~~=---""i%---Sf'HERtCAL WASHERS (FLANGE! (MK 29° AfiO 29b1

- - - OUTER O*RINO J.--------'"----~..,;.;,,-.,r- GASKET IMK 251 VESSELFl.M'GE 0::::::::::::::::::::::::+:::::::::::~~~--lNNER O-RINO GASKET I MK 26)

OUTLET NOZZLE IMl< 12, l3 ANO 14] /

VESSEL Sllf'l"OOT _/

PAO INTERY.EOIATE SI-IELL COU!,SE I

Sl£LL COURSE CORE SUPPORr----,:,i.

OUlDE [MK16l BOTTOM HEAD - -

SPH!:RICAL RING INS'TRllMENTATION NOZZLES IMK 19 ANO 20) h"EAO CAP Figure 3-1 RPV Base Metal Material Identifications for North Anna Units 1 and 2

Note: This figure is representative of the RPV with the original RPV closure heads.

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Westinghouse Non-Proprietary Class 3 3-5 Best-Estimate Cu and Ni Weight Percent Values, Initial RTNnT Values, and Initial USE Values for the North Anna Unit 1 RPV Beltline and Extended Beltline MateriaJsCa)

Heat Flux Type Wt.% Wt.% RT NDT(U)(b) Initial USE Material Description Number (Lot) Cu Ni (OF) (ft-lbs}

Reactor Vessel Beltline Materials 990286 I Upper Shell Forging 05 295213

- 0.16 0.74 1 72 Upper to Intermediate Sl\1IT 89 Shell Circumferential 25295 0.352 0.125 -40 112 (1170)

Weld (94% OD)

Upper to Intermediate Sl\1IT 89 Shell Circumferential 4278 0.12 0.11 -4 105 (1211)

Weld (6%ID)

Intermediate Shell 990311 I Forging 04 298244 - 0.12 0.82 -6 91 Intermediate to Lower Sl\1IT 89 Shell Circumferential 25531 0.098 0.124 -2 95 (1211)

Weld 990400 I Lower Shell Forging 03 292332 - 0.156 0.817 33 85 Reactor Vessel Extended Beltline Materials Inlet/Outlet Nozzle to o.35(c) 1.13(c) 30(d) 72(c)

RotterdamCd)

Upper Shell Welds Inlet Nozzle Forging 09 990290-11 - 0.13 0.80 -14 2". 71 Inlet Nozzle Forging 10 990290-12 - 0.13 0.79 -10 2". 58 Inlet Nozzle Forging 11 990268-21 - 0.18 0.78 8 56(c)

Outlet Nozzle Forging 12 990290-31 - 0.13 0.80 -6 2". 66 Outlet Nozzle Forging 13 990290-22 - 0.13 0.81 -7 2". 59 Outlet Nozzle Forging 14 990290-21 - 0.13 0.81 8 2". 59 Reactor Vessel Surveillance Program Materials(,)

990400 I Lower Shell Forging 03 292332

- 0.158 0.823 - -

Intermediate to Lower Sl\1IT 89 Shell Circumferential 25531 (1211) 0.098 0.124 - -

Weld Notes: *1 (a) Unless otherwise noted, the information is extracted from PWROG-18005-NP (Reference 7). Dashes indicate when category a is not applicable to the material. *

(b) All RTNDT(U) values are based on measured data which are used in conjunction with ASME Code Section III (Reference 9) and/or BTP 5-3 (Reference 8) methods; thus, am value of0°F can be used with these RTNDT(U) values per WCAP-14040-A, Revision 4 (Reference 3). -,

(c) Generic value developed in PWROG-17090-NP-A (Reference 12). Justification for the use of these values, consistent with the NRC Safety Evaluation, are presented in Appendix D.

(d) The specific heat, flux type, and flux lot numbers of these welds could not be identified in the available information; therefore, conservative generic numbers will be used to describe these welds. The RTNDT(U) value was determined using ASME Code Section III minimum criteria at the time of fabrication and BTP 5-3 (Reference 8), Position 1.1(4) guidance. Since this is a maximum possible value based on measured data that satisfied the ASME requirements, them associated with this RTNDT(U) is zero.

(e) The reactor vessel surveillance material data is taken from Dominion Energy calculation SM- I 008 (Reference 25).

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Westinghouse Non-Proprietary Class 3 3-6 Table 3-2 Best-Estimate Cu and Ni Weight Percent Values, Initial RTNoT Values, and Initial USE Values for the North Anna Unit 2 RPV Beltline and Extended Beltline MaterialsCa)

Heat Flux Type Wt.% Wt.% RTNDT(U)(b) Initial USE Material Description (OF)

Number (Lot) Cu Ni (ft-lbs)

Reactor Vessel Beltline Materials 990598 I Upper Shell Forging 05 291396

- - 0.08 0.77 8 72 Upper to Intermediate SMIT 89 Shell Circumferential 4278 0.12 0.11 -4 105 (1211)

Weld (94% OD)

Upper to Intermediate SMIT 89 75(c)

Shell Circumferential 801 0.18 0.11 10 (1211)

Weld (6%ID)

Intermediate Shell 990496 I Forging 04 292424

- 0.107 0.857 69 72 Intermediate to Lower LW320 Shell Circumferential 716126 0.066 0.046 -67 109 (26)

Weld 990533 I Lower Shell Forging 03 - 0.13 0.83 37 80 297355 Reactor Vessel Extended Beltline Materials LW320 8816 (28)

LW320 20459 Inlet/Outlet Nozzle to (26) 0.23(c) 0.56(c) 30(d) 75(c)

Upper Shell Welds LW320 27622 (26)

LW320 27622 (28)

Inlet Nozzle Forging 09 990426 - 0.19 0.82 11 56(c)

Inlet Nozzle Forging 10 54567-2 - 0.14 0.79 5 ?. 77 Inlet Nozzle Forging 11 54590-2 - 0.155 0.77 -31 ?. 75 Outlet Nozzle Forging 12 990426-22 - 0.19 0.80 8 ?. 60 Outlet Nozzle Forging 13 990426-31 - 0.19 0.79 1 56(c)

Outlet Nozzle Forging 14 791291 - 0.12 0.82 -22 ?. 74

Reactor Vessel Surveillance Program Materials(e)

Intermediate Shell 990496 I Forging 04 292424

- 0.116 0.886 - *. -

Intermediate to Lower LW320 Shell Circumferential 716126 0.067 0.052 - -

(26)

Weld Notes contained on the following page.

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Westinghouse Non-Proprietary Class 3 3-7 Notes:

(a) Unless otherwise noted, the information is extracted from PWROG-18005-NP (Reference 7). Dashes indicate when a category is not applicable to the material.

(b) All RTNDT(U) values are based on measured data which are used in conjunction with ASME Code Section III (Reference 9) and/or BTP 5-3 (Reference 8) methods; thus, a 01 value of0°F can be used with these RTNDT(U) values per WCAP-14040-A, Revision 4 (Reference 3).

(c) Generic value developed in PWROG-17090-NP-A (Reference 12). Justification for the use of these values, consistent with the NRC Safety Evaluation, are presented in Appendix D.

(d) The records do not identify which weld heats are associated with which specific nozzle welds. Therefore, the bounding material properties will be conservatively associated with all Unit 2 nozzle welds. The RTNDT(U) value was determined using ASME Code Section III minimum criteria at the time of fabrication and BTP 5-3 (Reference 8), Position 1.1(4) guidance.

Since this is a maximum possible value based on measured data that satisfied the ASME requirements, the 01 associated with this RTNDT(U) is zero.

(e) The reactor vessel surveillance material data is taken from Dominion Energy calculation SM- I 008 (Reference 25).

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Westinghouse Non-Proprietary Class 3 3-8 Table3-3 Initial RTNDT Values for the North Anna Units 1 and 2 Replacement Reactor Vessel Closure Head and Vessel Flange Materials<a)

Unit 1 Unit2 Reactor Vessel Material Initial RTNDT Initial RTNDT (OF) {°F)

Replacement Closure Head -76 -49 Vessel Flange -22 -22 Note:

(a) The infonnation is extracted from PWROG-18005-NP (Reference 7).

Table3-4 Calculation of Position 2.1 CF Values for North Anna Unit 1 Surveillance Materials Capsule Fluence<a> ARTNDT(c) FF*ARTNDT Material Capsule FF(b) FF2 (x 10 19 n/cm2, {°F) {°F)

E>l.OMeV)

Lower Shell V 0.306 0.675 51 34.44 0.456 Forging 03 u 0.914 0.975 116 113.08 0.950 (Tangential) w 2.05 1.196 93 111.19 1.429 Lower Shell V 0.306 0.675 29 19.59 0.456 Forging 03 u 0.914 0.975 72 70.19 0.950 (Axial) w 2.05 1.196 96 114.77 1.429 SUM: 463.25 5.671 CFSurveillance Forging :E(FF

L\RTNDT) + k (FF)2 (463.25) + (5,671) = 81.68°F Surveillance V 0.306 0.675 88 59.43 0.456 Weld Metal u 0.914 0.975 30 29.24 0.950 (Heat# 25531) w 2.05 1.196 86 ,102.82 1.429

\ SUM: 191.50 2.836 CFsuiveillanceWeld= :E (FF* L\RTNDT) + :E (FF) 2 (191.50) + (2.836) = 67.53°F Notes:

(a) The fluence values are taken from Table 2-1 ofthls report.

(b) FF fluence factor= f(O.zs-o.io*iog(!))_

( c) ARTNDT values are extracted from BAW-2356 (Reference 13 ). Chemistry adjustments are not performed because the beltline and surveillance materials are identical and/or not adjusting for chemical composition is conservative.

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Westinghouse Non-Proprietary Class 3 3-9 Table 3-5 Calculation of Position 2.1 CF Values for North Anna Units 1 and 2 Welds with Data from Other Plant Surveillance Programs Capsule Fluence<a) ARTND/a)(c) FF*L\RTNDT FF2 Material Capsule FF(b)

(x 1019 n/cm2, (OF) (OF)

E> 1.0 MeV) 123.79 T 0.241 0.615 76.11 0.378 (127.79) 140.92 u 0.693 0.897 (144.92) 126.43 0.805 Sequoyah 1 155.02 X 1.16 1.041 161.44 1.085 Surveillance (159.02)

Weld Material y 159.80 (Heat # 25295) 1.97 1.185 189.39 1.405 (163.80)

SUM: 553.37 3.672 CFsurveillance Weld= I:(FF

ilRTNDr) + L (FF) 2 = (553.37) + (3.672) = 150.69°F 70.56 T 0.244 0.618 43.60 0.382 (74.56) 126.38 u 0.654 0.881 (130.38) 111.34 0.776 Sequoyah 2 40.22 X 1.16 1.041 41.89 1.085 Surveillance (44.22)

Weld Material y 82.91 (Heat# 4278)<d) 2.02 1.192 98.81 1.420 (86.91)

SUM: 295.63 3.663 CFsurveillanceWeld= L (FF* ilRTNor) + L (FF)2 = (295.63) + (3.663) = 80.71°F(d)

Notes:

(a) Sequoyah Units 1 and 2 surveillance data are taken from WCAP-17539-NP (Reference 17).

(b) FF= fluence factor= f(O.zs-o.io*Iog (f))_

(c) The surveillance weld ~RTNDr values have been decreased by 4°F (547°F - 551 °F) to account for the difference in the operating temperature between the Sequoyah and North Anna units. Pre-adjusted values are listed in parentheses. Chemistry adjustments are not performed since the Sequoyah Units 1 and 2 surveillance weld CF values are 178.7°F and 67.9°F, respectively. This results in Position 1.1 CF ratios less than l; therefore, not adjusting for chemical composition is conservative.

(d) Since North Anna Units 1 and 2 have same vessel weld CF and inlet temperature, the results apply to both units.

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Westinghouse Non-Proprietary Class 3 3-10 Table 3-6 Calculation of Position 2.1 CF Values for North Anna Unit 2 Surveillance Materials Capsule Fluence<a> FF(b) ARTNDic) FF*ARTNDT Material Capsule FF2 (x 10 19 n/cm2, E (OF) {°F)

> 1.0MeV)

Intermediate V 0.286 0.658 19 12.50 0.433 Shell Forging 04 u 0.985 0.996 33 32.86 0.992 (Tangential) w 2.08 1.199 86 103.14 1.438 V 0.286 0.658 21 13.82 0.433 Intermediate Shell Forging 04 u 0.985 0.996 66 65.72 0.992 (Axial) w 2.08 1.199 65 77.96 1.438 SUM: 306.00 5.726 CFsurveillance Forging :E(FF

LiRTNDT) + :E (FF)2 = (306.00) (5.726) = 53.44°F Surveillance V 0.286 0.658 18 11.84 0.433 Weld Metal u 0.985 0.996 8 7.97 0.992 (Heat# 716126) w 2.08 47 56.37 1.438 1.199 SUM: 76.18 2.863 CFsurveillance Weld= :E (FF* LiRTNDT) :E (FF)2 = (76.18) + (2.863) = 26.61°F Notes:

(a) The fluence values are taken from Table 2-2 of this report.

(b) FF= fluence factor fC0*28 -o.1o*iog(t)J_

(c) liRTl\'DT values are extracted from BAW-2376 (Reference 14). Chemistry adjustments are not performed because the beltline and surveillance materials are identical aud/or not adjusting for chemical composition is conservative.

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Westinghouse Non-Proprietary Class 3 3-11 Table 3-7 Summary of the North Anna Unit 1 RPV Beltline, Extended Beltline, and Surveillance Material CF Values based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 Chemistry Factor Heat Flux Type Material Position 1.1 (a) Position 2.1 (h)

Number (Lot)

(OF) (OF)

Reactor Vessel Beltline Materials 990286 I Upper Shell Forging 05 295213 - 121.50 -

Upper to Intermediate Shell SMIT 89 25295 163.25 150.69 Circumferential Weld (OD 94%) (1170)

Upper to Intermediate Shell SMIT89 4278 63.00 80.71 Circumferential Weld (ID 6%) (1211) 990311 I Intermediate Shell Forging 04 298244

- 86.oo* -

Intermediate to Lower Shell SMIT89 25531 56.22 67.53 Circumferential Weld (1211) 990400 I Lower Shell Forging 03 - 119.97 81.68 292332 Reactor Vessel Extended Beltline Materials Inlet/Outlet Nozzle to Upper Shell Welds Rotterdam - 293.45 -

Inlet Nozzle Forging 09. 990290-11 - 96.00 -

Inlet Nozzle Forging 10 990290-12 - 95.75 -

Inlet Nozzle Forging 11 990268-21 - 140.30 -

Outlet Nozzle Forging 12 990290-31 - 96.00 -

Outlet Nozzle Forging 13 990290-22 - 96.00 -

Outlet Nozzle Forging 14 990290-21 - 96.00 -

Reactor Vessel Surveillance Program Materials 990400 I Lower Shell Forging 03 - 121.63 -

292332

'.Intermediate to Lower Shell SMIT89 Circumferential Weld 25531 (1211) 56.22 -

Notes:

(a) All values are based on Tables I and 2 ofRegulatory Guide L99, Revision 2 (Position I.I) using the Cu and Ni weight percent values given in Table 3-1 of this report. Dashes indicate when a category is not applicable to the material.

(b) Values are from Tables 3-4 and 3-5 of this report.

WCAP-18364-NP March 2020 Revision 1

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Westinghouse Non-Proprietary Class 3 3-12 Table 3-8 Summary of the North Anna Unit 2 RPV Beltline, Extended Beltline, and Surveillance Material CF Values based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 Chemistry Factor Flux Type Material Heat Number Position 1.1 Ca) Position 2.1 (b)

(Lot)

(OF) (OF)

Reactor Vessel Beltline Materials 990598 I Upper Shell Forging 05 291396

- 51.00 -

Upper to Intermediate Shell S:rv1IT89 4278 63.00 80.71 Circumferential Weld (OD 94%) (1211)

Upper to Intermediate Shell SWT 89 Circumferential Weld (ID 6%)

801 (1211) 87.80 -

990496 /

Intermediate Shell Forging 04 292424 - 74.00 53.44 Intermediate to Lower Shell LW320 716126 36.09 26.61 Circumferential Weld (26) 990533 /

Lower Shell Forging 03 297355

- 96.00 -

Reactor Vessel Extended Beltline Materials 8816 Inlet/Outlet Nozzle to LW320 20459 163.20 -

Upper Shell Welcls (26 & 28) 27622 Inlet Nozzle Forging 09 990426 - 150.40 -

Inlet Nozzle Forging 10 54567-2 - 104.75 -

Inlet Nozzle Forging 11 54590-2 - 118.25 -

Outlet Nozzle Forging 12 990426-22 - 150.00 -

Outlet Nozzle Forging 13 990426-31 - 149.60 -

Outlet Nozzle Forging 14 791291 - 86.00 -

Reactor Vessel Surveillance Program Materials I

990496 /

Intermediate Shell Forging 04 292424 82.40 -

Intermediate to Lower Shell LW320 Circumferential Weld 716126 (26) 37.08 -

Notes:

(a) All values are based on Tables 1 and 2 ofRegulatory Guide 1.99, Revision 2 (Position 1.1) using the Cu and Ni weight percent values given in Table 3-2 of this report. Dashes indicate when a category is not applicable to the material.

(b) Value;s are from Tables 3-5 and 3-6 of this report.

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Westinghouse Non-Proprietary Class 3 4-1 4 PRESSURIZED THERMAL SHOCK A limiting condition on RPV integrity known as 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 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 [Reference 15]) 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 oflicense, 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 (Reference 11 ).

These accepted methods were used with the surface fluence values of Section 2 to calculate the following RTPTS values for the North Anna Units 1 and 2 RPV materials. The SLR RTPTS calculations are presented in Tables 4-1 and 4-2 for North Anna Units 1 and 2, respectively.

PTS Conclusion All of the beltline and extended beltline materials in the North Anna Units 1 and 2 reactor vessel are below the RTPTs screening criteria values of 270°F for base metal and/or longitudinal welds, and 300°F for circumferentially oriented welds through SLR (72 EFPY). These RTPTS values are based on the revised initial RTNDT values in PWROG-18005-NP (Reference 7), which are developed using ASME Section III (Reference 9) and, if needed, NUREG-0800, BTP 5-3 (Reference 8) methodologies. Limiting fluence values corresponding to the lowest extent of the nozzle welds were used to calculate the RTPTS values for both the nozzle welds and nozzle forgings.

The North Anna Units 1 and 2 limiting RTPTs value for base metal and longitudinal welds at 72 EFPY is 212.2°F (see Table 4-1 and Table 4-2), which corresponds to North Anna Unit 2 Lower Shell Forging 03 based on Regulatory Guide 1.99, Position 1.1. Note that the RTPTs value calculated for Unit 1 Lower Shell Forging 03 per Regulatory Guide 1.99, Revision 2, Position 1.1 is higher. However, the use of the lesser of the Regulatory Guide 1.99, Revision 2, Position 1.1 (i.e., without surveillance data) and 2.1 (i.e., with surveillance data) CFs with non-credible data and a full margin term is justified* since none of the surveillance data are more than two times sigma-delta above the Position 1.1 CF trend line. This determination is documented in Appendix C. The North Anna Units 1 and 2 limiting RTPTS value for circumferentially oriented welds at 72 EFPY is 136.3°F (see Table 4-1 and Table 4-2), which corresponds to the North Anna Unit 1 Intermediate to Lower Shell Circumferential Weld Heat # 25531 based on Regulatory Guide 1.99, Position 1.1. Note that the RTPTs value for this material using the Position 2.1 CF is higher. However, per the discussion in Appendix C, the lower of the Position 1.1 and Position 2.1 CF can be used for this material. These limiting materials are consistent with the previous EOLE analysis, SM-1008 (Reference 25).

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Westinghouse Non-Proprietary Class 3 4-2 The Alternate PTS Rule (10 CFR 50.61a [Reference 16]) was published in the Federal Register by the NRC in 2010. This alternate rule is less restrictive than the PTS Rule (10 CFR 50.61) and is intended to be used for situations in which the 10 CFR 50.61 criteria cannot be met. North Anna Units 1 and 2 meet the criteria for the PTS Rule through SLR and therefore do not utilize the Alternate PTS Rule at this time.

WCAP-18364-NP March 2020 Revision 1

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Westinghouse Non-Proprietary Class 3 4-3 Table 4-1 Calculation of North Anna Unit 1 RTPTS Values for 72 EFPY (SLR) at the Clad/Base Metal Interface() a Surface Heat Flux Type CF(b) Fluence(c) Surf. RTNDT(U)(e) Predicted au<t) O'A(g) M RTPTS Material 19 FF(d) ARTNDT (OF)

Number (Lot) (x 10 n/cm2, {°F) {°F) {°F) {°F)

{°F)

E> 1.0MeV)

Reactor Vessel Beltline Materials 990286 I Upper Shell Forging 05 295213

- 121.50 0.304 0.674 1 81.9 0.0 17.0 34.0 116.9 Upper to Intermediate S:MIT 89 Shell Circumferential 25295 163.25 0.351 0.711 -40 116.1 0.0 28.0 56.0 132.1 Weld (OD 94%)(k) (1170)

Using credible surveillance data(ll) 150.69 0.351 0.711 -40 107.2 0.0 14.0 28.0 95.2 Upper to Intermediate S:MIT 89 Shell Circumferential 4278 63.00 0.351 0.711 -4 44.8 0.0 22.4 44.8 85.6 (1211)

Weld (ID 6%)

Using non-credible surveillance data(iJ 80.71 0.351 0.711 -4 57.4 0.0 28.0 56.0 109.4 Intermediate Shell 990311 /

Forging 04 298244 - 86.00 7.07 l.464 -6 125.9 0.0 17.0 34.0 153.9 Intermediate to Lower SMIT89 Shell Circumferential 25531 56.22 7.04 1.464 -2 82.3 0.0 28.0 56.0 136.3 (1211)

Weld Using non-credible surveillance dataaJ 67.53 7.04 1.464 -2 98.9 0.0 28.0 56.0 152.9 990400 I Lower Shell Forging 03 292332

- 119.97 7.20 1.467 33 176.0 0.0 17.0 34.0 243.0 Using non-credible surveillance data@ 81.68 7.20 1.467 33 119.9 0.0 17.0 34.0 186.9 WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-4 Table 4-1 Calculation of North Anna Unit 1 RTPTs Values for 72 EFPY (SLR) at the Clad/Base Metal Interface(a)

Surface Predicted Heat Flux Type CF(b) Fluence<c) Surf. RT NDT(U}e) o-u<fJ 0',1 (g) M RTPTS Material 19 FF(d) AR'fNoT (OF)

Number (Lot) (x 10 n/cm2, (OF) (OF) (OF) (OF)

(OF)

E>l.OMeV)

Reactor Vessel Extended Beltline Materials Inlet Nozzle Forging 09 0.00898 0 30 0.0 0.0 0.0 0.0 30.0 to Upper Shell Weld Inlet Nozzle Forging 10 0.0313 0.225 30 66.1 0.0 28.0 56.0 152.1 to Upper Shell Weld Inlet Nozzle Forging 11 0.0120 0.124 30 36.4 0.0 18.2 36.4 102.7 to Upper Shell Weld Rotterdam - 293.45 Outlet Nozzle Forging 12 0.0182 0.162 30 47.6 0.0 23.8 47.6 125.2 to Upper Shell Weld Outlet Nozzle Forging 13 0.00522 0 30 0.0 0.0 0.0 0.0 30.0 to Upper Shell Weld Outlet Nozzle Forging 14 0.00697 0 30 0.0 0.0 0.0 0.0 30.0 to Upper Shell Weld Inlet Nozzle Forging 09 990290-11 - 96.00 0.00898 0 -14 0.0 0.0 0.0 0.0 -14.0 InletNozzleForging 10 990290-12 - 95.75 0.0313 0.225 -10 21.6 0.0 10.8 21.6 33.1 Inlet Nozzle Forging 11 990268-21 - 140.30 0.0120 0.124 8 17.4 0.0 8.7 17.4 42.8 Outlet Nozzle Forging 12 990290-31 - 96.00 0.0182 0.162 -6 15.6 0.0 7.8 15.6 25.1 Outlet Nozzle Forging 13 990290-22 - 96.00 0.00522 0 -7 0.0 0.0 0.0 0.0 -7.0 Outlet Nozzle Forging 14 990290-21 - 96.00 0.00697 0 8 0.0 0.0 0.0 0.0 8.0 Notes contained on the following page.

WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-5 Notes:

(a) The IO CFR 50.61 (Reference 15) methodology was utilized in the calculation of the RTPTs values.

(b) Chemistry factors are taken from Table 3-7 of this report.

(c) Fluence is taken from Table 2-3 of this report. The surface fluence at the lowest extent of the nozzle to upper shell weld centerline was used to represent the inlet and outlet nozzle forgings and associated welds. *

(d) FF= fluence factor= ro.zs-o.1o*Iog (I))_ Embrittlement effects are considered only if the fluence is greater than 10 17 n/cm2* For materials with fluence less than 10 17 n/cm2 the FF is set equal to O.

(e) RTNDT(U) values are taken from Table 3-1 of this report.

(f) The initial RTNDT values are based on measured values; therefore cm= 0°F.

(g) Per IO CFR 50.61 (Reference 15), the base metal cr8 = l 7°F for Position 1.1 and Position 2.1 with non-credible surveillance data. Also, per 10 CFR 50.61, the weld metal cr8 =

28°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and cr8 = 14°F for Position 2.1 with credible surveillance data. However, cr8 need not exceed 0.5*8RTNDT.

(h) The surveillance data for weld Heat# 25295 from the Sequoyah Unit 1 surveillance program were deemed credible per WCAP-17539-NP (Reference 17).

(i) The surveillance data for weld Heat# 4278 from the Sequoyah Unit 2 surveillance program were deemed non-credible per WCAP-17539-NP (Reference 17).

G) The credibility evaluation for the North Anna Unit 1 surveillance data in Appendix A.1 of this report determined that the Lower Shell Forging 03 and weld Heat # 25531 surveillance data are deemed non-credible.

(k) While 10 CFR 50.61 specifically requires the analysis be performed at the surface and this material is not present at the surface, it is still included as it represents the majority of the weld.

WCAP-18364-NP March 2020 Revision 1

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Westinghouse Non-Proprietary Class 3 4-6 Table 4-2 Calculation of North Anna Unit 2 RT PTS Values for 72 EFPY (SLR) at the Clad/Base Metal Interface<al Surface Flux Predicted Heat CF(b) Fluence<cl Surf. RTNDT(U)(c) o-u<t) 0'11(g) M RTPTs Material Type 19 FF(d) (OF) ARTNDT (OF)

Number (x 10 n/cm2, {°F) {°F) (OF)

(Lot) {°F)

E>l.OMeV)

Reactor Vessel Beltline Materials 990598 I Upper Shell Forging 05 - 51.00 0.307 0.676 8 34.5 0.0 17.0 34.0 76.5 291396 Upper to Intermediate Shell SMIT 89 Circumferential Weld 4278 63.00 0.355 0.714 -4 45.0 0.0 22.5 45.0 86.0 (1211)

(OD 94%)Gl Using non-credible surveillance data(!,) 80.71 0.355 0.714 -4 57.6 0.0 28.0 56.0 109.6 Upper to Intermediate Shell SMIT 89 Circumferential Weld 801 87.80 0.355 0.714 10 62.7 0.0 28.0 56.0 128.7 (1211)

(ID 6%)

Intermediate Shell 990496 I Forging 04 292424 - 74.00 7.20 1.467 69 108.6 0.0 17.0 34.0 211.6 Using non-credible surveillance data(iJ 53.44 7.20 1.467 69 78.4 0.0 17.0 34.0 181.4 Intermediate to Lower LW320 716126 36.09 7.18 1.467 -67 52.9 0.0 26.5 52.9 38.9 Shell Circumferential Weld (26)

Using credible surveillance data(iJ 26.61 7.18 1.467 -67 39.0 0.0 14.0 28.0 0.0 990533 I Lower Shell Forging 03 - 96.00 7.34 1.470 37 141.2 0.0 17.0 34.0 212.2 297355 WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-7 Table 4-2 Calculation of North Anna Unit 2 RTPTs Values for 72 EFPY /SLR) at the Clad/Base Metal Interface(a)

Surface Flux Predicted Heat CF(b) Fluence(c) Surf. RT NDT(Ule) O'u(I) O'A(g) M RTPTS Material Type FF(d) ARTNDT Number (x 1019 n/cm2, (OF) {°F) {°F) {°F) {°F)

(Lot) {°F)

E>l.OMeV)

Reactor Vessel Extended Beltline Materials Inlet Nozzle Forging 09 to 30(k) 163.20 0.00826 0 0.0 0.0 0.0 0.0 30.0 Upper Shell Weld Inlet Nozzle Forging 10 to 30(k) 163.20 0.0314 0.226 36.8 0.0 18.4 36.8 103.6 Upper Shell Weld Inlet Nozzle Forging 11 to 30(k) 8816 163.20 0.0118 0.123 20.0 0.0 10.0 20.0 70.0 Upper Shell Weld LW320 20459 Outlet Nozzle Forging 12 (26 & 28) 27622 163.20 0.0182 0.162 30(k) 26.5 0.0 13.2 26.5 82.9 to Upper Shell Weld Outlet Nozzle Forging 13 30(k) 163.20 0.00479 0 0.0 0.0 0.0 0.0 30.0 to Upper Shell Weld Outlet Nozzle Forging 14 30(k) 163.20 0.00687 0 0.0 0.0 0.0 0.0 30.0 to Upper Shell Weld Inlet Nozzle Forging 09 990426 - 150.40 0.00826 0 11 0.0 0.0 0.0 0.0 11.0 Inlet Nozzle Forging 10 54567-2 - 104.75 0.0314 0.226 5 23.6 0.0 11.8 23.6 52.3 Inlet Nozzle Forging 11 54590-2 - 118.25 0.0118 0.123 -31 14.5 0.0 7.2 14.5 -2.0 Outlet Nozzle Forging 12 990426-22 - 150.00 0.0182 0.162 8 24.3 0.0 12.2 24.3 56.7 Outlet Nozzle Forging 13 990426-31 - 149.60 0.00479 0 1 0.0 0.0 0.0 0.0 1.0 Outlet Nozzle Forging 14 791291 - 86.00 0.00687 0 -22 0.0 0.0 0.0 0.0 -22.0 Notes contained on the following pag~.

WCAP-18364-NP March2020 Revision 1

""* This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 4-8 Notes:

(a) The 10 CFR 50.61 (Reference 15) methodology was utilized in the calculation of the RTPTs values.

(b) Chemistry factors are taken from Table 3-8 of this report.

(c) Fluence is taken from Table 2-4 of this report. The surface fluence at the lowest extent of the nozzle to upper shell weld centerline was used to represent the inlet and outlet nozzle forgings and associated welds.

(d) FF= fluence factor= f( 0-28 -o.1o*iog(fJ)_ Embrittlement effects are considered only if the fluence is greater than 10 17 n/cm2

For materials with fluence less than 10 17 n/cm2 the FF is set equal to O. *

(e) RTNDT(U) values are taken from Table 3-2 of this report.

(t) The initial RTNDT values are based on measured values; therefore cru = 0°F.

(g) Per 10 CFR 50.61 (Reference 15), the base metal cr~ = l 7°F for Position 1.1 and Position 2.1 with non-credible surveillance data. Also, per IO CFR 50.61, the weld metal cr~ =

28°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and cr~ = 14°F for Position 2.1 with credible surveillance data. However, cr~ need not exceed 0.5* LIB.TNDT-(h) The surveillance data for weld Heat# 4278 from the Sequoyah Unit 2 surveillance program were deemed non-credible per WCAP-17539-NP (Reference 17).

(i) The credibility evaluation for the North Anna Unit 2 surveillance data in Appendix A.2 of this report determined that the Intermediate Shell Forging 04 surveillance data are deemed non-credible; however, the weld Heat# 716126 surveillance data are deemed credible.

G) While IO CFR 50.61 specifically requires the analysis be performed at the surface and this material is not present at the surface, it is still included as it represents the majority of the weld.

(k) The RTNDT(U) is based on the highest RTNDT(U) of the heats associated with this weld.

WCAP-18364-NP March 2020 Revision 1

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Westinghouse Non-Proprietary Class 3 5-1 5 UPPER-SHELF ENE,RGY The decrease in Charpy upper-shelf energy (USE) is associated with the determination of acceptable RPV toughness during the license renewal period when the vessel is exposed to additional irradiation.

The requirements on USE are included in 10 CFR 50, Appendix G (Reference 10). 10 CFR 50, Appendix G requires utilities to submit an analysis at least 3 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 (Reference 11). For vessel beltline 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 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. Per Regulatory Guide 1.99, Revision 2, when credible data exist, the Position 2.2 projected USE value should be used in preference to the Position 1.2 projected USE value. Note, if data from the surveillance materials is determined to be non-credible for determination of 11RTNDT by Credibility Criterion 3 of Regulatory Guide 1.99, Revision 2, then "they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given inASTM E 185-82."

The 72 EFPY Position 1.2 USE values of the vessel materials can be predicted using the corresponding 1/4T fluence projections, the copper content of the materials, and Figure 2 in Regulatory Guide 1.99, Revision 2 (see Figures 5-1 and 5-3 of this report).

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 reduced plant surveillance data was obtained from Table 7-6 of BAW-2356 (Reference 13) for North Anna Unit 1. The reduced plant surveillance data was obtained from Table 7-6 of BAW-2376 (Reference 14) for North Anna Unit 2. The surveillance data was plotted in Regulatory Guide 1.99, Revision 2, Figure 2 (see Figures 5-2 and 5-4 of this report) using the surveillance capsule fluence values documented in Table 2-1 of this report for North Anna Unit 1 and Table 2-2 of this report for North Anna Unit 2. This data was fitted by drawing a line parallel to the existing lines as the upper bound of all the surveillance data. These reduced lines were used instead of the existing lines to determine the Position 2.2 SLR USE values.

The projected USE values were calculated to determine if the North Anna Units 1 and 2 beltline and extended beltline materials remain above the 50 ft-lb criterion at 72 EFPY (SLR). These calculations are summarized in Tables 5-1 and 5-2. Fluence values corresponding to the lowest extent of the nozzle welds at the surface were used to conservatively calculate the projected USE values for the nozzle forgings.

WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-2 USE Conclusion For North Anna Unit 1, the limiting USE value at 72 EFPY is 50.0 ft-lb (see Table 5-1); this value corresponds to Inlet Nozzle Forging 11 using Position 1.2. The Unit 1 Inlet Nozzle 11 USE value set equal to 50 ft-lbs results in a projected drop of 10.7%. A review of Regulatory Guide 1.99, Revision 2, Figure 2 resulted in a conservative estimate of approximately 11 %, but the figure has limited precision. A decrease of 10.7% is considered appropriate based on the following conservativism in the calculations. The estimated% decrease is based on a fluence of 2 x 10 17 n/cm 2 (E > 1.0 MeV), which is the lowest fluence line displayed in Regulatory Guide 1.99, Revision 2, Figure 2. The actual fluence is projected to be roughly half this, i.e. 1.20 x 10 17 n/cm2 (E > 1.0 MeV), at the lowest extent of the nozzle weld and would be even lower at higher axial elevations. In addition, the fluence would be further decreased if attenuation to the 1/4T location were considered. These additional decreases influence would raise the projected USE of Unit 1 Inlet Nozzle 11 above 50 ft-lbs. As shown in Table 5-1, all North Anna Unit 1 reactor vessel materials are projected to remain at or above the USE screening criterion value of 50 ft-lbs at 72 EFPY.

For North Anna Unit 2, the limiting USE value at 72 EFPY is 48.2 ft-lb (see Table 5-2); this value corresponds to the Intermediate Shell Forging 04 using Position 2.2. Position 2.2 was used to determine the Unit 2 Intermediate Shell Forging 04 USE value even though its surveillance data was deemed non-credible per Appendix A. Per Regulatory Guide 1.99, Revision 2, this is appropriate since the upper shelf can be clearly determined from the surveillance test results. As shown in Table 5-2, all other North Anna Unit 2 reactor vessel materials are projected to remain above the USE screening criterion value of 50 ft-lbs at72 EFPY.

The North Anna Unit 2 Intermediate Shell Forging 04 reactor vessel material, which is projected to drop below 50 ft-lbs USE at SLR, is addressed in the equivalent margins analysis (EMA) being performed under PWROG PA-MSC-1481 to qualify the material at 72 EFPY. The material-specific EMA in PA-MSC-1481 is underway, and must be submitted at least 3 years prior to the USE dropping below 50 ft-lbs. The Unit 2 Intermediate Shell Forging 04 is projected to drop below 50 ft-lbs at 52.3 EFPY (EOLE), which is projected to occur in 2040.

In addition to the material discussed above, PA-MSC-1481 includes EMAs for all of the following materials at each Unit for conservativism.

Upper Shell Forging

Intermediate Shell Forging

Inlet Nozzle Forgings

Outlet Nozzle Forgings

Inlet Nozzle Welds

Outlet Nozzle Welds WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-3 Table 5-1 Predicted USE Values at 72 EFPY (SLR) for North Anna Unit 1 Surf. SLR 1/4TSLR Projected Unirradiated Projected Reactor Vessel Wt.% Fluence<hl Fluence<c) USE Heat# USE<*l SLR USE Material cu<*) (x 10 19 n/cm2, (x 10 19 n/cm2, Decrease

l.OMeV) E>l.OMeV) (%) Position 1.2 990286 I Upper Shell Forging 05 0.16 0.304 0.192 72 17.0 59.8 295213 Upper to Intermediate Shell Circumferential 25295 0.352 0.351 0.221 112 34.0 73.9 Weld (OD 94%) Upper to Intermediate Shell Circumferential 4278 0.12 0.351 0.221 105 18.5 85.6 Weld (ID 6%)<gl Intermediate Shell 990311 / 0.12 7.07 4.46 91 30.0 63.7 Forging 04 298244 Intermediate to Lower Shell Circumferential 25531 0.098 7.04 4.44 95 34.5 62.2 Weld 990400 I Lower Shell Forging 03 0.156 7.20 4.54 85 36.0 54.4 292332 Inlet Nozzle Forging 09 o.o<f) 0.00898 0.00898 72 72.0 to Unner Shell Weld Inlet Nozzle Forging 10 0.0313 0.0313 72 26.0 53.3 to Upper Shell Weld Inlet Nozzle Forging 11 0.0120 0.0120 72 24.0 54.7 to Unner Shell Weld Rotterdam 0.35 Outlet Nozzle Forging 0.0182 0.0182 72 24.0 54.7 12 to Upper Shell Weld Outlet Nozzle Forging 0.o<f) 0.00522 0.00522 72 72.0 13 to Upper Shell Weld Outlet Nozzle Forging 0.o<f) 0.00697 0.00697 72 72.0 14 to Unner Shell Weld Inlet Nozzle Forging 09 990290-11 0.13 0.00898 0.00898 71 0.o<f) 71.0 Inlet Nozzle Forging 10 990290-12 0.13 0.0313 0.0313 58 10.0 52.2 Inlet Nozzle Forging 11 990268-21 0.18 0.0120 0.0120 56 10.7(h) 50.0(h) Outlet Nozzle 990290-31 0.13 0.0182 0.0182 66 9.0 60.1 Forging 12 Outlet Nozzle o.o<f) 990290-22 0.13 0.00522 0.00522 59 59.0 Forging 13 Outlet Nozzle 0.o<f) 990290-21 0.13 0.00697 0.00697 59 59.0 Forging 14 Position 2.2< 0> Intermediate to Lower Shell Circumferential 25531 0.098 7.04 4.44 95 27.0 69.4(i) Weld 990400 I 54.4(i) Lower Shell Forging 03 0.156 7.20 4.54 85 36.0 292332 Notes on the following page. WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-4 Notes: (a) Copper weight percent values and unirradiated USE values were taken from Table 3-1 of this report. (b) Surface fluence values taken from Table 2-3. (c) The l/4T fluence values were calculated from the surface fluence, the reactor vessel beltline thickness (7.677 inches) and equation f = fsurr

e-0-24 (x) from Regulatory Guide 1.99, Revision 2, where x = the depth into the vessel wall (inches). The surface fluence at the lowest extent of the nozzle to upper shell weld centerline was used to represent the inlet and outlet nozzle forgings and associated welds. Fluence values above 10 17 n/cm2 (E > 1.0 MeV) but below 2 x 10 17 n/cm2 (E > 1.0 MeV) were rounded to 2 x 10 17 n/cm2 (E > 1.0 MeV) when determining the % decrease because 2 x 10 17 n/cm2 is the lowest fluence displayed in Figure 2 of the Guide.

( d) The Position 1.2 USE decrease values were calculated by plotting the 1/4 T fluence values onto Figure 2 of Regulatory Guide 1.99, Revision 2 and using the material-specific Cu wt. % values. Base metal and weld Cu wt. % 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. ( e) Calculated using surveillance capsule measured percent decrease in USE from BA W-2356 (Reference 13) and Regulatory Guide 1.99, Revision 2, Position 2.2; see Figure 5-2. (f) Embrittlement effects only need to be considered if the fluence is greater than 1017 n/cm2 (E > 1.0 MeV). (g) Since this inner diameter (ID) weld is only 6% of the vessel thickness, the weld is not present at the 1/4T location; hence, it is not applicable to this calculation. It is presented for information only. (h) The Unit 1 Inlet Nozzle 11 USE value is set equal to 50 ft-lbs which results in a projected drop of 10.7%. A review of Regulatory Guide 1.99, Revision 2, Figure 2 resulted in a conservative estimate of approximately 11 %, but the figure has limited precision. A decrease of 10. 7% is considered appropriate based on the following conservativism in the calculations. The estimated% decrease is based on a fluence of2 x 10 17 n/cm2 (E > 1.0 MeV), which is the lowest fluence line displayed in Regulatory Guide 1.99, Revision 2, Figure 2. The actual fluence is projected to be roughly half this, i.e. 1.20 x 10 17 n/cm2 (E > 1.0 MeV), at the lowest extent of the nozzle weld and would be even lower at higher axial elevations. In addition, the fluence would be further decreased if attenuation to the 1/4T location were considered. These additional decreases influence would raise the projected USE of Unit 1 Inlet Nozzle 11 above 50 ft-lbs. (i) Position 2.2 was used to determine the Unit 1 Lower Shell Forging 03 and the Intermediate to Lower Shell Weld USE value even though the surveillance data were deemed non-credible per Appendix A. Per Regulatory Guide 1.99, Revision 2, this is appropriate since the upper shelf can be clearly determined from the surveillance test results. WCAP-18364-NP March 2020 Revision 1

- - This record was final approved.on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-5 Table 5-2 Predicted USE Values at 72 EFPY (SLR) for North Anna Unit 2 Surf. SLR l/4TSLR Projected Unirradiated Projected Wt.% Fluence<bl Fluence(c) USE Reactor Vessel Material Heat# USE<a> SLR USE Cu(a) (x 10 19 n/cm2, (x 10 19 n/cm2 , Decrease(dl (ft-lbs) (ft-lbs) E>l.OMeV) E>l.OMeV) (%) Position 1.2 990598 I Upper Shell Forging 05 0.08 0.307 0.194 72 13.0 62.6 291396 Upper to Intermediate Shell Circumferential Weld 4278 0.12 0.355 0.224 105 18.5 85.6 (OD94%) Upper to Intermediate Shell Circumferential Weld 801 0.18 0.355 0.224 75 23.0 57.8 (ID 6%)(gl Intermediate Shell 990496 / 0.107 7.20 4.54 72 28.0 51.8 Forging 04 292424 Intermediate to Lower Shell 716126 0.066 7.18 4.53 109 29.0 77.4 Circumferential Weld 990533 / Lower Shell Forging 03 0.13 7.34 4.63 80 32.0 54.4 297355 Inlet Nozzle Forging 09 to 0.00826 0.00826 75 o.o<0 75.0 Uooer Shell Weld Inlet Nozzle Forging 10 to 0.0314 0.0314 75 17.0 62.3 Unner Shell Weld Inlet Nozzle Forging 11 to 8816 0.0118 0.0118 75 15.0 63.8 Uooer Shell Weld 20459 0.23 Outlet Nozzle Forging 12 to 27622 0.0182 0.0182 75 15.0 63.8 Uooer Shell Weld Outlet Nozzle Forging 13 to 0.00479 0.00479 75 o.o<o 75.0 Uooer Shell Weld Outlet Nozzle Forging 14 to 0.00687 0.00687 75 o.o<o 75.0 Unner Shell Weld Inlet Nozzle Forging 09 990426 0.19 0.00826 0.00826 56 O.O(O 56.0 Inlet Nozzle Forging 10 54567-2 0.14 0.0314 0.0314 77 10.5 68.9 Inlet Nozzle Forging 11 54590-2 0.155 0.0118 0.0118 75 10.0 67.5 Outlet Nozzle Forging 12 990426-22 0.19 0.0182 0.0182 60 11.5 53.1 Outlet Nozzle Forging 13 990426-31 0.19 0.00479 0.00479 56 0.0(I) 56.0 Outlet Nozzle Forging 14 791291 0.12 0.00687 0.00687 74 o.oco 74.0 \ Position 2.z(e) Intermediate Shell 990496 / 48.2(h) 0.107 7.20 4.54 72 33.0 Forging 04 292424 Intermediate to Lower Shell 716126 0.066 7.18 4.53 109 28.0 78.5 Circumferential Weld Notes on the following page. WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-6 Notes: (a) Copper weight percent values and unirradiated USE values were taken from Table 3-2 of this report. (b) Surface fluence values taken from Table 2-4. (c) The I/4T fluence values were calculated from the surface fluence, the reactor vessel beltline thickness (7.677 inches) and equation f= Tuurf

e-0-24 (x) from Regulatory Guide 1.99, Revision 2, where x = the depth into the vessel wall (inches). The surface fluence at the lowest extent of the nozzle to upper shell weld centerline was used to represent the inlet and outlet nozzle forgings and associated welds. 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 10 17 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 the Guide.

(d) The Position 1.2 USE decrease values were calculated by plotting the 1/4T fluence values onto Figure 2 ofRegulatory Guide 1.99, Revision 2 and using the material-specific Cu wt.% values. Base metal and weld Cu wt. % lines were extended into the low fluence area of Regulatory Guide 1.99, Revision 2, Figure 2, i.e., below 10 18 n/cm2 , in order to determine the USE% decrease as needed. (e) Calculated using surveillance capsule measured percent decrease in USE from BAW-2376 (Reference 14) and Regulatory Guide 1.99, Revision 2, Position 2.2; see Figure 5-4. (f) Embrittlement effects only need to be considered if the fluence is greater than 1017 n/cm2 (E > 1.0 MeV). (g) Since this inner diameter (ID) weld is only 6% of the vessel thickness, the weld is not present at the l/4T location; hence, it is not applicable to this calculation. It is presented for information only. (h) Position 2.2 was used to determine the Unit 2 Intermediate Shell Forging 04 USE value even though its surveillance data was deemed non-credible per Appendix A. Per Regulatory Guide 1.99, Revision 2, this is appropriate since the upper shelf can be clearly determined from the surveillance test results. WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-7 100.0 -- % Copper Base Metal Weld 0.35 0.30 n '.VI 0.25 0.25 n,,o I 0.20 Upper Limit t --- 0.15 0.15 n.10 0.10 0.05 \ ~ ~ ,,,,,,,.,,,,,,, --- ... ,,,,,,,.

........ -- . . ~ ... . -

~ ~

........ ~ i,.----

~

w u,

-- ....- ~~

- =---- ~

~ .,...i,.,

.5 a.

....- .... ~ ....- ....-

~

.... t:::::==- .... ---- -~,,,,,,,.--

e ~~

Q ~ Intermediate to Lowar Shell Girth Wekl

~ ~~ '

Ill 1:11

_.i,., 72 EFPY 1/4T Fluence

4."3 x 1019 n/cm2

! 10.0 &

C Intermediate SheH Forging Ill ~

72 EFPY 1/4T Fluence

4.45 x 10 19 n/c;m2 e

Ill Inlet Nozzle Forging 10 & I Q. Nozzle to Upper Shell Weld I

...- 72 EFPY Surface Fluence

3.13 x 10 11 ntc~

f--

J I i . - Shen Forging 72 EFPY 1/4T Fluence

4.53 x 10 11 nlcm2 I I I Inlet Nozzle Forging 11 &

- Nozzle to Upper Shell Weld 72 EFPY Surface Fluenoe

1.20 x 1017 n/cm2 I j 1- - I""!

Upper/Intermediate Girth Weld 72 EFPY 1/4T Fluence = 2.21 x 1011 nfcmZ OUtlel Nozzle Forging 12 &

Nozzle to Upper Shel Weld 72 EFPY Surf- Fluence

I I I III I I I 1.82 X 1017 n/cm2 I I 1 11

.~ I 1 I I Upper Shen Forging 72 EFPY 1/4T Fluence = 1.91 x 1011 ntcm2 I I I I I 1.0 1.DOE+17 1.00E+18 1.00E+19 1.00E+20 Neutron Fluence, n/cm 2 (E > 1 MeV)

Figure 5-1 Regulatory Guide 1.99, Revision 2, Position 1.2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for North Anna Unit 1 at SLR (72 EFPY)

WCAP-18364-NP March 2020 Revision I

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-8

% Copper

~m:Mfllll W.m1 0.35 0.30 n-,n I n '>r::

100.0 n,r:: I I n -,n 0.20 I I I 0.15 Llmiling Forging Percent USE Decreaset 0.15 I\ I n rn 29.6% from Capsule W I 0.10 I I I 0.05 (langential orientation) I

- I I I I I I I I I

--- --~-

\ \\

\\

I Upper Limit

\ ~

-- -- .... ........ -- ~ ------ --

.... --~ ___ ,----~ -. -

~

... i--

r-.

Forging Une I

~- -

~ ~

w

\\

....... :::::::=:::

=---- ::::::: ---- -- ~ i--

f-J I Weld Heat 12ss31 I

-- \ l.

t/J

, ~

.5

--, £;: ---- -- -- -- -~-- -----= -- -- I Line t::: ........ r..--- ---- -- - ---- ---- ----

Q.

~

ICII ~

~

I Llmtting Weld Heat # 25531 Peroent USE Decrease is 22.1% from Capsule W J! 10.0

~

C ti

~

QI

~

Intermediate to Lower Shell Girth Weld I 72 EFPY 1/4T Fluenoe

4.43 x 1010 n/cm2 I

a,

~

Measured Forging Data for lower Shell

Measul9d Weld Data for Heat# 25531 I n EFPY 1/4T La.wr Shel Forging Fluence

4.53 x 1019 nfcm2 I

1.0 1.00e+1 7 1.00E+18 1.00!+19 1.00E+20 Neutron Fluence, n/cm 2 {E > 1 MeV)

Figure 5-2 Regulatory Guide 1.99, Revision 2, Position 2.2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for North Anna Unit 1 at SLR (72 EFPY)

WCAP-1 8364-NP March 2020 Revision I

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its val idation)

Westinghouse Non-Proprietary Class 3 5-9 100.0

,..._ % Copper

,..._ ~H!: M!:1!11 ~

- 0.35 n~n 0 .30 n?i::

n?i:: n?n t ' -

f----

0.20 0.15 Upper Limit f----

10.1,: 0 .10 I

- \

In 10 nn,:

~ .... -- -- ~ ~ .,,...-

1,,,.- i-

-- .... ---- --e------

~ '---

w

. . :::::::=:::

=---- =E ~

i,..- 1,,,.-

-- --- --- -;:::::::--..,-- ~

~

Cl)

, ~

i-

-- --[:. ~

~

.5 a.

i,..-

_:..-- --- --- .... ... -r--

1,,,.-

Q 2

~

i,..-

~

ti tll J!

C ti I:!

ti a.

10.0 i_.-

~

i,..-

Inlet Nozzle Forging 10 &

Nozzle to Upper Shell Weld lntenne<late to Lower Shell Girth Weld 72 EFPY 1/4T Fluence

4.52 x 101* n/om2 lnlermediate Sheff Forging 72 EFPY 1/4T Fluence

4.53 x I I I I 1019 n/rml

,.. 72 EFPY Surface Fluence

3.14 x 10 11 nlcm' I Lower Shell Forging L-Inlet Nozzle Forging 11 &

1 72 EFPY 1/4TFklence

4.62 x 1011 n/om2 1 Nozzle to Upper Shell Weld

~

72 EFPY Sooaoe Fluence * -

1.18 X 1017 n/cm2 Outlet Nozzle Forging 12 & _ _

L I -,I Uppernn1ennediate Girth Weld 72 EFPY 1/4T Fluence

2.24 x 101* n/cm2 Nozzle to Upper Shell Weld l

+-

72 EFPY Surface Fluenoe

1.82 X 1017 n/cm1 l l l I l 11 I

_ Upper Shel Forging 1.0 1.00E+17 I I I 1111 I 72 EFPY 1/-4T Fluence

1.93 x 1011 n/cmZ I I I I I I I I 1.00E+18 1.00E+19 1.00E+20 Neutron Fluence, n/cm 2 (E > 1 MeV)

Figure 5-3 Regulatory Guide 1.99, Revision 2, Position 1.2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for North Anna Unit 2 at SLR (72 EFPY)

WCAP-1 8364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 5-10

% Copper Base Metal Weld 0.35 0.30 100.0 0.30 I 0.25 n25 I I O?n I I I I I I I n?n I I I 0.15 0.15 I I o 10 I Lmlllng Forging Percent USE Decrease(

0.11 \ 005 --+-+-+*--!~---, LJ 16.7%from Capsule V 1 1--- --+-- - t - - - - t-1-+--t-+-t-t-1

- I n r- L - - ~(~ax~~~l ~one ~* ~m~~~~~n1.-,_ _ J---r--r--rtit=rHrH

\ \ \ Upper Limit I --::

\ \ \ I __,,,__,-- '---

\ . ,... ... --:------ - Forging I

\\

\,

" [.:---- _/-~--:::,......

i---:::::--~r:..;--- --- ,. .... - - ... "" -

Ii=

I

\~ ---~~:==~~~~-- --------

WeldHeat#

w 716126 Line

-~ ~--

en

, -.....-.....--... ~-- ~lo"'- ~ _ _ .. _

.5 Q

Q.

e .---:: ~~--~~i--""'

____. ~ ~

- --~--- ,_..........___.__,___..__,_ ____....__....,

--- - - - - - .........._Hlimtting Weld Heat# 716126 P*cent USE II CII J!

C II

~

II 10.0 -

~ - * ------ Decrease 14.4% from Caosule V D.

Intermediate to Lowar Shell Girth Weld 72 EFPY 1/.fT Fluence

4.52 x 10" nlcm2

- -----I---+-+-+--+--+--+-<>-<

Intermediate Shell FOl'ging 72 EFPY 1/.fT Fluence

4.53 x 10" n/cm2 Measured Forging Data for lntermedate Shell

Measured Weld Data for Heat# 716126 1.0 -l --!:=============i::-1----...L--L--1._L.LLLL-1-----1.--L_J-1..L..L...LLJ~

1.00E+17 1.00E+18 1.00E+19 1.00E+20 Neutron Fluence, n/cm 2 (E > 1 MeV)

Figure 5-4 Regulatory Guide 1.99, Revision 2, Position 2.2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for North Anna Unit 2 at SLR (72 EFPY)

WCAP-18364-NP March 2020 Revision I

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Westinghouse Non-Proprietary Class 3 6-1 6 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Heatup and cooldown limit curves are calculated using the most limiting value of RTNOT (reference nil-ductility transition temperature) corresponding to the limiting material in the beltline region of the RPV.

The most limiting RTNoT of the material in the core (beltline) region of the RPV is determined by using the unirradiated RPV material fracture toughness properties and estimating the irradiation-induced shift (LlRTNDT),

6.1 ADJUSTED REFERENCE TEMPERATURES CALCULATION RTNOT increases as the material is exposed to fast-neutron irradiation; therefore, to find the most limiting RTNDT at any time period in the reactor's life, LlRTNOT due to the radiation exposure associated with that time period must be added to the original unirradiated RTNoT, Using the adjusted reference temperature (ART) values, pressure-temperature (P-T) limit curves are determined in accordance with the requirements of 10 CFR Part 50, Appendix G (Reference 10), as augmented by Appendix G to Section XI of the ASME Boiler and Pressure Vessel Code (Reference 18).

P-T limit curves for SLR (72 EFPY) do not need to be submitted as part of the North Anna Units 1 and 2 License Renewal Application since P-T limit curves are available as a part of the current license. However, new P-T limit curve development or an extension of the applicability of the current curves must be completed prior to the expiration of the current curves as specified in the North Anna Units 1 and 2 licensing basis.

Nozzle forgings and their associated welds are evaluated in this report at EOLE and SLR as part of the extended beltline. The extended beltline materials evaluated are documented in Tables 6.1-3, 6.1-6, 6.1-9, and 6.1-12. All extended beltline materials were evaluated without consideration of attenuation through the vessel wall, i.e., with the fluence at the clad/base metal interface. The nozzle forging materials were assigned the fluence values at the postulated 1/4T flaw location for each specific nozzle in Tables 2-3 and 2-4. Only those materials with neutron fluence values greater than 1 x 10 17 n/cm 2 (E > 1.0 MeV) need to consider embrittlement. In order to fully assess the North Anna Units 1 and 2 P-T limit curves applicability to 72 EFPY, a nozzle corner fracture mechanics analysis was completed. These nozzle P-T limit curves were generated and compared to the beltline P-T Limit curves to ensure that the beltline curves are bounding. The detailed nozzle forging fracture mechanics evaluation and comparison to the beltline P-T limit curves was doc;umented in WCAP-18363-NP (Reference 19). Based on this, analysis, the current beltline P-T limit cu~es in the North Anna Power Station Technical Specifications \\;ere confirmed to be bounding. ' '.

The P-T limit curves for normal heatup and cooldown of the primary reactor coolant system for North Anna Units 1 and 2 were previously developed in WCAP-15112 (Reference 20). The existing P-T limit curves are based on the limiting beltline material ART values, which are influenced by both the fluence and the initial material properties of that material. Since the development of the curves, the fluence values and initial material properties used to calculate ART values have been updated.

To confirm or update the applicability of the current P-T limit curves at EOLE and generate new P-T limit curves for SLR, the limiting reactor vessel material ART values with consideration of the updated TLAA WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-2 fluence values, revised Position 2.1 chemistry factor values, and updated initial RTNDT values must be determined. The Regulatory Guide 1.99, Revision 2 (Reference 11) methodology was used along with the surface fluence of Section 2 to calculate ART values for the North Anna Units 1 and 2 reactor vessel materials at BOLE (50.3 EFPY and 52.3 EFPY for Units 1 and 2, respectively) ,and SLR (72 EFPY for both Units 1 and 2). Nozzle P-T limit curves were developed based on surfac6 fluence values to ensure conservatism in WCAP-18363-NP. The ART calculations are summarized in Tables 6.1-1 through 6.1-12 for North Anna Units 1 and 2.

\

WCAP-18364-NP March2020 Revision 1

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Westinghouse Non-Proprietary Class 3 6-3 Table 6.1-1 Calculation of the North Anna Unit 1 ART Values at the l/4T Location for the Reactor Vessel Beltline Materials at 50.3 EFPY<a)

R.G. Surface l/4T -

Flux Fluence<c) Fluence(d) Predicted Heat 1.99, CF(b) 1/4T RTNDT(U)(t) 0'1 O',\(g) M ART Material Type** FF<el LiRTNDT Number Rev.2 (x 1019 n/cm2, (x 10 19 n/cm2, {°F) {°F) (°F) (OF) (OF)

(Lot) (OF)

Position E> 1.0 MeV) E> 1.0 MeV)

Upper Shell 990286 I Forging 05 295213

- 1.1 121.50 0.215 0.136 0.481 1 58.4 0.0 17.0 34.0 93.4 Upper to Intermediate Shell SMIT89 25295 1.1 163.25 0.248 0.156 0.512 -40 83.6 0.0 28.0 56.0 99.6 Circumferential (1170)

Weld (OD 94%)

Using credible SW'Veillance data(h) 2.1 150.69 0.248 0.156 0.512 -40 77.2 0.0 14.0 28.0 65.2 Upper to Intermediate Shell SMIT89 4278 1.1 63.00 0.248 0.156 0.512 -4 32.3 0.0 16.1 32.3 60.6 Circumferential (1211)

Weld (ID 6%)(kl Using non-credible surveillance data<iJ 2.1 80.71 0.248 0.156 0.512 -4 41.3 0.0 20.7 41.3 78.7 Intermediate Shell 990311 /

Forging 04 298244 - 1.1 86.00 5.03 3.17 1.304 -6 112.1 0.0 17.0 34.0 140.1 Intermediate to Lower Shell SMIT 89 25531 1.1 56.22 5.02 3.17 1.304 -2 73.3 0.0 28.0 56.0 127.3 Circumferential (1211)

Weld - ..

Using non-credible surveillance data@ 2.1 67.53 5.02 3.17 1.304 -2 88.0 0.0 28.0 56.0 142.0 Lower Shell 990400 I Forging 03 292332 - 1.1 119.97 5.13 3.24 1.309 33 157.0 0.0 17.0 34.0 224.0 Using non-credible surveillance dataOJ 2.1 81.68 5.13 3.24 1.309 33 106.9 0.0 17.0 34.0 173.9 Notes contained on the following page.

WCAP-18364-NP March2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-4 Notes:

(a) The Regulatory Guide 1.99, Revisiop..2 (Reference 11) methodology was utilized in the calculation of the ART values.

(b) Chemistry factors are taken from Table 3-7 of this report.

(c) Fluence is taken from Table 2-3 of this report.

(d) The I/4T fluence values were calculated from the surface fluence, the reactor vessel beltline thickness (7.677 inches), and equation f= fsurf

e-0 -24 (xJ from Regulatory Guide 1.99, Revision 2, where x = the depth into the vessel wall (inches).

(e) FF= fluence factor= f(O.zs-o.ro*Iog (f)J_

(f) RTNDT(UJ values are taken from Table 3-1 of this report.

(g) Per the guidance of Regulatory Guide 1.99, Revision 2 (Reference 11 ), the base metal cra = l 7°F for Position I. I and Position 2.1 with non-credible surveillance data, and cra =

8.5°F for Position 2.1 with credible surveillance data. Also, per Regulatory Guide 1.99, Revision 2, the weld metal cra = 28°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and cra = 14°F for Position 2.1 with credible surveillance data. However, cra need not exceed 0.5*dRTNDT for either forgings or welds with or without surveillance data.

(h) The surveillance data for weld Heat# 25295 from the Sequoyah Unit I surveillance program were deemed credible per WCAP-17539-NP (Reference 17).

(i) The surveillance data for weld Heat# 4278 from the Sequoyah Unit 2 surveillance program were deemed non-credible per WCAP-17539-NP (Reference 17).

G) The credibility evaluation for the North Anna Unit I surveillance data in Appendix A. I of this report determined that the Lower Shell Forging 03 and weld Heat# 25531 surveillance data are deemed non-credible.

(k) Since this inner diameter (ID) weld is only 6% of the vessel thickness, the weld is not present at the I/4T location; hence, it is not applicable to this calculation. It is presented for information only.

WCAP-18364-NP March 2020 Revision 1

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Westinghouse Non-Proprietary Class 3 Table 6.1-2 Calculation of the North Anna Unit 1 ART Values at the 3/4T Location for the Reactor Vessel Beltline Materials at 50.3 EFPY(al R.G. Surface 3/4T Flux Predicted Heat 1.99, CF(b) FluenceCcJ FluenceCd) 3/4T RTNDT(U)(f) UI ut/g) M ART Material Type FF(el (OF) A.RTNDT Number Rev.2 (x 1019 n/cm2, (x 1019 n/cm2, {°F) (OF) {°F) {°F)

(Lot) (OF)

Position E> 1.0 MeV) E> 1.0 MeV)

Upper Shell 990286 I Forging 05 295213

- 1.1 121.50 0.215 0.0540 0.305 1 37.1 0.0 17.0 34.0 72.1 Upper to Intermediate Shell SMIT 89 \

25295 1.1 163.25 0.248 0.0623 0.329 -40 53.7 0.0 26.8 53.7 67.4 Circumferential (1170)

Weld (OD 94%)

Using credible surveillance data(h) 2.1 150.69 0.248 0.0623 0.329 -40 49.6 0.0 14.0 28.0 37.6 Upper to Intermediate Shell SMIT 89 4278 1.1 63.00 0.248 0.0623 0.329 -4 20.7 0.0 10.4 20.7 37.4 Circumferential (1211)

Weld (ID 6%)(k)

Using non-credible surveillance data(iJ 2.1 80.71 0.248 0.0623 0.329 -4 26.5 0.0 13.3 26.5 49.1 Intermediate Shell 990311 /

Forging 04 298244 - ,, 1.1 86.00 5.03 1.26 1.065 -6 91.6 0.0 17.0 34.0 119.6 Intermediate to Lower Shell SMIT 89 25531 1.1 56.22 5.02 1.26 1.065 -2 59.8 0.0 28.0 56.0 113.8 Circumferential (1211)

Weld Using non-credible surveillance dataO) 2.1 67.53 5.02 1.26 1.065 -2 71.9 0.0 28.0 56.0 125.9 Lower Shell 990400 I Forging 03 292332 - 1.1 119.97 5.13 1.29 1.070 33 128.4 0.0 17.0 34.0 195.4 Using non-credible surveillance dataOJ 2.1 81.68 5.13 1.29 1.070 33 87.4 0.0 17.0 34.0 154.4 Notes contained on the following page.

WCAP-18364-NP March2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-6 Notes:

(a) The Regulatory Guide 1.99, Revision 2 (Reference 11) methodology was utilized in the calculation of the ART values.

(b) Chemistry factors are taken from Table 3-7 of this report.

(c) Fluence is taken from Table 2-3 of this report.

(d) The 3/4T fluence values were calculated from the surface fluence, the reactor vessel beltline thickness (7.677 inches), and equation f= fsmf

e* 0*24 (x) from Regulatory Guide 1.99, Revision 2, where x = the depth into the vessel wall (inches).

(e) FF= fluence factor= f 0*28 10*Iog(f))_

(f) RTNDT(U) values are taken from Table 3-1 of this report.

(g) Per the guidance of Regulatory Guide 1.99, Revision 2 (Reference 11 ), the base metal 0'6 = l 7°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and 0'6 =

8.5°F for Position 2.1 with credible surveillance data. Also, per Regulatory Guide 1.99, Revision 2, the weld metal 0'6 = 28°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and 0'6 = 14°F for Position 2.1 with credible surveillance data. However, 0'6 need not exceed 0.5*.iRTNDT for either forgings or welds with or without surveillance data.

(h) The surveillance data for weld Heat# 25295 from the Sequoyah Unit I surveillance program were deemed credible per WCAP-17539-NP (Reference 17).

(i) The surveillance data for weld Heat# 4278 from the Sequoyah Unit 2 surveillance program were deemed non-credible per WCAP-17539-NP (Reference 17).

G) The credibility evaluation for the North Anna Unit I surveillance data in Appendix A. I of this report determined that the Lower Shell Forging 03 and weld Heat# 25531 surveillance data are deemed non-credible.

(k) Since this inner diameter (ID) weld is only 6% of the vessel thickness, the weld is not present at the 3/4T location; hence, it is not applicable to this calculation. It is presented for information only.

WCAP-18364-NP March 2020 Revision 1

- - This record was-final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-7 Table 6.1-3 Calculation of the North Anna Unit 1 ART Values for the Reactor Vessel Extended Beltline Materials at 50.3 EFPY(a)

R.G. Surface Predicted Heat Flux Type 1.99, CF(b) Fluence(c) Surf. RTNDT(U)(e) GI GA (f) M ART Material FF(d) ARTNDT Number (Lot) Rev. 2 (x 10 19 n/cm2, {°F) (OF) {°F) C°F) {°F)

{°F)

Position E> 1.0MeV)

Inlet Nozzle Forging 09 to Upper Shell Weld Rotterdam - 1.1 293.45 0.00617 0 30 0.0 0.0 0.0 0.0 30.0 Inlet Nozzle Forging 10 to Upper Shell Weld Rotterdam - 1.1 293.45 0.0219 0.182 30 53.4 0.0 26.7 53.4 136.8 Inlet Nozzle Forging 11 to Upper Shell Weld Rotterdam - 1.1 293.45 0.00833 0 30 0.0 0.0 0.0 0.0 30.0 Outlet Nozzle Forging 12 to Upper Shell Weld Rotterdam - 1.1 293.45 0.0127 0.129 30 37.8 0.0 18.9 37.8 105.5 Outlet Nozzle Forging 13 to Upper Shell Weld Rotterdam - 1.1 293.45 0.00359 0 30 0.0 0.0 0.0 0.0 30.0 Outlet Nozzle Forging 14 to Upper Shell Weld Rotterdam - 1.1 293.45 0.00484 0 30 0.0 0.0 0.0 0.0 30.0 Inlet Nozzle Forging 09 990290-11 - 1.1 96.00 0.00292 0 -14 0.0 0.0 0.0 0.0 -14.0 Inlet Nozzle Forging 10 990290-12 - 1.1 95.75 0.0104 0.113 -10 10.8 0.0 5.4 10.8 11.6 Inlet Nozzle Forging 11 990268-21 - 1.1 140.30 0.00394 0 8 0.0 0.0 0.0 0.0 8.0 Outlet Nozzle Forging 12 990290-31 - 1.1 96.00 0.00612 0 -6 0.0 0.0 0.0 0.0 -6.0 Outlet Nozzle Forging 13 990290-22 - 1.1 96.00 0.00172 0 -7 0.0 0.0 0.0 0.0 -7.0 Outlet Nozzle Forging 14 990290-21 - 1.1 96.00 0.00233 0 8 0.0 0.0 0.0 0.0 8.0 Notes contained on the following page.

WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-8 Notes:

(a) The Regulatory Guide 1.99, Revision 2 (Reference 11) methodology was utilized in the calculation of the ART values.

(b) Chemistry factors are taken from Table 3-7 of this report.

(c) Fluence is taken from Table 2-3 of this report. The fluence values for the nozzle forgings are taken at the postulated I/4T flaw axial location. The fluence values for the inlet/outlet nozzle to upper shell welds are taken at the lowest extent of the nozzle weld centerline. Analysis of the nozzle forgings and associated welds are conservatively performed using the surface fluence, neglecting attenuation through the reactor vessel wall. Embrittlement effects are considered only if the fluence is greater than 1017 n/cm2*

For materials with fluence less than 1017 n/cm2 the FF is set equal to 0.

(d) FF= fluence factor= f(O.zs-o.io*tog (fll.

(e) RTNDT(U) values are taken from Table 3-1 of this report.

(f) Per the guidance of Regulatory Guide 1.99, Revision 2 (Reference 11), the base metal at.= l 7°F for Position 1.1. Also, per Regulatory Guide 1.99, Revision 2, the weld metal at.= 28°F for Position 1. 1. However, at. need not exceed O.S*~RTNDT for either forgings or welds with or without surveillance data.

WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-9 Table 6.1-4 Calculation of the North Anna Unit 2 ART Values at the 1/4T Location for the Reactor Vessel Beltline Materials at 52.3 EFPY(a)

R.G. Surface 1/4T Flux Fluence<c) Predicted Heat 1.99, CF(b) Fluence<d) 1/4T RTNDT(U)(t) G1 Ge..(g) M ART Material Type FF<e> ARTNDT Number Rev. 2 (x 10 19 n/cm2, (x 10 19 n/cm2, (OF) {°F) {°F) (OF) {°F)

(Lot) {°F)

Position E> 1.0MeV) E>l.OMeV)

Upper Shell 990598 I Forging 05 291396

- 1.1 51.00 0.223 0.141 0.489 8 24.9 0.0 12.5 24.9 57.8 Upper to Intermediate Shell SMIT 89 4278 1.1 63.00 0.258 0.163 0.521 -4 32.8 0.0 16.4 32.8 61.7 Circumferential (1211)

Weld (OD 94%)

Using non-credible surveillance data(hJ 2.1 80.71 0.258 0.163 0.521 -4 42.1 0.0 21.0 42.1 80.1 Upper to Intermediate Shell SMIT 89 801 1.1 87.80 0.258 0.163 0.521 10 45.8 0.0 22.9 45.8 101.5 Circumferential (1211)

Weld (ID 6% )Gl Intermediate Shell 990496 I Forging 04 292424 - 1.1 74.00 5.25 3.31 1.314 69 97.2 0.0 17.0 34.0 200.2

/

Using non-credible surveillance data(i) 2.1 53.44 5.25 3.31 1.314 69 70.2 0.0 17.0 34.0 173.2 Intermediate to Lower Shell LW320 716126 1.1 36.09 5.24 3.31 1.314 -67 47.4 0.0 23.7 47.4 27.8 Circumferential (26)

Weld Using credible surveillance data(iJ 2.1 26.61 5.24 3.31 1.314 -67 35.0 0.0 14.0 28.0 -4.0 Lower Shell 990533 I Forging 03 297355 - 1.1 96.00 5.36 3.38 1.319 37 126.6 0.0 17.0 34.0 197.6 Notes contained on the following page.

WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-10 Notes:

(a) The Regulatory Guide 1.99, Revision 2 (Reference 11) methodology was utilized in the calculation of the ART values.

(b) Chemistry factors are taken from Table 3-8 of this report.

(c) Fluence is taken from Table 2-4 of this report.

(d) The 1/4T fluence values were calculated from the surface fluence, the reactor vessel beltline thickness (7.677 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).

(e) FF = fluence factor = fC0*28 - o. rn*Iog (l)J.

(f) RTNDT(U) values are taken from Table 3-2 of this report.

(g) Per the guidance of Regulatory Guide 1.99, Revision 2 (Reference 11 ), the base metal crt>. = l 7°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and crt>. =

8.5°F for Position 2.1 with credible surveillance data. Also, per Regulatory Guide 1.99, Revision 2, the weld metal crt>. = 28°F for Position I. 1 and Position 2.1 with non-credible surveillance data, and crt>. = 14°F for Position 2.1 with credible surveillance data. However, crt>. need not exceed 0.5*6RTNDT for either forgings or welds with or without surveillance data.

(h) The surveillance data for weld Heat# 4278 from the Sequoyah Unit 2 surveillance program were deemed non-credible per WCAP-17539-NP (Reference 17).

(i) The credibility evaluation for the North Anna Unit 2 surveillance data in Appendix A.2 of this report determined that the Intermediate Shell Forging 04 surveillance data are deemed non-credible; however, the weld Heat# 716126 surveillance data are deemed credible.

G) Since this inner diameter (ID) weld is only 6% of the vessel thickness, the weld is not present at the I/4T location; hence, it is not applicable to this calculation. It is presented for information only.

WCAP-18364-NP March 2020 Revision I

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-11 Table 6.1-5 Calculation of the North Anna Unit 2 ART Values at the 3/4T Location for the Reactor Vessel Beltline Materials at 52.3 EFPY(a)

R.G. Surface 3/4T Flux Fluence<cl Fluence<d) Predicted Heat 1.99, CF(b) 3/4T RTNDT(U)(t) 61 <;,.,_(g) M ART Material Type FF<*> ARTNDT Number Rev. 2 (x 10 19 n/cm2, (x 10 19 n/cm2, {°F) (OF) {°F) {°F) {°F)

(Lot) (OF)

Position E>l.OMeV) E> 1.0 MeV)

Upper Shell 990598 /

Forging 05 291396

- 1.1 51.00 0.223 0.0560 0.311 8 15.9 0.0 7.9 15.9 39.7 Upper to Intermediate Shell SMIT89 4278 1.1 63.00 0.258 0.0648 0.336 -4 21.1 0.0 10.6 21.1 38.3 Circumferential (1211)

Weld (OD 94%)

Using non-credible surveillance data(hJ 2.1 80.71 0.258 0.0648 0.336 -4 27.1 0.0 13.5 27.1 50.2 Upper to Intermediate Shell SMIT 89 Circumferential 801 (1211) 1.1 87.80 0.258 0.0648 0.336 10 29.5 0.0 14.7 29.5 68.9 Weld (ID 6%)m Intermediate Shell 990496 /

Forging 04 292424 - 1.1 74.00 5.25 1.32 1.077 69 79.7 0.0 17.0 34.0 182.7 Using non-credible surveillance datar) 2.1 53.44 5.25 1.32 1.077 69 57.5 0.0 17.0 34.0 160.5 Intermediate to Lower Shell LW320 716126 1.1 36.09 5.24 1.32 1.076 -67 38.8 0.0 19.4 38.8 10.7 Circumferential (26)

Weld Using credible surveillance datarJ 2.1 26.61 5.24 1.32 1.076 -67 28.6 0.0 14.0 28.0 -10.4 Lower Shell 990533 I Forging 03 297355

- 1.1 96.00 5.36 1.35 1.083 37 103.9 0.0 17.0 34.0 174.9 Notes contained on the following page.

WCAP-18364-NP March2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-12 Notes:

(a)* The Regulatory Guide 1.99, Revision 2 (Reference 11) methodology was utilized in the calculation of the ART values.

(b) Chemistry factors are taken from Table 3-8 of this report.

(c) Fluence is taken from Table 2-4 of this report.

(d) The 3/4T fluence values were calculated from the surface fluence, the reactor vessel beltline thickness (7.677 inches), and equation f= fsurf

e-o.24 (xl from Regulatory-Guide 1.99, Revision 2, where x = the depth into the vessel wall (inches).

(e) FF= fluence factor= f(O.zs-o.1o*Iog(t))_

(f) RTNDT(U) values are taken from Table 3-2 of this report.

(g) Per the guidance of Regulatory Guide 1.99, Revision 2 (Reference 11 ), the base metal cl",\,= l 7°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and c,8 =

8.5°F for Position 2.1 with credible surveillance data. Also, per Regulatory Guide 1.99, Revision 2, the weld metal c,8 = 28°F for Position 1. 1 and Position 2.1 with non-credible surveillance data, and c,8 = 14°F for Position 2.1 with credible surveillance data. However, c,8 need not exceed 0.5*LlRTNDT for either forgings or welds with or without surveillance data.

(h) The surveillance data for weld Heat# 4278 from the Sequoyah Unit 2 surveillance program were deemed non-credible per WCAP-17539-NP (Reference 17).

(i) The credibility evaluation for the North Anna Unit 2 surveillance data in Appendix A.2 of this report determined that the Intermediate Shell Forging 04 surveillance data are deemed non-credible; however, the weld Heat# 716126 surveillance data are deemed credible.

G) Since this inner diameter (ID) weld is only 6% of the vessel thickness, the weld is not present at the 3/4T location; hence, it is not applicable to this calculation. It is presented for information only.

WCAP-18364-NP March2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-13 Table 6.1-6 Calculation of the North Anna Unit 2 ART Values for the Reactor Vessel Extended Beltline Materials at 52.3 EFPY<al R.G. Surface Fluence<c) Predicted Heat Flux Type 1.99, CF(b) Surf. RTNDT(U)(e) 0'1 O't,.(f) M ART Material FF(d) ARTNDT Number (Lot) Rev. 2 (x 10 19 n/cm2, (OF) (OF) {°F) (OF) (OF)

(OF)

Position E> 1.0MeV)

Inlet Nozzle Forging 09 30(g) 1.1 163.20 0.00603 0 0.0 0.0 0.0 0.0 30.0 to Upper Shell Weld Inlet Nozzle Forging 10 30(g) 1.1 163.20 0.0227 0.186 30.3 0.0 15.2 30.3 90.7 to Upper Shell Weld Inlet Nozzle Forging 11 3o<s) 8816 1.1 163.20 0.00858 0 0.0 0.0 0.0 0.0 30.0 to Upper Shell Weld LW320 20459 Outlet Nozzle Forging 12 (26 & 28) 27622 1.1 163.20 0.0132 0.132 30(g) 21.5 0.0 10.8 21.5 73.1 to Upper Shell Weld Outlet Nozzle Forging 13 30(g) 1.1 163.20 0.00350 0 0.0 0.0 0.0 0.0 30.0 to Upper Shell Weld Outlet Nozzle Forging 14 30(g) 1.1 163.20 0.00498 0 0.0 0.0 0.0 0.0 30.0 to Upper Shell Weld Inlet Nozzle Forging 09 990426 - 1.1 150.40 0.00285 0 11 0.0 0.0 0.0 0.0 11.0 Inlet Nozzle Forging 10 54567-2 - 1.1 104.75 0.0107 0.115 5 12.0 0.0 6.0 12.0 29.0 Inlet Nozzle Forging 11 54590-2 - 1.1 118.25 0.00405 0 -31 0.0 0.0 0.0 0.0 -31.0 Outlet Nozzle Forging 12 990426-22 - 1.1 150.00 0.00633 0 8 0.0 0.0 0.0 0.0 8.0 Outlet Nozzle Forging 13 990426-31 - 1.1 149.60 0.00168 0 1 0.0 0.0 0.0 0.0 1.0 Outlet Nozzle Forging 14 791291 - 1.1 86.00 0.00239 0 -22 0.0 0.0 0.0 0.0 -22.0 Notes contained on the following page.

WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-14 Notes:

(a) The Regulatory Guide 1.99, Revision 2 (Reference 11) methodology was utilized in the calculation of the ART values.

(b) Chemistry factors are taken from Table 3-8 of this report.

(c) Fluence is taken from Table 2-4 of this report. The fluence values for the nozzle forgings are taken at the postulated I/4T flaw axial location. The fluence values for the inlet/outlet nozzle to upper shell welds are taken at the lowest extent of the nozzle weld centerline. Analysis of the nozzle forgings and associated welds are conservatively performed using the surface fluence, neglecting attenuation through the reactor vessel wall. Embrittlement effects are considered only if the fluence is greater than 10 17 n/cm2*

For materials with fluence less thap. 10 17 n/cm2 the FF is set equal to 0.

(d) FF= fluence factor= 1° 0-10

108 <!))_

(e) RTNDT(UJ values are taken from Table 3-1 of this report.

(f) Per the guidance of Regulatory Guide 1.99, Revision 2 (Reference 11), the base metal 0"6 = l 7°F for Position 1.1. Also, per Regulatory Guide 1.99, Revision 2, the weld metal 0"6 = 28°F for Position I.I. However, 0"6 need not exceed 0.5*iiRTNDT for either forgings or welds with or without surveillance data.

(g) The RTNDT(U) is based on the highest RTNDT(U) of the heats associated with this weld.

WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-15 Table 6.1-7 Calculation of the North Anna Unit 1 ART Values at the 1/4T Location for the Reactor Vessel Beltline Materials at 72 EFPY(a)

R.G. Surface 1/4T Flux Predicted Heat 1.99, CF(b) FluenceCc) Fluence<dJ 1/4T RTNDT(U}f) 0"1 O"A(g) M ART Material Type FF(e> ARTNDT Number Rev.2 (x 10 19 n/cm2, (x 1019 n/cm 2, (OF) {°F) {°F) {°F) {°F)

(Lot) .. {°F)

Position E> 1.0 MeV) E>l.OMeV)

Upper Shell 990286 /

Forging 05 295213

- 1.1 121.50 0.304 0.192 0.559 1 68.0 0.0 17.0 34.0 103.0 Upper to Intermediate Shell SMIT 89 25295 1.1 163.25 0.351 0.221 0.594 -40 97.0 0.0 28.0 56.0 113.0 Circumferential (1170)

Weld (OD 94%)

Using credible surveillance data(hJ 2.1 150.69 0.351 0.221 0.594 -40 89.5 0.0 14.0 28.0 77.5 Upper to Intermediate Shell SMIT89 4278 1.1 63.00 0.351 0.221 0.594 -4 37.4 0.0 18.7 37.4 70.8 Circumferential (1211)

Weld (ID 6%)(k)

Using non-credible surveillance data(iJ 2.1 80.71 0.351 0.221 0.594 -4 47.9 0.0 24.0 47.9 91.9 Intermediate Shell 990311 /

Forging 04 298244 - 1.1 86.00 7.07 4.46 1.379 -6 118.6 0.0 17.0 34.0 146.6 Intermediate to Lower Shell SMIT89 25531 1.1 56.22 7.04 4.44 1.378 -2 77.5 0.0 28.0 56.0 131.5 Circumferential (1211)

Weld Using non-credible surveillance dataOJ 2.1 67.53 7.04 4.44 1.378 -2 93.1 0.0 28.0 56.0 147.1 Lower Shell 990400 I Forging 03 292332 - 1.1 119.97 7.20 4.54 1.383 33 165.9 0.0 17.0 34.0 232.9 Using non-credible surveillance data(i} 2.1 81.68 7.20 4.54 1.383 33 113.0 0.0 17.0 34.0 180.0 Notes contained on the following page.

WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-16 Notes:

(a) The Regulatory Guide 1.99, Revision 2 (Reference 11) methodology was utilized in the calculation of the ART values.

(b) Chemistry factors are taken from Tab!~ 3-7 of this report.

(c) Fluence is taken from Table 2-3* of this report.

(d) The l/4T fluence values were calculated from the surface fluence, the reactor vessel beltline thickness (7.677 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).

(e) FF= fluence factor= f( 0.2s-o.10*1og(f))_

(f) RTNDT(U) values are taken from Table 3-1 of this report.

(g) Per the guidance of Regulatory Guide 1.99, Revision 2 (Reference 11), the base metal 0'6 = l 7°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and 0'6 =

8.5°F for Position 2.1 with credible surveillance data. Also, per Regulatory Guide 1.99, Revision 2, the weld metal 0'6 = 28°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and 0'6 = 14°F for Position 2.1 with credible surveillance data. However, 0'6 need not exceed 0.5*llRTNDT for either forgings or welds with or without surveillance data.

(h) The surveillance data for weld Heat# 25295 from the Sequoyah Unit 1 surveillance program were deemed credible per WCAP-17539-NP (Reference 17).

(i) The surveillance data for weld Heat# 4278 from the Sequoyah Unit 2 surveillance program were deemed non-credible per WCAP-17539-NP (Reference 17).

G) The credibility evaluation for the North Anna Unit 1 surveillance data in Appendix A. I of this report determined that the Lower Shell Forging 03 and weld Heat # 25531 surveillance data are deemed non-credible.

(k) Since this inner diameter (ID) weld is only 6% of the vessel thickness, the weld is not present at the 1/4T location; hence, it is not applicable to this calculation. It is presented for information only.

WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-17 Table 6.1-8 Calculation of the North Anna Unit 1 ART Values at the 3/4T Location for the Reactor Vessel Beltline Materials at 72 EFPY(a)

R.G. Surface 3/4T Flux Predicted aig)

Heat 1.99, CF(b)

Fluence(c) Fluence<d) 3/4T RTNDT(U)(t) O'I M ART Material Type FF<e> ARTNDT Number Rev.2 (x 10 19 n/cm2, (x 1019 n/cm2, (OF) {°F) (OF) {°F) (OF)

(Lot) {°F)

Position E>l.OMeV) E>l.OMeV)

Upper Shell 990286 /

Forging 05 295213

- 1.1 121.50 0.304 0.0763 0.365 1 44.4 0.0 17.0 34.0 79.4 Upper to Intermediate Shell SMIT 89 25295 1.1 163.25 0.351 0.0881 0.392 -40 64.0 0.0 28.0 56.0 80.0 Circumferential (1170)

Weld (OD 94%)

Using credible surveillance datarl) 2.1 150.69 0.351 0.0881 0.392 -40 59.1 0.0 14.0 28.0 47.1 Upper to Intermediate Shell . 4278 SMIT 89 1.1 63.00 0.351 0.0881 0.392 -4 24.7 0.0 12.4 24.7 45.4 Circumferential (1211)

Weld (ID 6%)(k)

Using non-credible surveillance data(/J 2.1 80.71 0.351 0.0881 0.392 -4 31.6 0.0 15.8 31.6 59.3 Intermediate Shell 990311 /

Forging 04 298244

- 1.1 86.00 7.07 1.78 1.158 -6 99.6 0.0 17.0 34.0 127.6 Intermediate to Lower Shell SMIT89

25531 1.1 56.22 7.04 1.77 1.157 -2 65.0 0.0 28.0 56.0 119.0 Circumferential (1211)

Weld Using non-credible surveillance dataOJ 2.1 67.53 7.04 1.77 1.157 -2 78.1 0.0 28.0 56.0 132.1 Lower Shell 990400 I Forging 03 292332

- 1.1 119.97 7.20 1.81 1.163 33 139.5 0.0 17.0 34.0 206.5 Using non-credible surveillance dataOJ 2.1 81.68 7.20 1.81 1.163 .,.,

"" 95.0 0.0 17.0 34.0 162.0 Notes contained on the following page.

WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-18 Notes:

(a) The Regulatory Guide 1.99, Revision 2 (Reference 11) methodology was utilized in the calculation of the ART values.

(b) Chemistry factors are taken from Table 3-7 of this report.

(c) Fluence is taken from Table 2-3 of this report.

(d) The 3/4T fluence values were calculated from the surface fluence, the reactor vessel beltline thickness (7.677 inches), and equation f= f,urf

e* 0*24 (x) from Regulatory Guide 1.99, Revision 2, where x = the depth info the vessel wall (inches).

(e) FF = fluence factor = fC0*28 - O. lO*log Cf)).

(f) RTNDT(U) values are taken from Table 3-1 ofthis report.

(g) Per the guidance of Regulatory Guide 1.99, Revision 2 (Reference 11 ), the base metal crt. = l 7°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and crt. =

8.5°F for Position 2.1 with credible surveillance data. Also, per Regulatory Guide 1.99, Revision 2, the weld metal crt. = 28°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and crt. = 14°F for Position 2.1 with credible surveillance data. However, crt. need not exceed 0.5*~TNDT for either forgings or welds with or without surveillance data.

(h) The surveillance data for weld Heat# 25295 from the Sequoyah Unit I surveillance program were deemed credible per WCAP-17539-NP (Reference 17).

(i) The surveillance data for weld Heat# 4278 from the Sequoyah Unit 2 surveillance program were deemed non-credible per WCAP-17539-NP (Reference 17).

G) The credibility evaluation for the North Anna Unit I surveillance data in Appendix A. I of this report determined that the Lower Shell Forging 03 and weld Heat # 25531 surveillance data are deemed non-credible.

(k) Since this inner diameter (ID) weld is only 6% of the vessel thickness, the weld is not present at the 3/4T location; hence, it is not applicable to this calculation. It is presented for information only.

WCAP-18364-NP March 2020 Revision I

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-19 Table 6.1-9 Calculation of the North Anna Unit 1 ART Values for the Reactor Vessel Extended Beltline Materials at 72 EFPY(a)

R.G. Surface Predicted Heat Flux Type 1.99, CF(b) Fluence<c) Surf. RTNDT(U/*> 0'1 O'A(f) M ART Material FF(d) ARTNDT Number (Lot) Rev.2 (x 10 19 n/cm2, (OF) (OF) (OF) (OF) (OF)

(°F)

Position E>l.OMeV)

Inlet Nozzle Forging 09 to Upper Shell Weld Rotterdam - 1.1 293.45 0.00898 0 30 0.0 0.0 0.0 0.0 30.0 Inlet Nozzle Forging 10 to Upper Shell Weld Rotterdam - 1.1 293.45 0.0313 0.225 30 66.1 ,o.o 28.0 56.0 152.1 Inlet Nozzle Forging 11 to Upper Shell Weld Rotterdam - 1.1 293.45 0.0120 0.124 30 36.4 0.0 18.2 36.4 102.7 Outlet Nozzle Forging 12 Rotterdam - 1.1 293.45 0.0182 0.162 30 47.6 0.0 23.8 47.6 125.2 to Upper Shell Weld Outlet Nozzle Forging 13 to Upper Shell Weld Rotterdam - 1.1 293.45 0.00522 0 30 0.0 0.0 0.0 0.0 30.0 Outlet Nozzle Forging 14 to Upper Shell Weld Rotterdam - 1.1 293.45 0.00697 0 30 0.0 0.0 0.0 0.0 30.0

' Forgmg Inlet Nozzle . 09 990290-11 - 1.1 96.00 0.00425 0 -14 0.0 0.0 0.0 0.0 -14.0 Inlet Nozzle Forging 10 990290-12 - 1.1 95.75 0.0148 0.142 -10 13.6 0.0 6.8 13.6 17.2 Inlet Nozzle Forging 11 990268-21 - 1.1 140.30 0.00568 0 8 0.0 0.0 0.0 0.0 8.0 Outlet Nozzle Forging 12 990290-31 - 1.1 96.00 0.00875 0 -6 0.0 0.0 0.0 0.0 -6.0 Outlet Nozzle Forging 13 990290-22 - 1.1 96.00 0.00251 0 -7 0.0 0.0 0.0 0.0 -7.0 Outlet Nozzle Forging 14 990290-21 - 1.1 96.00 0.00335 0 8 0.0 0.0 0.0 0.0 8.0 Notes contained on the following page.

WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-20 Notes:

(a) The Regulatory Guide 1.99, Revision 2 (Reference 11) methodology was utilized in the calculation of the ART values.

(b) Chemistry factors are taken from Table 3-7 of this report.

(c) Fluence is taken from Table 2-3 of this report. The fluence values for the nozzle forgings are taken at the postulated 1/4T flaw axial location. The fluence values for the inlet/outlet nozzle to upper shell welds are taken at the lowest extent of the nozzle weld centerline. Analysis of the nozzle forgings and associated welds are conservatively performed using the surface fluence, neglecting attenuation through the reactor vessel wall. Embrittlement effects are considered only if the fluence is greater than 10 17 n/cm2

For materials with fluence less than 10 17 n/cm2 the FF is set equal to 0.

(d) FF = fluence factor = f(0 o. IO'log (!)).

(e) RTNDT(U) values are taken from Table 3-1 ofthis report.

(f) Per the guidance of Regulatory Guide 1.99, Revision 2 (Reference 11), the base metal 0'8 = l 7°F for Position 1.1. Also, per Regulatory Guide 1.99, Revision 2, the weld metal 0'8 = 28°F for Position 1.1. However, 0'8 need not exceed 0.5*ti.RTNDT for either forgings or welds with or without surveillance data.

WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-21 Table 6.1-10 Calculation of the North Anna Unit 2 ART Values at the 1/4T Location for the Reactor Vessel Beltline Materials at 72 EFPY(al R.G. Surface 1/4T Flux I FluenceCc) Predicted aigl Heat 1.99, CF(b) FluenceCdl 1/4T RTNDT(U) (t) G1 M ART Material Type FF<e> ARTNDT Number Rev.2 (x 10 19 n/cm2, (x 10 19 n/cm2, {°F) (OF) (OF) (OF) (OF)

(Lot) (OF)

Position E>l.OMeV) E>l.OMeV)

Upper Shell 990598 I Forging 05 291396

- 1.1 51.00 0.307 0.194 0.562 8 28.7 0.0 14.3 28.7 65.3 Upper to Intermediate Shell SMIT 89 4278 1.1 63.00 0.355 0.224 0.597 -4 37.6 0.0 18.8 37.6 71.2 Circumferential (1211)

Weld (OD 94%)

Using non-credible surveillance data(!,) 2.1 80.71 0.355 0.224 0.597 -4 48.2 0.0 24.1 48.2 92.3 Upper to Intermediate Shell SMIT 89 801 1.1 87.80 0.355 0.224 0.597 10 52.4 0.0 26.2 52.4 114.8 Circumferential (1211)-

Weld (ID 6%)Gl Intermediate Shell 990496 I Forging 04

- 1.1 74.00 7.20 4.54 1.383 69 102.3 0.0 17.0 34.0 205.3 292424 Using non-credible surveillance data(iJ 2.1 53.44 7.20 4.54 1.383 69 73.9 0.0 17.0 34.0 176.9 Intermediate to Lower Shell LW320 716126 1.1 36.09 7.18 4.53 1.382 -67 49.9 0.0 24.9 49.9 32.8 Circumferential (26)

Weld Using credible surveillance data(iJ 2.1 26.61 7.18 4.53 1.382 -67 36.8 0.0 14.0 28.0 -2.2 Lower Shell 990533 I

- 1.1 96.00 7.34 4.63 1.387 37 133.2 0.0 17.0 34.0 204.2 Forging 03 297355 Notes contained on the following page.

WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-22 Notes:

(a) The Regulatory Guide 1.99, Revision 2 (Reference 11) methodology was utilized in the calculation of the ART values.

(b) Chemistry factors are taken from Table 3-8 of this report.

(c) Fluence is taken from Table 2-4 of this report.

(d) The l/4T fluence values were calculated from the surface fluence, the reactor vessel beltline thickness (7.677 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).

(e) FF = fluence factor = f( 0 o. lO*log (f)J.

(f) RTNDT(U) values are taken from Table 3-2 of this report.

(g) Per the guidance of Regulatory Guide 1.99, Revision 2 (Reference 11 ), the base metal cr,-. = l 7°F for Position 1. 1 and Position 2.1 with non-credible surveillance data, and cr,-. =

8.5°F for Position 2.1 with credible surveillance data. Also, per Regulatory Guide 1.99, Revision 2, the weld metal crt. = 28°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and cr,-. = 14°F for Position 2.1 with credible surveillance data. However, crt. need not exceed 0.5*~RTNDT for either forgings or welds with or without surveillance data (h) The surveillance data for weld Heat# 4278 from the Sequoyah Unit 2 surveillance program were deemed non-credible per WCAP-17539-NP (Reference 17).

(i) The credibility evaluation for the North Anna Unit 2 surveillance data in Appendix A.2 of this report determined that the Intermediate Shell Forging 04 surveillance data are deemed non-credible; however, the weld Heat# 716126 surveillance data are deemed credible.

G) Since this inner diameter (ID) weld is only 6% of the vessel thickness, the weld is not present at the l/4T location; hence, it is not applicable to this calculation. It is presented for information only.

WCAP-18364-NP March 2020 Revision 1

- - This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-23 Table 6.1-11 Calculation of the North Anna Unit 2 ART Values at the 3/4T Location for the Reactor Vessel Beltline Materials at 72 EFPY(a)

R.G. Surface 3/4T Flux Predicted Heat 1.99, CF(b)

FluenceCcl Fluence<dl 3/4T RT NDT(U)(f) UJ Ut. (g) M ART Material Type FF<e) ARTNDT Number Rev.2 (x 1019 n/cm 2, (x 1019 n/cm2, {°F) {°F) {°F) {°F) {°F)

(Lot) {°F)

Position E>l.OMeV) E> 1.0 MeV)

Upper Shell 990598 /

Forging 05 291396

- 1.1 51.00 0.307 0.0771 0.367 8 18.7 0.0 9.4 18.7 45.4 Upper to Intermediate Shell SMIT 89 4278 1.1 63.00 0.355 0.0891 0.394 -4 24.8 0.0 12.4 24.8 45.7 Circumferential (1211)

Weld (OD 94%)

Using non-credible surveillance datafhJ. 2.1 80.71 0.355 0.0891 0.394 -4 31.8 0.0 15.9 31.8 59.6 Upper to Intermediate Shell SMIT89 801 1.1 87.80 0.355 0.0891 0.394 10 34.6 0.0 17.3 34.6 79.2 Circumferential (1211)

Weld (ID 6%)Cil -

Intermediate Shell 990496 /

Forging 04 292424

- 1.1 74.00 7.20 1.81 1.163 69 86.0 0.0 17.0 34.0 189.0 Using non-credible surveillance data(iJ 2.1 53.44 7.20 1.81 1.163 69 62.1 0.0 17.0 34.0 165.1 Intermediate to Lower Shell LW320 716126 1.1 36.09 7.18 1.80 1.162 -67 41.9 0.0 21.0 41.9 16.9 Circumferential (26)

Weld Using credible surveillance data(i) 2.1 26.61 7.18 1.80 1.162 -67 30.9 0.0 14.0 28.0 -8.1 Lower Shell 990533 I

- 1.1 96.00 7.34 1.84 1.168 37 112.1 0.0 17.0 34.0 183.1 Forging 03 297355 Notes contained on the following page.

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Westinghouse Non-Proprietary Class 3 6-24 Notes:

(a) The Regulatory Guide 1.99, Revision 2 (Reference 11) methodology was utilized in the calculation of the ART values.

(b) Chemistry factors are taken from Table 3-8 of this report.

(c) Fluence is taken from Table 2-4 of this report.

(d) The 3/4T fluence values were calculated from the surface fluence, the reactor vessel beltline thickness (7.677 inches), and equation f= fsurr

e* 0*24 (x) from Regulatory Guide 1.99, Revision 2, where x = the depth into the vessel wall (inches).

(e) FF = fluence factor = f( 0 o. 1o*Jog (f)).

(f) RTNDT(U) values are taken from Table 3-2 ofthis report.

(g) Per the guidance of Regulatory Guide 1.99, Revision 2 (Reference 11), the base metal 0'8 = 17°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and 0'8 =

8.5°F for Position 2.1 with credible surveillance data. Also, per Regulatory Guide 1.99, Revision 2, the weld metal 0'8 = 28°F for Position 1.1 and Position 2.1 with non-credible surveillance data, and 0'8 = 14°F for Position 2.1 with credible surveillance data. However, 0'8 need not exceed O.S*~TNDT for either forgings or welds with or without surveillance data.

(h) The surveillance data for weld Heat# 4278 from the Sequoyah Unit 2 surveillance program were deemed non-credible per WCAP-17539-NP (Reference 17).

(i) The credibility evaluation for the North Anna Unit 2 surveillance data in Appendix A.2 of this report determined that the Intermediate Shell Forging 04 surveillance data are deemed non-credible; however, the weld Heat# 716126 surveillance data are deemed credible.

G) Since this inner diameter (ID) weld is only 6% of the vessel thickness, the weld is not present at the 3/4T location; hence, it is not applicable to this calculation. It is presented for information only.

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Westinghouse Non-Proprietary Class 3 6-25 Table 6.1-12 Calculation of the North Anna Unit 2 ART Values for the Reactor Vessel Extended Beltline Materials at 72 EFPY<al R.G. Surface Predicted Heat Flux Type 1.99, CF(b) Fluence<c) Surf. RT NDT(U/el 0'1 0'11.(f) M ART Material FF(d) ARTNDT Number (Lot) Rev. 2 (x 10 19 n/cm2, (OF) {°F) (OF) {°F) (OF)

{°F)

Position E> 1.0MeV)

Inlet Nozzle Forging 09 30(g) 1.1 163.20 0.00826 0 0.0 0.0 0.0 0.0 30.0 to Upper Shell Weld Inlet Nozzle Forging 10 30(g) 1.1 163.20 0.0314 0.226 36.8 0.0 18.4 36.8 103.6 to Upper Shell Weld Inlet Nozzle Forging 11 30(g) 8816 1.1 163.20 0.0118 0.123 20.0 0.0 10.0 20.0 70.0 to Upper Shell Weld LW320 20459 Outlet Nozzle Forging 12 (26 & 28) 27622 1.1 163.20 0.0182 0.162 30(g) 26.5 0.0 13.2 26.5 82.9 to Upper Shell Weld Outlet Nozzle Forging 13 30(g) 1.1 163.20 0.00479 0 0.0 0.0 0.0 0.0 30.0 to Upper Shell Weld Outlet Nozzle Forging 14 30(g) 1.1 163.20 0.00687 0 0.0 0.0 0.0 0.0 30.0 to Upper Shell Weld Inlet Nozzle Forging 09 990426 - 1.1 150.40 0.00390 0 11 0.0 0.0 0.0 0.0 11.0 Inlet Nozzle Forging 10 54567-2 - 1.1 104.75 0.0148 0.142 5 14.9 0.0 7.4 14.9 34.8 Inlet Nozzle Forging 11 54590-2 - 1.1 118.25 0.00559 0 -31 0.0 0.0 0.0 0.0 -31.0 Outlet Nozzle Forging 12 990426-22 - 1.1 150.00 0.00875 0 8 0.0 0.0 0.0 0.0 8.0 Outlet Nozzle Forging 13 990426-31 - 1.1 149.60 0.00230 0 1 0.0 0.0 0.0 0.0 1.0 Outlet Nozzle Forging 14 791291 - 1.1 86.00 0.00330 0 -22 0.0 0.0 0.0 0.0 -22.0 Notes contained on the following page.

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Westinghouse Non-Proprietary Class 3 6-26 Notes:

(a) The Regulatory Guide 1.99, Revision 2 (Reference 11) methodology was utilized in the calculation of the ART values.

(b) Chemistry factors are taken from Table 3-8 of this report.

-(c) Fluence is taken from Table 2-4 of this report. The fluence values for the nozzle forgings are taken at the postulated 1/4T flaw axial location. The fluence values for the inlet/outlet nozzle to upper shell welds are taken at the lowest extent of the nozzle weld centerline. Analysis of the nozzle forgings and associated welds are conservatively performed using the surface fluence, neglecting attenuation through the reactor vessel wall. Embrittlement effects are considered only if the fluence is greater than 10 17 n/cm2*

For materials with fluence less than 10 17 n/cm2 the FF is set equal to 0.

(d) FF = fluence factor = f 0*28 - o. rn*Iag (f)J.

(e) RTNDT(UJ values are taken from Table 3-2 ofthis report.

(f) Per the guidance of Regulatory Guide 1.99, Revision 2 (Reference 11), the base metal cr11 = 17°F for Position I.I. Also, per Regulatory Guide 1.99, Revision 2, the weld metal cr11 = 28°F for Position 1.1. However, cr11 need not exceed 0.5*1'1RTNDr for either forgings or welds with or without surveillance data.

(g) The RTNDr(U) is based on the highest RTNDT(UJ of the heats associated with this weld.

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Westinghouse Non-Proprietary Class 3 6-27 6.2 P-T LIMIT CURVES APPLICABILITY This section determines the applicability term of the BOLE P-T limit curves by comparing the ART values contained in the analysis of record (AOR) with the ART values calculated using the updated fluence projections and materials information contained herein. If the ART values used in the previous analysis are higher or equal to the ART values calculated using the updated fluence and material properties, then the applicability term of the current curves will remain unchanged. If the ART values used in the previous analysis are lower than the ART values calculated using the updated fluence and material properties, then the applicability term of the current curves may need to be shortened. This new period of applicability can be calculated based on a comparison of the ART values and linear interpolation using the fluence projections.

Tables 6.1-7 through 6.1-12 calculates the extended beltline, l/4T, and 3/4T ART calculations for North Anna Units 1 and 2 at SLR (72 EFPY for both Units 1 and 2). The limiting SLR ART values for North Anna Units 1 and 2 correspond to the North Anna Unit 2 Lower Shell Forging 03. Note, the ART values calculated for Unit 1 Lower Shell Forging 03 and Unit 2 Intermediate Shell Forging 04 based on Regulatory Guide 1.99, Revision 2, Position 1.1 are higher. However, the use of the lesser of the Regulatory Guide 1.99, Revision 2, Position 1.1 and 2.1 CFs with non-credible data and a full margin term is justified since none of the surveillance data are more than two times sigma-delta above the Position 1.1 CF trend line.

This determination is documented in Appendix C.

Table 6.2-1 compares the TLAA limiting ART values at SLR to the limiting ART values used in development of the existing EOLE P-T limit curves implemented in the Technical Specifications which are based on WCAP-15112 (Reference 20). The 1/4T and 3/4T limiting ART values at 72 EFPY are less than the 1/4T and 3/4T ART values used in the current Technical Specifications. This decrease is driven by the reduction in the initial RTNDT of the limiting material, i.e. the Unit 2 Lower Shell Forging 03 initial RTNnT reduced from 56°F to 37°F.

Table 6.2-1 Summary of the Limiting ART Values l/4T Location 3/4T Location 218.5 195.6 ART in Current Technical Specifications<hl Limiting Material: Unit 2 Lower Shell Forging 03

('.'F) ( developed using Position 1.1 data) 180.0 162.0 Unit 1 Limiting ART at SLR Limiting Material: Unit 1 Lower Shell Forging 03 (OF) ( developed using Position 2.1 with non-credible surveillance data and full margin term) 204.2 183.1 Unit 2 Limiting ART at SLR Limiting Material: Unit 2 Lower Shell Forging 03 (OF)

( developed using Position 1.1 data)

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Westinghouse Non-Proprietary Class 3 6-28 Table 6.2-1 shows that the SLR ART values at the 1/4T and 3/4T locations remain bounded by the ART values used in the current P-T limit curves. Thus, the P-T limit curves implemented in the North Anna Units 1 and 2 Technical Specifications will remain valid through SLR (72.0 EFPY) for the cylindrical shell materials.

The P-T limit curves also consider the 10 CFR 50, Appendix G "flange-notch" and the ASME Lowest Service Temperature (LST) requirements. For the flange notch requirements, the limiting reactor vessel closure head and flange initial RTNDT value of -22°F identified in Table 3-3 remains consistent with the limiting flange RTNnT value in WCAP-15112 (Reference 20); thus, the P-T limit curves flange notch requirements requires no change or further consideration. The LST requirements are not applicable to North Anna Units 1 and 2 because the plants are Westinghouse-designed and utilize stainless steel reactor coolant system piping. See Appendix C of WCAP-18363-NP for further details. It is noted that slightly different reactor vessel dimensions were used in WCAP-15112 than in the latest P-T limits analysis in WCAP-18363-NP, but this difference is negligible.

WCAP-18363-NP contains limiting nozzle P-T limit curves for Unit 1 and Unit 2 through SLR, and compares these nozzle P-T limit curves to the P-T limit curves in WCAP-15112, which is the basis for the current Technical Specifications P-T limit curves. WCAP-18363-NP concludes that the beltline P-T limits curves in WCAP-15112 bound the inlet/outlet nozzle P-T limit curves through SLR.

Conclusion Based on the SLR ART values calculated herein and the P-T limits analysis completed in WCAP-18363-NP (Reference 19), the P-T limit curves currently in the Technical Specifications based on evaluations in WCAP-15112 will remain valid through 72 EFPY.

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Westinghouse Non-Proprietary Class 3 7-1 7 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULES This section provides recommended capsule withdrawal schedules for North Anna Units 1 and 2 as well as technical justifications and demonstration of the schedules' compliance with ASTM E185-82 (Reference 21), as prescribed by 10 CFR 50, Appendix H (Reference 22). As shown below, each unit requires the withdrawal of 4 capsules in order to meet the requirements of ASTM El 85-82. To meet this, one additional capsule needs to be withdrawn from each unit at the 80-year equivalent fluence. To assist in asset management, Dominion plans to supplement the requirements of ASTM E185-82 by withdrawing an additional capsule from each unit at the 100-year equivalent.fluence.

10 CFR 50, Appendix H (Reference 22) states:

The design of the surveillance program and the withdrawal schedule must meet the requirements of the edition of ASTM E 185 that is current on the issue date of the ASME Code to which the reactor vessel was purchased. Later editions ofASTM E 185 may be used, but including only those editions through 1982. For each capsule withdrawal, the test procedures and reporting requirements must meet the requirements of ASTM E 185-82 to the extent practicable for the configuration of the specimens in the capsule.

The original North Anna Units 1 and 2 reactor vessels were designed and constructed to ASME Section III, 1968 Edition through Winter 1968 Addenda. Thus, per 10 CFR 50, Appendix H, the North Anna Units 1 and 2 surveillance program withdrawal schedules may meet the requirements of any version of the ASTM E185 standard from the 1966 version (which was current on the issue date of the ASME Code to which the reactor vessels were purchased) through the 1982 version. Per WCAP-8771 (Reference 27) and WCAP-8772 (Reference 28), the North Anna Units 1 and 2 surveillance capsule programs were designed in 1976 to the ASTM E185-73 (Reference 29) standard, which was the version active at that time. Therefore, the requirements of 10 CFR 50, Appendix H were met at the time of the design of the reactor vessel surveillance program.

Since that time North Anna has implemented the capsule withdrawal schedules in Updated Final Safety Analysis Report (UFSAR) (Reference 30), Tables 5 .4-2 and 5 .4-3 to meet the requirements of ASTM El 85-82 (Reference 21 ). The first step in determining the surveillance capsule withdrawal schedule compliance is to determine the minimum number of capsules to withdraw and/or test. ASTM E-185-82 bases the number of capsules on the maximum ~T NDT projected at the vessel surface for all reactor vessel materials.

Per Tables 4-1 and 4-2 of this report, the maximum t1RTNnT values for the North Anna Units 1 and 2 are 176.0°F and 14 l .2°F, respectively. Since the maximum t1RTNDT are projected to be above 100°F, but below 200°F, four (4) capsules are required to be pulled from each unit per Table 1 of ASTM E185-82. To date, three {3) capsules have been pulled and tested from each unit. The fourth capsules are currently scheduled for withdrawal and testing.

Because ASTM E185-82 is based on plant operation during the original 40-year license term, the requirements are supplemented herein using NUREG-1801, Revision 2 (GALL, Reference 23) and NUREG-2191 (GALL-SLR, Reference 24). The latest recommended surveillance capsule withdrawal schedules for North Anna Units 1 and 2, which include future capsule withdrawals to support the current 60-year operating licenses and the proposed 80-year operating licenses, are provided in Tables 7-1 and 7-2, respectively. These schedules meet the recommendations of ASTM E185-82 as required by 10 CFR 50, WCAP-18364-NP March 2020 Revision I

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Westinghouse Non-Proprietary Class 3 7-2 Appendix H and satisfy the guidance contained in the GALL and GALL-SLR. Specific details are described herein.

For North Anna Unit 1, it is recommended that, in order to satisfy the withdrawal requirements for the 60-year and 80-year operating licenses, Capsule X be withdrawn and tested at the refueling outage nearest to but following 39.1 EFPY (projected to occur in 2025), which is when fluence value of the capsule is

projected to reach the projected 80-year (72 EFPY) peak vessel fluence (7.20 x 10 19 n/cm2). This fluence value is also below twice the 60-year RV peak fluence to support the current 60-year license (50.3 EFPY),

and therefore also satisfies the existing license requirement for surveillance capsule withdrawal and testing.

To assist in asset management, Capsule Z should be withdrawn at the vessel refueling outage nearest to but following 59.4 EFPY of plant operation (projected to occur in 2046), which is when the fluence on the capsule will have reached at least the projected 100-year (90 EFPY) peak vessel fluence (8.92 x 10 19 n/cm 2).

This fluence is also below twice the 80-year RV pea.1<: fluence, and therefore, also satisfies the recommendations for an 80-year life. Standby Capsules T, Y, and S should remain in the reactor. For asset management consideratfons, if additional metallurgical data or dosimetry measurements are desired, withdrawal and testing of one of these capsules may be considered. Capsules T, Y, and S may remain in the RV through 72 EFPY.

For North Anna Unit 2, it is recommended that, in order to satisfy the withdrawal requirements for the 60-year and 80-year operating licenses, Capsule X be withdrawn and tested at the refueling outage nearest to but following 39.3 EFPY (projected to occur in 2026), which is when the fluence value of the capsule is projected to reach the projected 80-year (72 EFPY) peak vessel fluence (7.34 x 10 19 n/cm 2). This fluence value is also below twice the 60-year RV peak fluence to support the current 60-year license (52.3 EFPY),

and therefore also satisfies the existing license requirement for surveillance capsule withdrawal and testing.

To assist in asset management, Capsule Z should be withdrawn at the vessel refueling outage nearest to but following 56.1 EFPY of plant operation (projected to occur in 2043), which is when the fluence on the capsule will have reached at least the projected 100-year (90 EFPY) peak vessel fluence (9 .15 x 10 19 n/cm2).

This fluence is also below twice the 80-year RV peak fluence, and therefore, also satisfies the recommendations for an 80-year life. Standby Capsules T, Y, and S should remain in the reactor. For asset management considerations, if additional metallurgical data or dosimetry measurements are desired, withdrawal and testing of one of these capsules may be considered. Capsules T, Y, and S may remain in the RV through 72 EFPY.

Tables 7-3 and 7-4 compare the recommended schedules for North Anna Units 1 and 2, respectively, with the requirements of ASTM El 85~82, and the guidance of the GALL and GALL-SLR. The tables

.,demonstrate that these schedules meet the recommendations of -0,BTM E 185-82 as required by 10 CFR 50, Appendix H and satisfy the guidance contained in the GALL an~ GALL-SLR.

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Westinghouse Non-Proprietary Class 3 7-3 Table 7-1 North Anna Unit 1 Recommended Surveillance Capsule Withdrawal Schedule<a)

Estimated Capsule Lead Withdrawal<*) Insert<*) Fluence<*>

Capsule ID Status(dJ LocationCbl Factor< > 0 (EFPY I Year) (EFPY I Year) (x 10 19 n/cm2, E>l.OMeV) 1.1/1979 V 165° 1.61 Active NA 0.306 (EOC 1) 5.9 I 1987 u 65° 1.04 Active (EOC 6)

NA 0.914 14.8 I 1998 w 245° 1.16 Active (EOC 13)

NA 2.05 39 .1 / 2025<t)

X 285° 1.77 Active<t) NA 1.2o<t)

(estimated) z 305° 0.80 Active 16.2/2000 NA 1.52 16.2/2000 z 165° NA Active NA (Moved EOC 14) 1.52 59.4 I 2046<g>

z 165° 1.44 Active

~~( estimated)~~

~~NA 8.92(g)~~

T 55° 0.80 Stand-By(hl 16.2/2000 NA 1.52 16.2/2000 T 245° NA Stand-By(hl NA 1.52 (Moved EOC 14) 8.46 T 245° 1.13 Stand-By(hl NA NA (72 EFPY) y 9.17 295° 1.26 Stand-By(hl NA NA (72EFPY) 5.41 s 45° 0.73 Stand-By<hl NA NA (72 EFPY)

~~Notes (notes continued on next page):~~

~~(a) Withdrawal schedule meets requirements of ASTM E 185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, dated July I, 1982.~~

~~(b) See UFSAR Figure 5.4-5 for original capsule installation locations.~~

(c) Lead Factor is defined in ASTM El85-82 as the ratio of the neutron flux density at the location of the specimens in a surveillance capsule to the neutron flux density at the reactor pressure vessel inside surface at the peak fluence location. An entry of"NA" corresponds to a capsule relocation.

(d) Capsules with a withdrawal date intended to satisfy ASTM EI 85-8;l, GALL, GALL-SLR, and/or asset management objectives are designated Active. Capsules with no planned withdrawal dates, but which are maintained for contingencies, including further license renewal, are designated Standby.

(e) Surveillance capsule neutron fluence and EFPY estimates are take.n from Section 2 and consistent with WCAP-18015-NP (Reference 6). WCAP-18015-NP uses an NRC approved methodology, documented in WCAP-14040-A (Reference 3), that satisfies Regulatory Guide 1.190 (Reference 2). 50.3 EFPY corresponds to the estimated cumulative core bumup at the end of the 60-year license period. Historic capsule withdrawal dates are taken from UFSAR Table 5.4-2.

(f) Capsule withdrawal fluence is set equal to a projected 80-year peak clad/base metal fluence, i.e. 7.20 x 10 19 n/cm2 , projected to occur at 39.1 EFPY. The capsule should be withdrawn at the outage nearest to but following 39.1 EFPY of operation. The capsule is estimated to be withdrawn in 2025 (EOC 31 ). The withdrawal date is provided as an estimate only, and the EFPY should be treated as the official withdrawal schedule. The withdrawal dates are estimated based on an EFPY of 34.4 EFPY on Sept 8, 2019 (EOC 27). A capacity factor of95.4% is used based on 25 day outage every 18 months (1.5 year), or 100% -

~~.;;/ays ,~~

~~365 year *1.Syears to estimate the year of the outage following 39.1 EFPY.~~

~~Projected Date= Sept 8, 2019 + (Projected EFPY - 34.4)~~

365.25 days/ 95.4%

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Westinghouse Non-Proprietary Class 3 7-4 (g) In order to support asset management, capsule should be withdrawn at a fluence equal to a projected 100-year peak clad/base metal fluence, i.e. 8.92 x 10 19 n/cm2, projected to occur at 59.4 EFPY. The capsule should be withdrawn at the outage nearest to but following 59.4 EFPY of operation. The capsule is estimated to be withdrawn in 2046 (EOC 45). The withdrawal date is provided as an estimate only, and the EFPY should be treated as the official withdrawal schedule. The withdrawal dates are estimated based on Footnote (f) of this table.
(h) Capsules T, Y, and Sare available to satisfy potential fluence monitoring requirements during license renewal periods.
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Westinghouse Non-Proprietary Class 3 7-5 Table 7-2 North Anna Unit 2 Recommended Surveillance Capsule Withdrawal Schedule<a)
Estimated Capsule Lead Withdrawal<*J Insertion<*J Fluence<*J Capsule ID Status<dJ Location<bJ Factor<cJ (EFPY I Year) (EFPY I Year) (x 10 19 n/cm2, E>l.OMeV) 1.0 / 1982 V 165° 1.61 Active NA 0.286 (EOC 1) 6.2 I 1989 u 65° 1.17 Active (EOC 6)
NA 0.985 15.1/1999 w 245° 1.21 Active (EOC 13)
NA 2.08 39.3 I 2026<+/-)
X 285° 1.81 Active<f) NA 7.34(!)
(estimated) z 305° 0.85 Active 15.1 I 1999 NA 1.45 15.1/1999 z 165° NA Active NA (Moved EOC 13) 1.45 56.1 I 2043<sl z 165° 1.58 Active<g)
(estimated)
NA 9_15<s)
T 55° 0.85 Stand-By(h) 15.1/1999 NA 1.45 Stand-By(h) 15.1/1999 T 65° NA NA 1.45 (Moved EOC 13) 8.70 T 65° 1.15 Stand-By(hl NA NA (72EFPY) y 9.33 295° 1.27 Stand-By(hl NA NA (72 EFPY) 4.85 s 45° 0.67 Stand-By(hl NA NA (72 EFPY)
Notes (notes continued on next page):
(a) Withdrawal schedule meets requirements of ASTM E185-82, Standard Practice/or Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, dated July 1, 1982.
(b) See UFSAR Figure 5.4-5 for original capsule installation locations.
(c) Lead Factor is defined in ASTM E185-82 as the ratio of the neutron flux density at the location of the specimens in a surveillance capsule to the neutron flux density at the reactor pressure vessel inside surface at the peak fluence location. An entry of"NA" corresponds to a capsule relocation.
(d) Capsules with a withdrawal date intended to satisfy ASTM E 185-82, GALL, GALL-SLR, and/or asset management objectives are designated Active. Capsules with no planned withdrawal dates, but which are maintained for contingencies, including further license renewal, are designated Standby.
(e) Surveillance capsule neutron fluence and EFPY estimates are taken from Section 2 and consistent with WCAP-18015-NP (Reference 6). WCAP-18015-NP uses an NRC approved methodology, documented in WCAP-14040-A (Reference 3), that satisfies Regulatory Guide 1.190 (Reference 2). 52.3 EFPY corresponds to the estimated cumulative core burnup at the end of the 60-year license period. Historic capsule withdrawal dates are taken from UFSAR Table 5.4-3.
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Westinghouse Non-Proprietary Class 3 7-6 (f) Capsule withdrawal fluence is set equal to a projected 80-year peak clad/base metal fluence, i.e. 7.34 x 1019 n/cm2, projected to occur at 39.3 EFPY. The capsule should be withdrawn at the outage nearest to but following 39.3 EFPY of operation. The capsule is estimated to be withdrawn in 2026 (EOC 31 ). The withdrawal date is provided as an estimate only, and the EFPY should be treated as the official withdrawal schedule. The withdrawal dates are estimated based on an EFPY of 33.2 EFPY on March 3, 2019 (EOC 26). A capacity factor of 95.4% is used based on 25 day outage every 18 months (1.5 year), or 100% ~days , to estimate the year of the outage following 39.3 EFPY.
3 65 year *1.Syears Projected Date= March 3, 2019 + (Projected EFPY - 33.2)

365.25 days/ 95.4%

(g) In order to support asset management, capsule should be withdrawn at a fluence equal to a projected 100-year peak clad/base metal flucnce, i.e. 9.15 x 10 19 n/cm2, projected to occur at 56.l EFPY. The capsule should be withdrawn at the outage nearest to but following 56.1 EFPY of operation. The capsule is estimated to be withdrawn in 2043 (EOC 42). The withdrawal date is provided as an estimate only, and the EFPY should be treated as the official withdrawal schedule. The withdrawal dates are estimated based on footnote (f) of this table.
(h) Capsules T, Y, and S are available to satisfy potential fluence monitoring requirements during license renewal periods.
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Westinghouse Non-Proprietary Class 3 7-7 Table 7-3 Comparison of the North Anna Unit 1 Capsule Withdrawal Schedule with ASTM E185-82 and the GALL Reports Requirement for a 4 Capsule Withdrawal Supplemental Guidance

Schedule per ASTM E185-82 Table 1, from GALL and GALL-SLR Comments<a) 1 3 EFPYor NIA For Unit 1, Capsule V was withdrawn at EOC 1, with an exposure of At the time when the accumulated neutron fluence 1.1 EFPY and a fluence = 3.06 x 10 18 n/cm2

The maximum predicted of the capsule exceeds 5 x 1022 n/m2 ~RTNDT was 82°F.(b)

(5 x 10 18 n/cm 2), or at the time when the highest predicted ~RTNDT of all encapsulated materials is ASTM E185-82 is met because the capsule was withdrawn as soon as approximately 28°C (50°F), whichever comes first. practical after MTNDT was 50°F.
2 6EFPYor NIA For Unit 1, Capsule U was withdrawn at EOC 6, with an exposure of At the time when the accumulated neutron fluence 5.9 EFPY and a fluence = 9.14 x 10 18 n/cm2

The peak 1/4T fluence atEOL of the capsule corresponds to the approximate end- (32 EFPY, 3.38 x 10 19 n/cm2 ) was approximately 2.13 x 10 19 n/cm 2 .Cc) of-license (EOL) fluence at the reactor vessel 1/4T location, whichever comes first. _ ASTM El85-82 is met because the capsule was withdrawn at approximately 6 EFPY, which occurred prior to the capsule receiving a fluence equal to the peak reactor vessel fluence at the 1/4T location at end of license.

3 15 EFPYor NIA For Unit 1, Capsule W was withdrawn at EOC 13, with an exposure of Or at the time when the accumulated neutron 14.8 EFPY and a fluence = 2.05 x 10 19 n/cm 2

The clad/base metal fluence at fluence of the capsule corresponds to the EOL (32 EFPY) was approximately 3.38 x 10 19 n/cm2

approximate EOL fluence at the reactor vessel inner wall location, whichever comes first. ASTM E185-82 is met because the capsule was withdrawn at approximately 15 EFPY, which occurred prior to the capsule receiving a fluence equal to the peak reactor vessel fluence at end of license.

4 EOLor For Unit 1, it is recommended that Capsule X be withdrawn with an exposure Not less than once or greater than twice the peak Per the GALL, at least one capsule is of approximately 39.1 EFPY and a fluence approximately equal to EOL vessel fluence. This may be modified on the withdrawn and tested with a fluence of 7.20 x 10 19 n/cm2

Additionally, in order to support asset management, it is basis of previous tests. This capsule may be held between one and two times the recommended that Capsule Z be withdrawn with an exposure of without testing following withdrawal. 60-year peak reactor vessel wall approximately 59.4 EFPY and a fluence approximately equal to neutron fluence. 8.92 x 10 19 n/cm 2

The peak clad/base metal fluence projected at End-Of-License Extension (EOLE, 50.3 EFPY) is 5.13 x 10 19 n/cm2 and at SLR (72 Per the GALL-SLR, at least one EFPY) is 7.20 x 10 19 n/cm2 _(d) capsule is withdrawn and tested with a fluence of between one and two times ASTM E185-82, GALL, and GALL-SLR guidance are met because one the 80-year peak reactor vessel wall capsule would be withdrawn with fluence between one and two times the 60-neutron fluence. It is not acceptable to year peak reactor vessel wall neutron fluence and an additional capsule re-direct a 60-year capsule to also would be withdrawn with fluence between one and two times the 80-year fulfill t~is requirement. peak reactor vessel wall neutron fluence. It is noted that both Capsules X and Z would provide greater than 80-year data.

Notes contained on next page.
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Westinghouse Non-Proprietary Class 3 7-8 Notes:
(a) The peak clad/base metal fluence, capsule fluence values, cycle numbers, and EFPY values are taken from Table 2-1.
(b) The predicted ""RTNDT for Capsule Vis= CF* FF= 121.63°F

0.675. The CF ofl21.63°F is the maximum RG 1.99, Position 1.1 CF of the materials in the North Anna Unit 1 reactor vessel surveillance program (Lower Shell Forging and Intermediate to Lower Shell Circumferential Weld) as taken from Table 3-7. The FF of 0.675 is equal to

£(0-28 -0.IO'log (I)) = 0.306<0-28 -0.IO' log(O.J06ll, where 0.306 is the capsule fluence (in units ofl 0 19 n/cm2).
(c) The 1/4T fluence value was calculated from the surface fluence, the reactor vessel beltline thickness (7.677 inches), and equation f= fsurf

e*0-24 (x) from RG 1.99, Revision 2, where x = the depth into the vessel wall (inches).

(d) North Anna Unit 1 projected Capsule X fluence is set equal to the projected 80-year clad/base metal fluence and Capsule Z fluence is set equal to the projected 100-year clad/base metal fluence. The corresponding EFPY values were determined by interpolation.
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Westinghouse Non-Proprietary Class 3 7-9 Table 7-4 Comoanson of the North Anna Unit 2 Capsule Withdrawal Schedule with ASTM E185-82 and the GALL Reports Requirement for a 4 Capsule Withdrawal Supplemental Guidance

Schedule oer ASTM E185-82 Table 1, from GALL and GALL-SLR Comments<a) 3 EFPY or NIA For Unit 2, Capsule V was withdrawn at EOC 1, with an exposure of At the time when the accumulated neutron fluence 1.0 EFPY and a fluence = 2.86 x 10 18 n/cm2

The maximum predicted of the capsule exceeds 5 x 1022 n/m2 L'.RTNDT was 54°F_(b)

(5 x 10 18 n/cm 2), or at the time when the highest predicted L'.RTNoT of all encapsulated materials is ASTM E185-82 is met because the capsule was withdrawn t.RTNDT was approximately 28°C (50°F), whichever comes first. approximately 50°F.
2 6EFPYor NIA For Unit 2, Capsule U was withdrawn at EOC 6, with an exposure of At the time when the accumulated neutron fluence 6.2 EFPY and a fluence = 9.85 x 10 18 n/cm 2

The peak 1/4T fluence at EOL of the capsule corresponds to the approximate EOL (32 EFPY, 3.31 x 10 19 n/cm2) was approximately 2.09 x 10 19 n/cm2 _(c) fluence at the reactor vessel 1/4T location, whichever comes first. ASTM El 85-82 is met because the capsule was withdrawn at approximately 6 EFPY, which occurred prior to the capsule receiving a fluence equal to the peak reactor vessel fluence at the 1/4T location at end of license.

3 15 EFPYor NIA For Unit 2, Capsule W was withdrawn at EOC 13, with an exposure of Or at the time when the accumulated neutron 15.1 EFPY and a fluence = 2.08 x 10 19 n/cm2

The peak clad/base metal fluence of the capsule corresponds to the fluence at EOL (32 EFPY) was approximately 3.31 x 10 19 n/cm2

approximate EOL fluence at the reactor vessel inner wall location, whichever comes first. ASTM E185-82 is met because the capsule was withdrawn at approximately 15 EFPY, which occurred prior to the capsule receiving a fluence equal to the peak reactor vessel fluence at end of license.

4 EOLor Per the GALL, at least one capsule is For Unit 2, it is recommended that Capsule X be withdrawn with an exposure Not less than once or greater than twice the peak withdrawn and tested with a fluence of of approximately 39.3 EFPY and a fluence approximately equal to EOL vessel fluence. This may be modified on the between one and two times the 60- 7.34 x 10 19 n/cm2

Additionally, in order to support asset management, it is basis of previous tests. This capsule may be held year peak reactor vessel wall neutron recommended that Capsule Z be withdrawn with an exposure of without testing following withdrawal. fluence. approximately 56.1 EFPY and a fluence approximately equal to 9.15 x 10 19 n/cm 2.The peak clad/base metal fluence projected at EOLE (52.3 EFPY) is 5.36 x 10 19 n/cm2 and at SLR (72 EFPY) is Per the GALL-SLR, at least one 7.34 x 10 19 n/cm2 .(d) capsule is withdrawn and tested with a fluence of between one and two times ASTM El85-82, GALL, and GALL-SLR guidance are met because a the 80-year peak reactor vessel wall capsule would be withdrawn with a fluence between one and two times the neutron fluence. It is not acceptable to 60-year peak reactor vessel wall neutron fluence and an additional capsule re-direct a 60-year capsule to also would be withdrawn with fluence between one and two times the 80-year fulfill this requirement.

peak reactor vessel wall neutron fluence. It is noted that both Capsules X and Z would provide greater than 80-year data.
Notes contained on next page.
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Westinghouse Non-Proprietary Class 3 7-10 Notes:
(a) The peak clad/base metal fluence, capsule fluence values, cycles numbers, and EFPY values are taken from Table 2-2.
(b) The predicted IIB.TNDT for Capsule Vis= CF* FF= 82.4°F

0.658. The CF of 82.4°F is the maximum RG 1.99, Position 1.1 CF of the materials in the North Anna Unit 2 reactor vessel surveillance program (Intermediate Shell Forging and Intermediate to Lower Shell Circumferential Weld) as taken from Table 3-8. The FF of0.658 is equal to f(0-2s-o.io*Iog(t)) = 0.286<0 0-10

Iog(o. 286 >>, where 0.286 is the capsule fluence (in units of 10 19 n/cm2).

(c) The I/4T fluence value was calculated from the surface fluence, the reactor vessel beltline thickness (7.677 inches), and equation f= fsurf

e-0 -24 (x) from RG 1.99, Revision 2, where x = the depth into the vessel wall (inches).

(d) North Anna Unit 2 projected Capsule X fluence is set equal to the projected 80-year clad/base metal fluence and Capsule Z fluence is set equal to the projected 100-year clad/base metal fluence. The corresponding EFPY values were determined by interpolation.
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Westinghouse Non-Proprietary Class 3 8-1 8 REFERENCES

1. Code of Federal Regulations, 10 CFR Part 54.3, "Definitions," U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 72, dated August 28, 2007.

2. 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. [Agencywide Documents Access and Management System (ADAMS) Accession Number ML010890301]

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.

4. RSICC Computer Code Collection CCC-650, "DOORS 3.2, One, Two-, and Three- Dimensional Discrete Ordinates Neutron/Photon Transport Code System," April 1998.

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," March 1996.

6. Westinghouse Report WCAP-18015-NP, Revision 2 "Extended Beltline Pressure Vessel Fluence Evaluations Applicable to North Anna 1 & 2," September 2018.

7. Pressurized Water Reactor (PWR) Owners Group (PWROG) Report PWROG-18005-NP, Revision 2, "Determination ofUnirradiated RTNnT and Upper-Shelf Energy Values of the North Anna Units 1 and 2 Reactor Vessel Materials," September 2019.

8. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Chapter 5 of LWR Edition, Branch Technical Position 5-3, "Fracture Toughness Requirements," Revision 2, U.S. Nuclear Regulatory Commission, March 2007. [ADAMS Accession Number ML070850035]

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

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

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

11. 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]

12. PWROG Report PWROG-17090-NP-A, Revision 0, "Generic Rotterdam Forging and Weld Initial Upper Shelf Energy Determination," January 2020. [ADAMS Accession Number ML20024E238]

13. BAW-2356, "Analysis of Capsule W Virginia Power North Anna Unit No. 1 Nuclear Power Plant, Reactor Vessel Material Surveillance Program," September 1999.

14. BA W-2376, "Analysis of Capsule W Virginia Power North Anna Unit No. 2 Nuclear Power Plant, Reactor Vessel Material Surveillance Program," August 2000.

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Westinghouse Non-Proprietary Class 3 8-2

15. Code of Federal Regulations IO CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, dated December 19, 1995, effective January 18, 1996.

16. Code of Federal Regulations, 10 CFR 50.61a, "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," U.S. Nuclear Regulatory Commission, Federal Register, Volume 75, No. 1, dated January 4, 2010, with corrections dated February 3, 2010 (No. 22), March 8, 2010 (No. 44), and November 26, 2010 (No. 227).

17. Westinghouse Report WCAP-17539-NP, Revision 0, "Sequoyah Units 1 and 2 Time-Limited Aging Analysis on Reactor Vessel Integrity," March 2012. [ADAMS Accession Number MLI 3032A253]

18. 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."

19. Westinghouse Report WCAP-18363-NP, Revision 1, "North Anna Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," March 2020.

20. Westinghouse Report WCAP-15112, Revision 2, "North Anna Units 1 and 2 WOG Reactor Vessel 60-Year Evaluation Minigroup Heatup and Cooldown Limit Curves for Normal Operation," March 2001.

21. ASTM E185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society for Testing and Materials, 1982.

22. Code of Federal Regulations IO CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, December 19, 1995.

23. NUREG-1801, Revision 2, "Generic Aging Lessons Learned (GALL) Report," December 2010, U.S. Nuclear Regulatory Commission. [ADAMS Accession Number ML103490041]

24. NUREG-2191, Volume 2, "Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report," July 2017, U.S. Nuclear Regulatory Commission. [ADAMS Accession Number ML17187A204]

25. Dominion Calculation SM-1008, Revision 0, Addendum M.

26. NRC Letter "North Anna Power Station, Unit Nos. 1 and 2, Issuance of Amendments to Change the Reactor Coolant System Pressure and Temperature Limits Regarding Vacuum Fill Operations (TAC Nos. MF4707 and MF4708)," dated July 27, 2015. [ADAMS Accession Number MLJ5187A424]

27. WCAP-8771, Revision 0, "Virginia Electric and Power Company North' Anna Unit No. 1 Reactor Vessel Radiation Surveillance Program," September 1976.

28. WCAP-8772, Revision 0, "Virginia Electric and Power Company North Anna Unit No. 2 Reactor Vessel Radiation Surveillance Program," November 1976.

29. ASTM El85-73, "Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels," American Society for Testing and Materials, 1973.

30. North Anna Power Station (NAPS) Updated Final Safety Analysis Report (UFSAR), Amendment No. 54, September 2018.

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Westinghouse Non-Proprietary Class 3 A-1 APPENDIX A CREDIBILITY EVALUATION OF THE NORTH ANNA UNITS 1 AND 2 SURVEILLANCE PROGRAMS Regulatory Guide 1.99, Revision 2 (Reference A-1) describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Positions 2.1 and 2.2 of Regulatory Guide 1.99, Revision 2, describe the method for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data. The methods of Positions 2.1 and 2.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.
To date there have been three surveillance capsules removed and tested from each of the North Anna Units 1 and 2 reactor vessels. The Unit 1 forging and weld surveillance data are judged to be non-credible based on the five criteria in Regulatory Guide 1.99, Revision 2. The Unit 2 weld surveillance data are judged to be credible based on the five criteria in Regulatory Guide 1.99, Revision 2; however, the Unit 2 forging surveillance data are judged to be non-credible. Appendix C contains an explanation of the North Anna licensing basis for the use of credible/ non-credible surveillance data.
Table A-1 reviews the five criteria in Regulatory Guide 1.99, Revision 2. The following subsections evaluate each of these five criteria for North Anna Units 1 and 2 in order to determine the credibility of the surveillance data for use in neutron radiation embrittlement calculations.
It should be noted that this report also uses surveillance data from Sequoyah Units 1 and 2 surveillance programs. The credibility conclusions for the surveillance data from these programs are contained in WCAP-17539-NP (Reference A-2), Appendix A. The conclusions in WCAP-17539-NP will not be readdressed here as the use of surveillance data in this report does not affect the credibility conclusions.
TableA-1 Regulatory Guide 1.99, Revision 2, Credibility Criteria Criterion Description No.
Materials in the capsules should be those judged most likely to be controlling with regard to 1
radiation embrittlement.
Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated 2 conditions should be small enough to permit the determination of the 30 ft-lbs temperature and upper-shelf ener2:v unambiguously.
When there are two or more sets of surveillance data from one reactor, the scatter of ~RTNOT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be
\
\
less than 28°F for welds and l 7°F for base metal. Even if the fluence range is large (two or 3 , 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 inASTM E185-82.
The irradiation temperature of the Charpy specimens in the capsule should match the vessel 4
wall temperature at the cladding/base metal interface within +/- 25°F.
The surveillance data for the correlation monitor material in the capsule should fall within the 5
scatter band of the database for that material.
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Westinghouse Non-Proprietary Class 3 A-2 A.1 NORTH ANNA UNIT 1 CREDIBILITY EVALUATION Criterion 1: Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement.
The North Anna Unit I reactor vessel consists of the following beltline region materials, which likely would have been considered at the time the surveillance program was designed and licensed:
a) Upper Shell Forging b) Intermediate Shell Forging c) Lower Shell Forging d) Intermediate to Lower Shell Circumferential Weld e) Upper to Intermediate Shell Circumferential Weld At the time that the North Anna Unit 1 surveillance program was designed and licensed, the materials selected for use in the North Anna Unit 1 surveillance program (Lower Shell Forging 03 and the Intermediate to Lower Shell Circumferential Weld) were those judged to be most likely controlling with regard to radiation embrittlement according to the accepted methodology. These materials remain limiting with respect to fluence and ART. Thus, the North Anna Unit I surveillance program meets the intent of this criterion.
Criterion 2: Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-lbs temperature and upper-shelf energy unambiguously.
The surveillance capsule analysis report, BAW-2356 (Reference A-3), which supports the Position 2.1 chemistry factor calculations, was reviewed and it was determined that this criterion is met.
Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of Llli.TNoT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28°F for welds and l 7°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 inASTM E185-82.
The functional form of the least squares method as described in Regµlatory Position 2.1 will be utilized to determine*,
a best-fit line for this data and to determine if the scatter ofI these ~RTNoT values about this line is less than 28°F for welds and less than l 7°F for the forgings.
Following is the calculation of the best-fit line as described in Regulatory Position 2.1 of Regulatory Guide 1.99, Revision 2. In addition, the recommended NRC methods for determining credibility will be followed.
The NRC methods were presented to the industry at a meeting held by the NRC on February 12 and 13, 1998 (Reference A-4). At this meeting the NRC presented five cases. Of the five cases, Case 1
("Surveillance data available from plant but no other source") most closely represents the situation for the North Anna Unit I surveillance forging and weld material.
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Westinghouse Non-Proprietary Class 3 A-3 Case 1: Lower Shell Forging 03 and Weld Heat # 25531 Following the NRC Case 1 guidelines, the North Anna Unit 1 surveillance forging and weld metal (Heat#
25531) will be evaluated using the North Anna Unit 1 data. Only North Anna Unit 1 data is being considered; therefore, no temperature adjustment is required.
The scatter of 11RTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 is presented in Table A.1-1.
TableA.1-1 North Anna Unit 1 Surveillance Capsule Data Scatter about the Best-Fit Line Using All Available Surveillance Data
<17°F CF<a> Capsule Measured Predicted Scatter ARTNDT(c)
(Base Metal)
Material Capsule (Slopebest-fit) Fluence FF ARTNDT(b) ARTNDT
<28°F (OF) (x 10 19 n/cm 2 ) (°F) (OF) (OF)
(Weld)
V 81.68 0.306 0.675 51 55.2 4.2 Yes Lower Shell Forging 03 u 81.68 0.914 0.975 116 79.6 36.4 No (Tangential) w 81.68 2.05 1.196 93 97.7 4.7 Yes V 81.68 0.306 0.675 29 55.2 26.2 No Lower Shell Forging 03 u 81.68 0.914 0.975 72 79.6 7.6 Yes (Axial) w 81.68 2.05 1.196 96 97.7 1.7 Yes V 67.53 0.306 0.675 88 45.6 42.4 No Surveillance Weld Material u 67.53 0.914 0.975 30 65.8 35.8 No (Heat#25531) w 67.53 2.05 1.196 86 80.7 5.3 Yes Notes:
(a) Since the Position 2.1 CFs in Table 3-4 did not consider chemistry or temperature adjustments, the interim CFs are equal to the Position 2.1 CFs calculated in Table 3-4 ofreport.
(b) i:IB.TNDr values are the measured 30 ft-lbs shift values taken from BA W-2356 (Reference A-3).
(c) Scatter llRTNDr = Absolute Value [Predicted llRTNDr- Measured llRTNDr].
For North Anna (see Appendix C), if one or more of the surveillance data fall outside+/- lcr scatter band ot' the Position 2.1 CF trend line then the data is considered non~'credible. Table A.1-1 indicates that only four of the six surveillance data points fall inside the +/- 1cr of l 7°F scatter band for surveillance base metals. Therefore, the forging data is deemed "non-credible" per the third criterion.
Table A.1-1 indicates that two of the three surveillance data points fall outside the +/- 1cr of 28°F scatter band for surveillance weld materials. Therefore, the surveillance weld data is deemed "non-credible" per the third criterion.
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Westinghouse Non-Proprietary Class 3 A-4 Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +/- 25°F.
The capsule specimens are located in the reactor between the thermal shield and the vessel wall and are positioned opposite the center of the core. The test capsules are in guide tubes attached to the thermal shield. The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions such that the temperatures will not differ by more than 25°F. Hence, this criterion is met.
Criterion 5: The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the database for that material.
The North Anna Unit 1 surveillance program does not contain correlation monitor material. Hence, this criterion is not applicable to the North Anna Unit 1 surveillance program.
CONCLUSION:
Based on the preceding responses to the five criteria of Regulatory Guide 1.99, Revision 2, Section B, the Lower Shell Forging 03 and weld Heat# 25531 surveillance data are deemed non-credible.
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Westinghouse Non-Proprietary Class 3 A-5 A.2 NORTH ANNA UNIT 2 CREDIBILITY EVALUATION Criterion 1: Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement.
The North Anna Unit 2 reactor vessel consists of the following beltline region materials, which likely would have been considered at the time the surveillance program was designed and licensed:
a) Upper Shell Forging b) Intermediate Shell Forging c) Lower Shell Forging d) Intermediate to Lower Shell Circumferential Weld
. e) Upper to Intermediate Shell Circumferential Weld At the time that the North Anna Unit 2 surveillance program was designed and licensed, the materials selected for use in the North Anna Unit 2 surveillance program (Intermediate Shell Forging 04 and the Intermediate to Lower Shell Circumferential Weld) were those judged to be most likely controlling with regard to radiation embrittlement according to the accepted methodology. These materials remain limiting with respect to ART. Thus, the North Anna Unit 2 surveillance program meets the intent of this criterion.
Criterion 2: Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-lbs temperature and upper-shelf energy unambiguously.
The surveillance capsule analysis report, BAW-2376 (Reference A-5), which supports the Position 2.1 chemistry factor calculations, was reviewed and it was determined that this criterion is met.
Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of LlRTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28°F for welds and l 7°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 inASTM E185-82.
The functional form of the least squares method as described in Regulatory Position 2.1 will be utilized to determine a best-fit line for this data and to determine if the scatter of these LlRTNDT values about this line is less than 28°F for welds and less than l 7°F for the forgings.
Following is the calculation of the best-fit line as described in Regulatory Position 2.1 of Regulatory Guide 1.99, Revision 2. In addition, the recommended NRC methods for determining credibility will be followed.
The NRC methods were presented to the industry at a meeting held by the NRC on February 12 and 13, 1998 (Reference A-4). At this meeting the NRC presented five cases. Of the five cases, Case 1
("Surveillance data available from plant but no other source") most closely represents the situation for the North Anna Unit 2 surveillance forging and weld material.
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Westinghouse Non-Proprietary Class 3 A-6 Case 1: Intermediate Shell Forging 04 and Weld Heat# 716126 Following the NRC Case 1 guidelines, the North Anna Unit 2 surveillance forging and weld metal (Heat#
716126) will be evaluated using the North Anna Unit 2 data. Only North Anna Unit 2 data is being considered; therefore, no temperature adjustment is required.
The scatter of ARTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 is presented in Table A.2-1.
TableA.2-1 North Anna Unit 2 Surveillance Capsule Data Scatter about the Best-Fit Line Using All Available Surveillance Data CF(al <17°F Capsule Measured Predicted Scatter ARTNDT(b) ARTNDT(c) (Base Metal)
Material Capsule (Slopebest-lit) Fluence FF ARTNDT
<28°F
{°F) (x 10 19 n/cm2) {°F) {°F) (OF)
(Weld)
V 53.44 0.286 0.658 19 35.2 16.2 Yes Intermediate Shell Forging 04 u 53.44 0.985 0.996 33 53.2 20.2 No (Tangential) w 53.44 2.08 1.199 86 64.1 21.9 No V 53.44 0.286 0.658 21 35.2 14.2 Yes Intermediate Shell Forging 04 u 53.44 0.985 0.996 66 53.2 12.8 Yes (Axial) w 53.44 2.08 1.199 65 64.1 0.9 Yes V 26.61 0.286 0.658 18 17.5 0.5 Yes Surveillance Weld Material u 26.61 0.985 0.996 8 26.5 18.5 Yes (Heat# 716126) w 26.61 2.08 1.199 47 31.9 15.1 Yes Notes:
(a) Since the Position 2.1 CFs in Table 3-6 did not consider chemistry or temperature adjustments the interim CFs are equal to the Position 2.1 CFs calculated in Table 3-6.
(b) ilRTNDT values are the measured 30 ft-lbs shift values taken from BAW-2376 (Reference A-5).
(c) Scatter t-.RTNDT = Absolute Value [Predicted t-.RTl\'DT-Measured ilRTNDT],
For North Anna (see Appendix C), if one or more of the surveillance data fall outside+/- lcr scatter band of the Position 2.1 CF trend line then the data is considered non-credible. Table A.2-1 indicates that only four of the six surveillance data points fall inside the+/- lcr of 17°F scatter band for surveillance base metals. Therefore, the forging data is deemed "non-credible" per the third criterion.
Table A.2-1 indicates that all three of the three surveillance data points for the surveillance weld materials fall inside the+/- lcr of 28°F scatter band. Therefore, the surveillance weld data is deemed "credible" per the third criterion.
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Westinghouse Non-Proprietary Class 3 A-7 Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +/- 25°F.
The capsule specimens are located in the reactor between the thermal shield and the vessel wall and are positioned opposite the center of the core. The test capsules are in guide tubes attached to the thermal shield. The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions such that the temperatures will not differ by more thap. 25°F. Hence, this criterion is met.
Criterion 5: The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the database for that material.
The North Anna Unit 2 surveillance program does not contain correlation monitor material. Hence, this criterion is not applicable to the North Anna Unit 2 surveillance program.
CONCLUSION:
Based on the preceding responses to the criteria of Regulatory Guide 1.99,

Revision 2, Section B, the Intermediate Shell Forging 04 surveillance data are deemed non-credible; however, the weld Heat# 716126 surveillance data are deemed credible.

A.3 REFERENCES A-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.
[ADAMS Accession Number ML003740284]
A-2. Westinghouse Report WCAP-17539-NP, Revision 0, "Sequoyah Units 1 and 2 Time-Limited Aging Analysis on Reactor Vessel Integrity," March 2012. [ADAMS Accession Number ML13032A253]
A-3. BAW-2356, "Analysis of Capsule W Virginia Power North Anna Unit No. 1 Nuclear Power Plant, Reactor Vessel Material Surveillance Program," September 1999.
A-4. K. Wichman, M. Mitchell, and A. Hiser, US NRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, "NRC/Industry Workshop on RPV Integrity Issues," February 12, 1998.
[ADAMS Accession Number MLJ 10070570]
A-5. BA W-2376, "Analysis of Capsule W Virginia Power North Anna Unit No. 2 Nuclear Power Plant, Reactor Vessel Material Surveillance Program," August 2000.
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Westinghouse Non-Proprietary Class 3 B-1 APPENDIXB EMERGENCY RESPONSE GUIDELINES 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 (Reference B-1 ). Generic categories of limits were developed for the guidelines based on the limiting inside surface RTNnT- These generic categories were conservatively generated for the Westinghouse Owners Group (WOG) to be applicable to all Westinghouse plants.
The highest value of RTNnT for which the generic category ERG limits were developed is 250°F for a longitudinal flaw and 300°F for a circumferential flaw. Therefore, if the limiting vessel material has an RTNDT that exceeds 250°F for a longitudinal flaw or 300°F for a circumferential flaw, plant-specific ERG P-T limits must be developed.
The ERG category is determined by the magnitude of the limiting RTNDT value, which is calculated the same way as the RTPTS values are calculated in Section 4 of this report. The material with the highest RTNDT defines the limiting material, which for North Anna Unit 1 is the Lower Shell Forging 03 (Position 2.1) and for North Anna Unit 2 is the Lower Shell Forging 03 (Position 1.1). Table B-1 identifies ERG category limits and the limiting material RTNDT values at 72 EFPY for North Anna Units 1 and 2.
Table B-1 Evaluation of North Anna Units 1 and 2 ERG Limit Category ERG Pressure-Temperature Limits (Reference B-1)
Applicable RTNoT Value(a) ERG P-T Limit Category RTNDT < 200°F Category I 200°F < RTNDT < 250°F Category II 250°F < RTNDT < 300°F Category III b Limiting RTNoT Value Limiting Reactor Vessel Material RTNoT Value@72 EFPY North Anna Unit 1 186.9°F(bl Lower Shell Forging 03 (Position 2.1)
North Anna Unit 2 212.2°F(c)
Lower Shell Forging 03 (Position 1.1)
Notes:
(a) Longitudinally oriented flaws are applicable only up to 250°F; circumferentially orient~d flaws are applicable up to 300°F.
(b) Value taken from Table 4-1. Note that the RTPTs value calculated for Unit 1 Lower Shell Forging 03 per Regulatory Guide 1.99, Revision 2, Position 1.1 is.higher. The use of the lesser of the Regulatory Guide 1.99, Revision 2, Position 1.1 and 2.1 CFs with non-credible data and a full margin term is justified since none of the surveillance data are more than two times sigma-delta above the Position 1.1 CF trend line. This evaluation is documented in Appendix C.
(c) Value taken from 4-2. Note that the RTPTs value calculated for Intermediate Shell Forging 04 per Regulatory Guide 1.99, Revision 2, Position 1.1 is higher. The use of the lesser of the Regulatory Guide 1.99, Revision 2, Position 1.1 and 2.1 CFs with non-credible data and a full margin term is justified since none of the surveillance data are more than two times sigma-delta above the Position 1. 1 CF trend line. This evaluation is documented in Appendix C.
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Westinghouse Non-Proprietary Class 3 B-2 Per the ERG limit guidance document (Reference B-1 ), 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.
Conclusion of ERG P-T Limit Categorization Per Table B-1, the limiting materials for North Anna Units 1 and 2 have RTNoT values less than 25O°F through 72 EFPY. Therefore, North Anna Units 1 and 2 are limited to ERG Category II through SLR (72 EFPY) and plant-specific ERG P-T limits do NOT need to be developed.
North Anna Unit 1 Lower Shell Forging 03 RTPTS will NOT exceed 2OO°F. Therefore, North Anna Unit 1 will remain in Category I through SLR (72 EFPY).
North Anna Unit 2 Lower Shell Forging 03 RTPTS will exceed 2OO°F, thus the plant's ERG Category changes from Category I to II, when the fluence exceeds 3.78 x 10 19 n/cm2 (E > 1.0 MeV). Table 2-4 in this report provides data points at 28.1 EFPY and 52.3 EFPY that can be used to determine when the ERG Category will change.

Lower Shell Forging 03 fluence@ 28.1 EFPY = 2.92 x 10 19 n/cm2

Lqwer Shell Forging 03 fluence@ 52.3 EFPY = 5.36 x 10 19 n/cm2 Interpolating between the below data points yields:.

3 70 2 92 EFPYRT- PTS=200°F = 28.1 + 5.36-2.92

- * * (52.3 - 28.1)

EFPYRT_PTS=20Q°F = 36.6 EFPY Therefore, North Anna Unit 2 must switch from ERG Category I to Category II prior to the cycle in which 36.6 EFPY is reached. Unit 2 will remain in ERG Category II through SLR (72 EFPY).
B.1 REFERENCES B-1. Westinghouse Owners G~oup Document HFO4BG, "Background Information for Westipghouse Owners Group Emergency Response Guidelines, Critical Safety Function Status Tree, F-O.4 Integrity, HP/LP-Rev. 3," March 2014.
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Westinghouse Non-Proprietary Class 3 C-1 APPENDIX C NORTH ANNA UNITS 1 AND 2 LICENSING BASIS FOR DETERMINING CHEMISTRY FACTOR WHEN SURVEILLANCE DATA IS AVAILABLE Regulatory Guide (RG) 1.99, Revision 2 (Reference C-1), indicates the Position 2.1 CFs may be used with a reduced margin term whenever the surveillance data has been deemed credible using the methodology described in Appendix A. Regulatory Guide 1.99, Revision 2 is less prescriptive if the surveillance data is deemed non-credible. However, additional guidance can be drawn from 10 CFR 50.61 (Reference C-2) and the NRC Generic Letter (GL) 92-01 Guidance (Reference C-3).
10 CFR 50.61 (c)(2) states that "licensees shall consider plant specific information that could affect the level of embrittlement". The plant specific information referred to in 10 CFR 50.61(c)(2) is data derived from reactor vessel materials surveillance programs that must be considered in the determination of L1RTNDT for the beltline material. GL 92-01 Guidance, published by NRC on February 12, 1998, describes the treatment of surveillance data for application to the corresponding reactor vessel beltline material, including: how to apply data from different data sources; how to correct surveillance data (L1RTNDT) values for differences in irradiation temperature and chemical composition relative to the reactor vessel beltline material being evaluated; how to evaluate the scatter of L1RTNDT data around a best fit Position 2.1 CF L1RTNDT trend line to determine surveillance data credibility; and how to compare surveillance data against a Position 1.1 CF trend line based on mean surveillance material chemical compositions to determine conservatism of the use of a Position 1.1 CF. As described in the NRC GL 92-01 Guidance, individual surveillance data points are not to be discarded on the basis of their deviation from a best fit Position 2.1 CF trend line alone; there must also be a recorded deficiency or a physical basis for classifying the data point as atypical. The same logic would apply to not discarding a data set.
When surveillance data is deemed non-credible per RG 1.99, NRC GL 92-01 Guidance considers the surveillance data to still be applicable for characterizing the beltline material through the direction of its use with a full margin term (cr,;) to establish the Position 2.1 RTNoT when the Position 1.1 CF is concluded to be non-conservative based on that same data (See case 3 in the NRC GL 92-01 Guidance): Dominion Energy proposed and licensed logically consistent application of surveillance data concluded to be non-credible due only to data scatter for both non-conservative and conservative RG 1.99, Position 1.1 CFs.
The Dominion licensing position is as follows:

a. The greater of the RG 1.99 Revision 2 Position 1.1 CF and 2.1 CF is used with a full margin term ( cr,; = l 7°F for base metal and cr,; = 28°F for welds) for evaluation of the reactor vessel beltline material whep. one or more of the surveillance data fall outside of the Position 2.1 CF trend line by more than one times cr,; (data is non-credible), and one or more ~f the surveillance data fall more than two times cr,; above the Position 1.1 CF trend line (Position

1. 1 CF is non-conservative)

b. The lesser of the RG 1.99 Revision 2 Position 1.1. and 2.1 CFs is used with a full margin term (cr,; = l 7°F for base metal and cr,; = 28°F for welds) for evaluation of the reactor vessel beltline material when one or more of the surveillance data fall outside of the Position 2.1 CF trend line by more than one times cr,; (data is non-credible), and none of the surveillance data fall more than two times cr,; above the Position 1.1 CF trend line (Position 1.1 CF is conservative).

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Westinghouse Non-Proprietary Class 3 C-2 Table C-1 contains the list of correspondences applicable to the submittal and approval of this licensing basis for North Anna Units 1 and 2. Appendixes C. l and C.2 evaluate whether the RG 1.99, Position 1.1 CFs are conservative based on the above licensing basis and the North Anna Units I and 2 surveillance data.
Table C-1 North Anna Units 1 and 2 Chemistry Factor Licensing Basis History Subject Content Relevant to North Anna Units 1 and 2 Reference Date Document(s) Chemistry Factor Licensing Basis Number(s)
Closure of the NRC review of the North Anna Units I and 2 GL 92-01 response.
NRC Letter This letter notes that for material with non-credible (Incoming June 23, surveillance but with all measured ARTNOT data points C-4 Correspondence No. 1999 below the RG 1.99, Position 1.1 ARTNOT predictions +
99-361) 2cr~, the RG 1.99, Position 2.1 CFs were used with a fuil margin value to calculate RTPTS*
Response to the NRC on discrepancies between the Reactor Vessel Integrity Database (RVID) and previously provided data.
VEPCO Letter Serial Sept. 1, This letter reiterates the use of non-credible surveillance C-5 No.99-361 1999 data with a full margin value when all measured ARTNOT data points are below the RG 1.99, Position 1.1 ARTNOT predictions + 2cr6.
Evaluation of the material properties based on the results from North Anna Unit 1, Capsule W.

VEPCO Letter Serial This letter reiterates the use of non-credible surveillance Nov. 19, C-6 No. 99-452A data with a full margin value when all measured ARTNOT 1999 data points are below the RG 1.99, Position 1.1 ARTNDT predictions+ 2cr6. It also includes portions of SM~1008.

Submittal of the 32.3 EFPY (Unit 1) and 34.3 EFPY (Unit 2) heatup and cooldown curves and Low Temperature Overpressure Protection System (LTOPS) setpoints.
VEPCO Letter Serial June 22, C-7 No.00-306 This letter includes a detailed evaluation based on the 2000 latest reactor vessel materials surveillance data, i.e. North Anna Unit 1, Capsule W. The North Anna Unit 1, Capsule '
W report, BAW-2356, is also attached.
VEPCO Letters Nos: Corrections and supplements to the above cited letters. C-8, C-9,01-020, 0 l-020A, No changes were made to the material property basis in Various C-10, C-11, 0l-020B,01-168, these letters. & C-12 0l-168A Revises the RVID for North Anna Unit 2 based on the Sequoyah Unit 2 surveillance data.
This letter demonstrates that conservatism or non-VEPCO Letter April 27, conservatism of the RG 1.99 ARTNOT prediction is C-13 No.01-262 2001 determined by whether all measured ARTNOT data points are below the RG 1.99, Position 1.1 LiRTNOT predictions +
2a6.
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Westinghouse Non-Proprietary Class 3 C-3 Table C-1 North Anna Units 1 and 2 Chemistry Factor Licensing Basis History Subject Content Relevant to North Anna Units 1 and 2 Reference Date Document(s) Chemistry Factor Licensing Basis Number(s)
Safety Evaluation (SE) for the 32.3 EFPY (Unit 1) and 34.3 EFPY (Unit 2) heatup and cooldown curves and NRC Letter LTOPS setpoints.
(Incoming May 2, Reiterates that for material with non-credible surveillance C-14 Correspondence No. 2001 data but all measured ~RTNDT data points below the RG 01-293) 1.99, Position 1.1 ~RTNDT predictions+ 2cr6., the RG 1.99, Position 2.1 CFs were used with a full margin value.
VEPCO Letter License Renewal Submittal May 29, C-15 No.01-282 2001 License Renewal RAI Response on RV embrittlement issues.
VEPCO Letter Oct. 15, C-16 No.02-601 Reiterates that conservatism is defined as below RG 2002 Position 1.1 ~RTNDT predictions +2cr6..
Submittal of the 50.3 EFPY (Unit 1) and 52.3 EFPY VEPCO Letter (Unit 2) heatup and cooldown curves and LTOPS July 1, C-17 No.04-380 setpoints. 2004 VEPCO Letter Provides RAI responses related to the Oct. 28, 50.3 EFPY/52.3 EFPY P-T curves submittal. C-18 No. 04-380A 2004 VEPCO Letter Editorial correction for the 50.3 EFPY/52.3 EFPY P-T Nov. 16, curves submittal. C-19 No. 04-380B 2004 NRC Letter (Incoming SE for the 50.3 EFPY (Unit 1) and 52.3 EFPY (Unit 2) July 8, heatup and cooldown curves and LTOPS setpoints. C-20 Correspondence 2005 No.05-460)
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Westinghouse Non-Proprietary Class 3 C-4 C.1 Regulatory Guide 1.99, Revision 2, Position 1.1 Conservativism Evaluation for North Anna Unit 1 Surveillance Data The purpose of this evaluation is to detennine if the North Anna Unit 1 non-credible surveillance data is less than 2crt. above of the RG 1.99, Position 1.1 prediction, thereby, detennining whether the RG 1.99, Position 1.1 CF is conservative. This evaluation is perfonned in Table C-2.
Table C-2 Conservatism Check for Position 1.1 for Non-Credible North Anna Unit 1 Surveillance Data CFC*> Capsule Measured Predicted Scatter Material Capsule (Pos. 1.1) FluenceChl FFCc> ARTNDT(b) ARTNDT(d) ARTNDr<*> <2aCf)
(OF) (x 10 19 n/cm2 ) (OF) (OF) (OF)
Lower Shell V 121.63 0.306 0.675 51 82.1 -31.1 Yes Forging 03 u 121.63 0.914 0.975 116 118.6 -2.6 Yes (Tangential) w. 121.63 2.05 1.196 145.4 -52.4 Yes 93 Lower Shell V 121.63 0.306 0.675 29 82.1 -53.1 Yes Forging 03 u 121.63 0.914 0.975 72 118.6 -46.6 Yes (Axial) w 121.63 2.05 1.196 96 145.4 -49.4 Yes Surveillance Weld V 56.22 0.306 0.675 88 38.0 50.0 Yes Material u 56.22 0.914 0.975 30 54.8 -24.8 Yes (Heat#25531) w 56.22 2.05 1.196 86 67.2 18.8 Yes Notes:
(a) CF values are taken from Table 3-7.
(b) Fluence and Measured iill.TNDT values are taken from Table 3-4.
(c) FF= fluence factor= f(0*28 -o.1o*Iog(f))_
(d) Predicted ~RTNDT =CF* FF (e) Scatter ~RTNDT = Measured iill.TNDT- Predicted iill.TNDT (f) crt,. = l 7°F for base metal and crt,. = 28°F for welds All data points are no more than 2crt. above of the RG 1.99, Position 1.1 prediction. Therefore, the RG 1.99, Position 1.1 CF is conservative, and the RG 1.99, Position 2.1 CF may be used with a full margin term for North Anna Unit 1 Lower Shell Forging 03 and Heat# 25531 weld material.
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Westinghouse Non-Proprietary Class 3 C-5 C.2 Regulatory Guide 1.99, Revision 2, Position 1.1 Conservativism Evaluation for North Anna Unit 2 Surveillance Data The purpose of this evaluation is to determine if the North Anna Unit 2 non-credible surveillance data is less than 2crt, above of the RG 1.99, Position 1.1 prediction, thereby, determining whether the RG 1.99, Position 1.1 CF is conservative. This evaluation is performed in Table C-3.
Note the North Anna Unit 2 surveillance weld Heat# 716126 was determined to be credible in Appendix A.
Therefore, the conservatism of the Position 1.1 CF does not need to be determined because the Position 2.1 CF with a reduced margin term will be used regardless.
Table C-3 Conservatism Check for Position 1.1 for Non-Credible North Anna Unit 2 Surveillance Data CF<*> Capsule Measured Predicted Scatter Material Capsule (Pos. 1.1) Fluence<h) FF<cl ARTNDib) ARTNDT(d) ARTNDT(e) <2a<I)
(OF) (x 10 19 n/cm 2) (OF) (OF) (OF)
Intermediate Shell V 82.40 0.286 0.658 19 54.2 -35.2 Yes Forging 04 u 82.40 0.985 0.996 33 82.1 -49.1 Yes (Tangential) w 82.40 2.08 1.199 86 98.8 -12.8 Yes Intermediate Shell V 82.40 0.286 0.658 21 54.2 -33.2 Yes Forging 04 u 82.40 0.985 0.996 66 82.1 -16.1 Yes (Axial) w 82.40 2.08 1.199 65 98.8 -33.8 Yes Notes:
(a) CF values are taken from Table 3-8.
(b) Fluence and Measured LIB.TNDT values are taken from Table 3-6.
(c) FF= fluence factor= f(0-2 B-o.io*Iog(I))_
(d) Predicted ~RTNDT =CF* FF (e) Scatter ~RTNDT = Measured ~TNDT- Predicted ~RTNDT (f) crll. = l 7°F for base metal and crll. = 28°F for welds All data points are no more than 2crt, above of the RG 1.99, Position 1.1 prediction. Therefore, the RG 1.99, Position 1.1 CF is conservative, and the RG 1.99, Position 2.1 CF may be used with a full margin term for North Anna Unit 2 Intermediate Shell Forging 04.
C.3 REFERENCES C-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.
[ADAMS Accession Number ML003740284]
C-2. 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, Volume 60, No. 243, dated December 19, 1995, effective January 18, 1996.
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Westinghouse Non-Proprietary Class 3 C-6 C-3. K. Wichman, M. Mitchell, and A. Hiser, US NRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, "NRC/Industry Workshop on RPV Integrity Issues," February 12, 1998.
[ADAMS Accession Number MLJ 10070570}
C-4. NRC Letter "Closure of the Review of the Response to Generic Letter 92-01, Revision 1, Supplement 1, 'Reactor Vessel Structural Integrity,' the North Anna Nuclear Power Plant, Units 1 and 2 (TAC Nos. MA0555 and MA0556)," June 23, 1999. [Dominion Serial No.99-361, Incoming NRC Letter]
C-5. VEPCO Letter 99-361, "Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, Response to NRC Request for Comments Generic Letter 92-01, Revision 1, Supplement 1," September 1, 1999.
C-6. VEPCO Letter 99-452A, "Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, Surry Power Station Units 1 and 2, Evaluation of Reactor Vessel Materials Surveillance Data," November 19, 1999.
C-7. VEPCO Letter 00-306, "Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, Proposed Technical Specifications Changes Requests for Exemption per 10 CFR 50.60(b)
Reactor Coolant System Pressure/Temperature Limits, L TOPS Setpoints, and L TOPS Enable Temperatures," June 22, 2000.
C-8. VEPCO Letter 01-020, "Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, Response to Request for Additional Information, Proposed Technical Specifications Changes, Reactor Coolant System Pressure/Temperature Limits, L TOPS Setpoints, and L TOPS Enable Temperatures," January 4, 2001.
C-9. VEPCO Letter O1-020A, "Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, Response to Request for Additional Information, Proposed Technical Specifications Changes, Reactor Coolant System Pressure/Temperature Limits, L TOPS Setpoints, and L TOPS Enable Temperatures," February 14, 2001.
C-10. VEPCO Letter O1-020B, "Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, Response to Request for Additional Information and Clarification of Exemption Request Regarding Proposed Technical Specifications Changes for Reactor Coolant System Pressure/Temperature Limits, LTOPS Setpoints, and LTOPS Enable Temperatures," March 13, 2001.
C-11. VEPCO Letter 01-168, "Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, Response to Request for Additional Information, Proposed Technical Specifications Changes, Reactor Coolant System Pressure/Temperature Limits, LTOPS Setpoints, and L TOPS Enable Temperatures," March 22, 2001.
C-12. VEPCO Letter O1-168A, "Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, Proposed Technical Specifications Changes, Editorial Correction to Proposed Reactor Coolant System Pressure/Temperature Limit Curves Applicable to Cool down," April 11, 2001.
C-13. VEPCO Letter 01-262, "Virginia Electric and Power Company, North Anna Power Station Unit 2, Application of Sequoyah 2 Surveillance Data to North Anna Unit 2 Reactor Vessel Weld Material Fabricated from Weld Wire Heat 4278," April 27, 2001.
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Westinghouse Non-Proprietary Class 3 C-7 C-14. NRC Letter "North Anna Power Station, Units l and 2 - Issuance of Amendments and Exemption from the Requirements of 10 CFR Part 50, Section 50.60(a) Re: Amended Pressure-Temperature Limits (TAC Nos. MA9343, MA9344, MA9347, and MA9348)," May 2, 2001. [Dominion Serial No.01-293, Incoming NRC Letter]
C-15. VEPCO Letter 01-282, "Virginia Electric and Power Company, Surry and North Anna Power Stations Units 1 and 2 License Renewal Applications - Submittal," May 29, 2001. [ADAMS Accession Number ML0I 1500496]
C-16. VEPCO Letter 02-601, "Virginia Electric and Power Company (Dominion), Surry and North Anna Power Stations Units 1 and 2, Response to Request for Supplemental Information, License Renewal Applications," October 15, 2002. [ADAMS Accession Number ML022960411]
C-17. VEPCO Letter 04-380, "Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, Proposed Technical Specifications Change Request, Reactor Coolant System Pressure/Temperature Limits, LTOPS Setpoints and LTOPS Enable Temperatures," July 1, 2004.
C-18. VEPCO Letter 04-380A, "Virginia Electric and Power Company (Dominion), North Anna Power Station Units 1 and 2, Request for Additional Information, Proposed Technical Specifications Change Request, Reactor Coolant System Pressure/Temperature Limits, L TOPS Setpoints and LTOPS Enable Temperatures," October 28, 2004.
C-19. VEPCO Letter 04-380B, "Virginia Electric and Power Company (Dominion), North Anna Power Station Units 1 and 2, Editorial Correction for Proposed Technical Specifications Change Request, Reactor Coolant System Pressure/Temperature Limits, LTOPS Setpoints and LTOPS Enable Temperatures," November 16, 2004.
C-20. NRC Letter "North Anna Power Station, Units 1 and 2 - Issuance of Amendments on Reactor Coolant System Pressure and Temperature Limits (TAC Nos. MC3705 and MC3706)," July 8, 2005. [Dominion Serial No.05-460, Incoming NRC Letter]
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Westinghouse Non-Proprietary Class 3 D-1 APPENDIXD Justification for the Use of PWROG-17090-NP-A PWROG-17090-NP-A (Reference D-1) was approved to provide generic values of the unirradiated Charpy Upper-Shelf Energy (USE) for American Society of Mechanical Engineers (ASME) SA508, Class 2 (or the corresponding American Society for Testing and Materials [ASTM] A508, Class 2) RV forgings that were fabricated by the Rotterdam Dockyard Company (Rotterdam) as well as generic values of unirradiated Charpy USE, weight percentage copper (Cu) content, and weight percentage nickel (Ni) content for RV Submerged Arc Welds (SAWs) _g11d Shielded Metal Arc Welds (SMAWs).
In the NRC's Safety Evaluation (Reference D-2) it is stipulated that plants citing the report must ensure that their reactor vessel materials meet the criteria set forth below.
The generic properties provided in the TR are for implementation as conservative generic estimates for the :rzaterial classes identified below only if no measured values of unirradiated Charpy USE, Cu content, and/or Ni content are available for the specific RV material under consideration. PWR plants that implement these generic estimates must identify their RV materials as follows:

A PWR plant with a Rotterdam RV proposing to use the generic unirradiated Charpy USE value of 56 ft-lbs. for its RVforging(s) must identify that its forging(s) are of the SA5 08, Class 2 or A5 08, Class 2 specification and that the forging(s) were supplied by Rheinstahl Huttenwerke AG.

A PWR plant with a Rotterdam RV proposing to use the generic unirradiated Charpy USE value 52 ft-lbs. for its RV forging(s) must identify that its forging(s) are of the SA508, Class 2 orA508, Class 2 specification. This generic unirradiated Charpy USE value may be used if the Rotterdam RVforging supplier is identified as Fried-Krupp Huttenwerke AG or if the forging supplier is unknown.

A PWR plant with a Rotterdam RV proposing to use the generic unirradiated Charpy USE value o/75 ft-lbs. for its RV weld(s) must identify that the weld(s) are of the SAW type, that the SAWs are not ofLinde 80 flux type, and that its SAW(s) were fabricated by Rotterdam.

A PWR plant with a Rotterdam RV proposing to use the generic Cu content of 0.23 percent and generic Ni content of 0.56 percent for its RV weld(s) must identify that the weld(s) are of the SAW type, that the SA Ws are not of Linde 80 flux type, and that its SAW(s) were fabricated by Rotterdam.

A PWR plant with a Rotterdam RV proposing to use the generic unirradiated Charpy USE value of 72 ft-lbs. for its RV weld(s) must identify that the weld(s) were fabricated by Rotterdam. This generic unirradiated Charpy USE value may be used if the Rotterdam RV weld is identified as a SMAW or if the Rotterdam RV weld type is unknown.

A PWR plant with a Rotterdam RV proposing to use the RG 1.99, Rev. 2, default Cu content of0.35 percent and generic Ni content of 1.13 percent for its RVweld(s) must identify that the weld(s) were fabricated by Rotterdam. These values may be used ifthe Rotterdam RV weld is identified as a SMAW or if the Rotterdam RV weld type is unknown.

WCAP-18364-NP March 2020 Revision 1

This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

'fwestinghouse Non-Proprietary Class 3 D-2 Table D-1 below demonstrates how the relevant North Anna Units 1 and 2 materials meet these stipulations; thus, justifying the use of the listed generic values from PWROG-17090-NP-A for these materials.
Table D-1 North Anna Units 1 and 2 Reactor Vessel Materials which Use PWROG-17090-NP-A Generic Values Initial Flux Wt.% Wt.%
Material Description Heat USE Justification Type Cu Ni (ft-lbs)
Unit 1 The inlet/outlet nozzle to upper shell Inlet/Outlet Nozzle to Unknown 0.35 1.13 72 welds were fabricated by Rotterdam Upper Shell Welds with an unknown weld type.
Forging was supplied by Rheinstahl Inlet Nozzle Forging 11 990268-21 - - - 56 Huttenwerke AG and is ASTM A508, Class 2 material.
Unit2 Upper to Intermediate The weld was fabricated by Shell Circumferential 801 SMIT 89 - - 75 Rotterdam with a SAW weld type Weld (6% ID) without Linde 80 flux.
8816 The inlet/outlet nozzle to upper shell Inlet/Outlet Nozzle to 20459 welds were fabricated by Rotterdam LW320 0.23 0.56 75 with a SAW weld type without Upper Shell Welds 27622 Linde 80 flux.
Inlet Nozzle Forging 09 990426 - - - 56 Forgings were supplied by Rheinstahl Huttenwerke AG and are Outlet Nozzle Forging 13 990426-31 - - - 56 ASTM A508, Class 2 material.
D.1 REFERENCES D-1. PWROG Report PWROG-17090-NP-A, Revision 0, "Generic Rotterdam Forging and Weld Initial Upper Shelf Energy Determination," January 2020. [ADAMS Accession Number ML20024E238]
D-2. NRC Safety Evaluation, "Final Safety Evaluation by the Office of Nuclear Reactor Regulation for the Pressurized Water Reactor Owners Group Topical Report PWROG-17090-NP, Revision 0,
'Generic Rotterdam Forging and Weld Initial Upper-Shelf Energy Determination' (EPID L-2018-TOP-0017)," December 12, 2019. [ADAMS Accession Numbers ML19345F0J5 and ML19345Fl37}
WCAP-18364-NP March2020 Revision 1

This record was final approved on 3/31/2020 8:40:22 AM. (This statement was added by the PRIME system upon its validation)

ML20246G701

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