NL-20-1011, Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 Version 2.0

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Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 Version 2.0
ML20253A311
Person / Time
Site: Vogtle  Southern Nuclear icon.png
Issue date: 09/09/2020
From: Gayheart C
Southern Nuclear Operating Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-20-1011
Download: ML20253A311 (44)
v • d • e


Text

Cheryl A. Gayheart 3535 Colonnade Parkway Regulatory Affairs Director Birmingham, AL 35243 205 992 5316 tel 205 992 7795 fax cagayhea@southernco.com September 9, 2020 Docket Nos.: 50-424 NL-20-1011 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Vogtle Electric Generating Plant, Units 1 & 2 Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 Version 2.0 Ladies and Gentlemen:

By letters dated December 11, 2019 (Agencywide Documents Access and Management System Accession No. ML19347B105), Southern Nuclear Operating Company (SNC) submitted a request for relief for the Vogtle Electric Generating Plant (VEGP), Units 1 and 2. SNC requests to increase the inspection interval for ASME Section XI, Table IWC-2500-1, exam Category C-B, item number C2.21 and C2.22, exams from 10 years to 30 years through for the remainder of the 6th ISI Interval.

SNC has discovered two minor errors in the Alternative after the Nuclear Regulatory Commission (NRC) staff's audit, and is resubmitting the Alternative to provide an editorial correction within the Inspection History section and a correction in the Section 6.0 conclusions regarding pressure test frequency. SNC is also submitting supplemental information requested by the NRC during the audit. Enclosure 1 resubmits the VEGP Alternative and supersedes Version 1.0 of proposed ISI Alternative VEGP-ISI-ALT-04-04. provides SI Report No. 1900064.406.R0, Evaluations to Address Limited Examination Coverage of Vogtle Electric Generating Plant Units 1 and 2 Steam Generator Main Steam and Feedwater Nozzle-to-Shell Welds and Nozzle Inside Radius Sections. contains SI Report No. 1900064.407.R2, Evaluations to Address Benchmarking of the PROMISE Software to Include the Effects of Inspections.

This letter contains no NRC commitments. If you have any questions, please contact Jamie Coleman at 205.992.6611.

Respectfully submitted, Cheryl A. Gayheart Regulatory Affairs Director

U. S. Nuclear Regulatory Commission NL-20-1011 Page 2 CAG/DSP/sm : Proposed Alternative VEGP-ISI-ALT-04-04, Version 2.0, in Accordance with 10 CFR 50.55a(z)(1) : SI Report No. 1900064.406.R0 : SI Report No. 1900064.407.R2 cc: Regional Administrator NRR Project Manager - Vogtle 1 & 2 Senior Resident Inspector - Vogtle 1 & 2 RType: CVC7000

Vogtle Electric Generating Plant, Units 1 & 2 Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 Enclosure 1 Proposed Alternative VEGP-ISI-ALT-04-04, Version 2.0, in Accordance with 10 CFR 50.55a(z)(1) to NL-20-1011 Proposed Alternative VEGP-ISI-ALT-04-04, Version 2.0, in Accordance with 10 CFR 50.55a(z)(1) 1.0 ASME CODE COMPONENTS AFFECTED:

Code Class: Class 2

Description:

Nozzle-to-shell welds and inside radius sections Examination Category: C-B (Pressure Retaining Nozzle Welds in Pressure Vessels,Section XI, Division 1)

Item Numbers: C2.21 - Nozzle-to-shell (nozzle-to-head or nozzle-to-nozzle) welds C2.22 - Nozzle inside radius sections Component IDs:

11201-B6-001-W18 32" STEAM OUTLET NOZZLE TO UPPER HEAD WELD 11201-B6-001-W19 16" MAIN FEEDWATER NOZZLE TO SHELL WELD 11201-B6-002-W18 32" STEAM OUTLET NOZZLE TO UPPER HEAD WELD 11201-B6-002-W19 16" MAIN FEEDWATER NOZZLE TO SHELL WELD 11201-B6-003-W18 32" STEAM OUTLET NOZZLE TO UPPER HEAD WELD 11201-B6-003-W19 16" MAIN FEEDWATER NOZZLE TO SHELL WELD 11201-B6-004-W18 32" STEAM OUTLET NOZZLE TO UPPER HEAD WELD 11201-B6-004-W19 16" MAIN FEEDWATER NOZZLE TO SHELL WELD 11201-B6-001-IR04 MAIN FEEDWATER NOZZLE INNER RADIUS 11201-B6-002-IR04 MAIN FEEDWATER NOZZLE INNER RADIUS 11201-B6-003-IR04 MAIN FEEDWATER NOZZLE INNER RADIUS 11201-B6-004-IR04 MAIN FEEDWATER NOZZLE INNER RADIUS 21201-B6-001-W18 32" STEAM OUTLET NOZZLE TO UPPER HEAD WELD 21201-B6-001-W19 16" MAIN FEEDWATER NOZZLE TO SHELL WELD 21201-B6-002-W18 32" STEAM OUTLET NOZZLE TO UPPER HEAD WELD 21201-B6-002-W19 16" MAIN FEEDWATER NOZZLE TO SHELL WELD 21201-B6-003-W18 32" STEAM OUTLET NOZZLE TO UPPER HEAD WELD 21201-B6-003-W19 16" MAIN FEEDWATER NOZZLE TO SHELL WELD 21201-B6-004-W18 32" STEAM OUTLET NOZZLE TO UPPER HEAD WELD 21201-B6-004-W19 16" MAIN FEEDWATER NOZZLE TO SHELL WELD 21201-B6-001-IR04 MAIN FEEDWATER NOZZLE INNER RADIUS 21201-B6-002-IR04 MAIN FEEDWATER NOZZLE INNER RADIUS 21201-B6-003-IR04 MAIN FEEDWATER NOZZLE INNER RADIUS 21201-B6-004-IR04 MAIN FEEDWATER NOZZLE INNER RADIUS 2.0 REQUESTED APPROVAL DATE:

Approval is requested by December 31, 2020.

3.0 APPLICABLE CODE EDITION AND ADDENDA:

The Fourth Inservice Inspection (ISI) Interval Code of record for Vogtle Units 1 & 2 is the 2007 Edition with 2008 Addenda of ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components.

E1-1 to NL-20-1011 Proposed Alternative VEGP-ISI-ALT-04-04, Version 2.0, in Accordance with 10 CFR 50.55a(z)(1) 4.0 APPLICABLE CODE REQUIREMENT:

ASME Section XI IWC-2500(a), Table IWC-2500-1, Examination Category C-B, Item No.

C2.21 requires surface and volumetric examination of all representative steam generator nozzles at terminal ends of piping runs once during each Section XI inspection interval.

ASME Section XI IWC-2500(a), Table IWC-2500-1, Examination Category C-B, Item No.

C2.22 requires volumetric examination of representative steam generator all nozzle at terminal ends of piping runs once during each Section XI inspection interval. The examination areas for Item Nos. C2.21 and C2.22 are shown in Figures IWC-2500-4(a),

(b), and (d).

5.0 REASON FOR REQUEST:

The Electric Power Research Institute (EPRI) performed an assessment [1] of the basis for the ASME Section XI examination requirements specified for Examination Category C-B of ASME Section XI, Division 1 for Steam Generator (SG) Main Steam (MS) and Feedwater (FW) Nozzle-to-Shell Welds and Nozzle Inside Radius Sections. The assessment includes a survey of inspection results from 74 units as well as flaw tolerance evaluations using probabilistic fracture mechanics (PFM) and deterministic fracture mechanics (DFM). The Reference [1] report concluded that the current ASME Code Section XI inspection interval of ten years can be increased significantly with no impact to plant safety. It is upon the basis of this conclusion that an alternate inspection interval is being requested. The Reference [1] report was developed consistent with the recommendations provided in EPRIs White Paper on PFM [14].

6.0 PROPOSED ALTERNATIVE AND BASIS FOR USE:

Southern Nuclear Company (SNC) is requesting an inspection alternative to the examination requirements of ASME Section XI, Table IWC-2500-1, Examination Category C-B, Item Nos. C2.21 and C2.22. The proposed alternative is to increase the inspection interval for these examination items to 30 years (from the current ASME Code Section XI 10-year requirement) for the remainder of the 6th Inservice Inspection (ISI)

Interval. Although the EPRI report [1] supports a longer inspection period, 30 years was selected as a prudent alternative to ensure that one more examination was conducted prior to the end of the current license period for Vogtle Units 1 & 2. A summary of the key aspects of the technical basis for this request are summarized below. The applicability of the technical basis to Vogtle Units 1 & 2 is shown in Appendix A.

Degradation Mechanism Evaluation An evaluation of degradation mechanisms that could potentially impact the reliability of the SG MS and FW Nozzle-to-Shell Welds and Nozzle Inside Radius Sections was performed in Reference [1]. Evaluated mechanisms included stress corrosion cracking (SCC), environmental assisted fatigue (EAF), microbiologically influenced corrosion (MIC), pitting, crevice corrosion, erosion-cavitation, erosion, flow accelerated corrosion (FAC), general corrosion, galvanic corrosion, and mechanical/thermal fatigue. Other than the potential for EAF and mechanical/thermal fatigue, there were no active degradation mechanisms identified that significantly affect the long-term structural integrity of the SG MS and FW nozzles.

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Stress Analysis Finite element analysis (FEA) was performed in Reference [1] to determine the stresses in the SG MS and FW Nozzle-to-Shell Welds and Nozzle Inside Radius Sections. The analysis was performed using representative pressurized water reactor (PWR) geometries, bounding transients, and typical material properties. The results of the stress analyses were used in a flaw tolerance evaluation. The applicability of the FEA analysis to Vogtle Units 1 & 2 is shown in Appendix A and confirms that all plant-specific requirements are met. Therefore, the evaluation results and conclusions of Reference

[1] are applicable to Vogtle Units 1 & 2.

Flaw Tolerance Evaluation Flaw tolerance evaluations were performed in Reference [1] consisting of PFM evaluations and confirmatory DFM evaluations. The results of the PFM analyses indicate that, after a preservice inspection (PSI), no other inspections are required for up to 60 years of plant operation to meet the U.S. Nuclear Regulatory Commissions (NRCs) safety goal of 10-6 failures per year. For the specific case of Vogtle Units 1 and 2 where PSI followed by three 10-year interval inspections have been performed, Table 8-10 of Reference [1] indicates that if the inspection interval is increased to 30 years after these previous inspections, the NRC safety goal is met (with considerable margin) for up to 80 years of plant operation. The DFM evaluations provide verification of the PFM results by demonstrating that it takes approximately 80 years for a postulated flaw with an initial depth equal to the ASME Code Section XI acceptance standards to grow to a depth where the maximum stress intensity factor (K) exceeds the ASME Code Section XI allowable fracture toughness.

Inspection History Plant Vogtle Unit 1 and 2 operating experience (including examinations performed to date, examination findings, inspection coverage, and Relief Requests) is presented in Appendix B. As shown in this Appendix, Item No. C2.21 (FW nozzle and MS nozzle) examinations have had limited coverage. Also, as shown in Appendix B, no flaws that exceeded the ASME Code,Section XI acceptance standards were identified during any examinations.

Industry inspection history for these components (as obtained from an industry survey

[1]) is presented in Appendix C. The results of the survey [1] indicate that these components are very flaw tolerant.

Conclusion It is concluded that the SG MS and FW Nozzle-to-Shell Welds and Nozzle Inside Radius Sections are very flaw tolerant. PFM and DFM evaluations performed as part of the technical basis [1] demonstrate that, after PSI, no other inspection is required until 60 years to meet the NRC safety goal of 10-6 failures per reactor year. Plant-specific applicability of the technical basis to Vogtle Units 1 & 2 is demonstrated in Appendix A.

An inspection interval of 30 years provides an acceptable level of quality and safety in E1-3 to NL-20-1011 Proposed Alternative VEGP-ISI-ALT-04-04, Version 2.0, in Accordance with 10 CFR 50.55a(z)(1) lieu of the ASME Examination Category C-B, Item Nos. C2.21 and C2.22 surface and volumetric examination 10-year inspection frequency.

Operating and examination experience demonstrates that these components have performed with very high reliability, mainly due to their robust design. As shown in Appendix B, to date, SNC has performed 20 inspections of SG MS and FW Nozzle-to-Shell Welds and Nozzle Inside Radius Sections at Vogtle Units 1 & 2. No flaws that exceeded the ASME Code,Section XI acceptance standards were identified during any examinations, as shown in Appendix B. Some of the inspections listed in Appendix B involved limited coverage ranging from 50% to 80%. Section 8.2.5 of Reference [1]

discusses limited coverage and determines that the conclusions of the report are applicable to components with limited coverage. In addition, it is important to note all other inspection activities, including the system leakage test (Examination Category C-H) conducted each inservice inspection period (approximately every other refueling outage), will continue to be performed, providing further assurance of safety.

Finally, as discussed in Reference [2], for situations where no active degradation mechanism is present, it was concluded that subsequent inservice inspections do not provide additional value after PSI has been performed and the inspection volumes have been confirmed to have no flaws that exceeded the ASME Code,Section XI acceptance standards. The Vogtle Units 1 & 2 SG MS and FW nozzles have received the required PSI examinations and 20 follow-on inservice inspections with no flaws that exceeded the ASME Code,Section XI acceptance standards.

Therefore, SNC requests that the NRC authorize this proposed alternative in accordance with 10 CFR 50.55a(z)(1).

7.0 DURATION OF PROPOSED ALTERNATIVE:

The proposed Alternative is requested for the remainder of the 4th Inservice Inspection through 6th Inspection (ISI) Interval for Vogtle Units 1 & 2, currently scheduled to end on 5/30/47.

8.0 PRECEDENT

No previous submittals have been made requesting relief from the ASME Examination Category C-B, Item Nos. C2.21 and C2.22 surface and volumetric examinations on the basis of the Reference [1] technical basis. However, the following is a list of approved Relief Requests related to inspections of SG MS and FW nozzles:

  • Letter from J. W. Clifford (NRC) to S. E. Scace (Northeast Nuclear Energy Company), Safety Evaluation of the Relief Request Associated with the First and Second 10-Year Interval of the Inservice Inspection (ISI) Plan, Millstone Nuclear Power Station, Unit 3 (TAC No. MA 5446), dated July 24, 2000, ADAMS Accession No. ML003730922.
  • Letter from R. L. Emch (NRC) to J. B. Beasley, Jr. (SNOC), Second 10-Year Interval Inservice Inspection Program Plan Requests for Relief 13, 14, 15, 21 and 33 for Vogtle Electric Generating Plant, Units 1 and 2 (TAC No. MB0603 and MB0604), dated June 20, 2001, ADAMS Accession No. ML011640178.

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  • Letter from M, Khanna (NRC) to D. A. Heacock (Dominion Nuclear Connecticut Inc.), Millstone Power Plant Unit No. 2 - Issuance of Relief Requests RR-89-69 Through RR-89-78 Regarding Third 10-Year Interval Inservice Inspection plan (TAC Nos. ME5998 Through ME6006), dated March 12, 2012, ADAMS Accession No. ML120541062.
  • Letter from R. J. Pascarelli (NRC) to E. D. Halpin (PG&E), Diablo Canyon Plant, Units 1 and 2 - Relief Request; NDE SG-MS-IR, Main Steam Nozzle Inner Radius Examination Impracticality, Third 10-Year Interval, American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Inservice Inspection Program (CAC Nos. MF6646 and MF6647), dated December 8, 2015, ADAMS Accession No. ML15337A021.

In addition, there are precedents related to similar requests for relief for Class 1 nozzles:

  • Based on studies presented in Reference [3], the NRC approved extending PWR reactor vessel nozzle-to-shell welds from 10 to 20 years in Reference [4].

Finally, there are precedents that used generic industry guidance in a similar approach to the approach requested in this submittal:

  • NRC relief was granted for the Vogtle and Farley requests for alternatives to the Reactor Pressure Vessel Threads in Flange examination requirements in the reference [13] Safety Evaluation.

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9.0 ACRONYMS

ASME American Society of Mechanical Engineers B&W Babcock and Wilcox BWR Boiling Water Reactor BWRVIP Boiling Water Reactor Vessel and Internals Program CE Combustion Engineering CFR Code of Federal Regulations DFM Deterministic fracture mechanics EAF Environmentally assisted fatigue EPRI Electric Power Research Institute FAC Flow accelerated corrosion FEA Finite element analysis FW Feedwater ISI Inservice Inspection MIC Microbiologically influenced corrosion MS Main Steam NPS Nominal pipe size NRC Nuclear Regulatory Commission NSSS Nuclear steam supply system PFM Probabilistic fracture mechanics PWR Pressurized Water Reactor SCC Stress corrosion cracking SG Steam Generator SNC Southern Nuclear Company E1-6 to NL-20-1011 Proposed Alternative VEGP-ISI-ALT-04-04, Version 2.0, in Accordance with 10 CFR 50.55a(z)(1)

10.0 REFERENCES

1. Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Inside Radius Sections. EPRI, Palo Alto, CA: 2019. 3002014590.
2. American Society of Mechanical Engineers, Risk-Based Inspection: Development of Guidelines, Volume 2-Part 1 and Volume 2-Part 2, Light Water Reactor (LWR)

Nuclear Power Plant Components. CRTD-Vols. 20-2 and 20-4, ASME Research Task Force on Risk-Based Inspection Guidelines, Washington, D.C., 1992 and 1998.

3. B. A. Bishop, C. Boggess, N. Palm, Risk-Informed extension of the Reactor Vessel In-Service Inspection Interval, WCAP-16168-NP-A, Rev. 3, October 2011.
4. US NRC, Revised Safety Evaluation by the Office of Nuclear Reactor Regulation; Topical Report WCAP-16168-NP-A, Revision 2, Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval, Pressurized Water Reactor Owners Group, Project No. 694, July 26, 2011, ADAMS Accession No. ML111600303.
5. BWRVIP-108: BWR Vessels and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA 2002. 1003557.
6. US NRC, Safety Evaluation of Proprietary EPRI Report, BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108), December 19, 2007, ADAMS Accession No. ML073600374.
7. BWRVIP-241: BWR Vessels and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA 2010. 1021005.
8. US NRC, Safety Evaluation of Proprietary EPRI Report, BWR Vessel and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii (BWRVIP-241), April 19, 2013, ADAMS Accession Nos. ML13071A240 and ML13071A233.
9. Code Case N-702, Alternate Requirements for Boiling Water Reactor (BWR)

Nozzle Inner Radius and Nozzle-to-Shell Welds, ASME Code Section XI, Division 1, Approval Date: February 20, 2004.

10. U. S. NRC Regulatory Guide 1.147, Revision 18, Inservice Inspection Code Case Acceptability, ASME Code Section XI, Division 1, dated March 2017.
11. Southern Nuclear Company, NL-16-0724, Vogtle Electric Generating Plant, Units 1

& 2, Proposed lnservice Inspection Alternative VEGP-ISI-ALT-11, Version 1.0, June 28, 2016, ADAMS Accession No. ML16180A046.

12. Southern Nuclear Company, NL-16-0723, Joseph M. Farley Nuclear Plant, Unit 1, Proposed lnservice Inspection Alternative FNP-ISI-ALT-19, Version 1.0, June 30, 2016, ADAMS Accession No. ML16182A475.
13. Michael T. Markley (NRC) to Charles R. Pierce (Southern Nuclear), Vogtle Electric Generating Plant, Units 1 and 2, and Joseph M. Farley Nuclear Plant, Unit 1 -

Alternative to Inservice Inspection Regarding Reactor Pressure Vessel Threads E1-7 to NL-20-1011 Proposed Alternative VEGP-ISI-ALT-04-04, Version 2.0, in Accordance with 10 CFR 50.55a(z)(1)

Inflange Inspection (CAC Nos. MF8061, MF8062, MF8070), January 26, 2017, ADAMS Accession No. ML ML17006A109.

14. N. Palm (EPRI), BWR Vessel & Internals Project (BWRVIP) Memo No. 2019-016, White Paper on Suggested Content for PFM Submittals to the NRC, February 27, 2019, ADAMS Accession No. ML19241A545.
15. Structural Integrity Associates, Inc. Calculation No, FP-VOG-323, Revision 0, FatiguePro Analysis of Plant Data for Vogtle Units 1 and 2 through 2018.

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APPENDIX A VOGTLE UNIT 1 AND UNIT 2 APPLICABILITY E1-9 to NL-20-1011 Proposed Alternative VEGP-ISI-ALT-04-04, Version 2.0, in Accordance with 10 CFR 50.55a(z)(1)

Plant-Specific Applicability Section 9 of Reference [1] provides requirements that must be demonstrated in order to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for Vogtle Units 1 & 2 is provided in Table A1.

Table A1 indicates that all plant-specific requirements are met for Vogtle Units 1 & 2.

Therefore, the results and conclusions of the EPRI report are applicable to Vogtle Units 1

& 2.

Table A1. Applicability of Reference [1] Representative Analyses to Vogtle Units 1

&2 Category Requirement from Reference Applicability to Vogtle Units 1 & 2

[1]

General The nozzle-to-shell weld shall The Vogtle Units 1 & 2 MS and FW Requirements be one of the configurations nozzle configurations are shown in shown in Figure 1-1 or Figure Figures A1 and A2, and are 1-2 of Reference [1]. representative of the configuration shown in Figure 1-1 of Reference [1].

The materials of the SG shell, The Vogtle Units 1 & 2 nozzles are FW nozzles, and MS nozzles fabricated of SA-508, Class 2A must be low alloy ferritic steels material, and the SG vessel which conform to the heads/shells are fabricated from SA-requirements of ASME Code, 533, Gr. A, Cl. 2 material. Both of Section XI, Appendix G, these materials conform to the Paragraph G-2110. requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

The number of transients The transient cycles in Table 5-5 of shown in Table 5-5 of Reference [1] meet or exceed the 60-Reference [1] are bounding for year projected cycles for Vogtle Units application over a 60-year 1 and 2 as shown in Table A2 [15].

operating life.

SG Feedwater The piping attached to the FW The Vogtle Units 1 & 2 FW piping Nozzle nozzle must be 14-inch to 18- lines are both 16-inch NPS.

inch NPS.

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Category Requirement from Reference Applicability to Vogtle Units 1 & 2

[1]

The FW nozzle design must The Vogtle Units 1 & 2 FW nozzle have an integrally attached configuration is shown in Figure A1 thermal sleeve and has an integrally attached thermal sleeve.

SG Main Steam For Westinghouse and CE Vogtle Units 1 & 2 are Westinghouse Nozzle plants, the piping attached to 4-loop PWRs. The Vogtle Units 1 &

the SG MS nozzle must be 28- 2 MS nozzles have 32 to 26 inch to 36-inch NPS. reducers. The pipe size of the attached reducer to the nozzle end is 32 NPS which satisfies the intent of this requirement.

For B&W SGs, the piping This requirement is not applicable for attached to the main steam Vogtle Units 1 & 2 because they are nozzle must be 22-inch to 26- both Westinghouse 4-loop units.

inch NPS The SG must have one main As shown in Figure A3, Vogtle Units steam nozzle that exits the top 1 & 2 both have one MS nozzle per dome of the SG. SG that exits the top dome of each SG.

The main steam nozzle shall The Vogtle Units 1 & 2 MS nozzle not significantly protrude into configuration is shown in Figures A2 the SG (e.g., see Figure 4-7 of and A3, and does not protrude Reference [1]) or have a unique significantly into the SG. The Vogtle nozzle weld configuration (e.g., Units 1 & 2 MS nozzles are NOT see Figure 4-6 of Reference unique. They are similar to the

[1]). configuration selected for analysis (Figure 4-8 of Reference [1]).

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Figure A1 Vogtle Units 1 & 2 SG Feedwater Nozzle Configuration E1-12 to NL-20-1011 Proposed Alternative VEGP-ISI-ALT-04-04, Version 2.0, in Accordance with 10 CFR 50.55a(z)(1)

Figure A2 Vogtle Units 1 & 2 SG Main Steam Nozzle Configuration Figure A3 Steam Generator Upper Head E1-13 to NL-20-1011 Proposed Alternative VEGP-ISI-ALT-04-04, Version 2.0, in Accordance with 10 CFR 50.55a(z)(1)

Table A2 Transient Cycles for Vogtle Units 1 and 2 in Comparison to the Requirements in Reference [1]

Transient Cycles From Unit 1 60-Year Unit 2 60-Year Allowable Table 5-5 of Projected Cycles Projected Cycles Cycles From EPRI From Table 3 of From Table 4 of Tables 3 and 4 Report [15] [15] of [15]

3002014590

[1]

Heatup/Cooldown 300 69 76/75(5) 200 Plant Loading(1) 5000 164 141 500 Plant Unloading(2) 5000 66 36 500 (3)

Loss of Load 360 119 89 760 (4)

Loss of Power 60 3 3 40 Notes:

(1) Transient listed as Plant Loading 0-15% Power in Tables 3 and 4 of [15].

(2) Transient listed as Plant Unloading 0 - 15% Power in Tables 3 and 4 of [15].

(3) Loss of Load transient is a bundled to conservatively envelope a combination of several transients listed in Tables 3 and 4 of [15]:

  • Loss of Load w/o Rx Trip
  • Loss of RC Flow 1 Loop @ Power
  • Large Step Load Decrease

(4) Transient listed as Loss of Offsite Power in Tables 3 and 4 of [15].

(5) Cycles for Heatup and Cooldown, respectively.

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APPENDIX B VOGTLE UNITS 1 & 2 INSPECTION HISTORY E1-15 to NL-20-1011 Proposed Alternative VEGP-ISI-ALT-04-04, Version 2.0, in Accordance with 10 CFR 50.55a(z)(1)

VOGTLE UNITS 1 & 2 INSPECTION HISTORY Currently, the MS and FW nozzle components for VEGP Units 1 & 2 satisfy all of the inspection requirements of ASME Code,Section XI, 2007 Edition including the 2008 Addenda.

(Note: The first digit in each Component ID depicts the unit for each component.)

MS Nozzle Date Interval/Period Components ID Exam Coverage(3)

Results Item 10/26/88 1st/1st 11201-B6-001-W18 NRI 50%

No. 10/2/97 2nd /1st 11201-B6-001-W18 NRI 50%

C2.21 4/1/08 3rd/1st 11201-B6-001-W18 NRI 50%

3/29/14 3rd/3rd 11201-B6-001-W18 NRI 50%

10/9/90 1st/1st 21201-B6-001-W18 RI(1) 50%

10/19/99 2nd /1st 21201-B6-001-W18 NRI 50%

10/02/08 3rd/1st 21201-B6-001-W18 NRI 50%

9/24/14 3rd/3rd 21201-B6-001-W18 NRI 50%

(1) Subsurface Planer flaw acceptable per IWC-3510-1.

(3) The following relief requests address <90% inspection coverage for the 1st, 2nd, and 3rd Intervals: RR-29, RR-14, and VEGP-ISI-RR-05.

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FW Nozzle Date Interval/Period Components ID Exam Coverage(3)

Results Item 9/30/94 1st/ 3rd 11201-B6-002-W19 NRI 50%

No. 3/29/05 2nd/3rd 11201-B6-002-W19 NRI 50%

C2.21 10/7/09 3rd/1st 11201-B6-002-W19 NRI 80%

10/10/96 1st/3rd 21201-B6-002-W19 RI(2) 50%

10/1/05 2nd/3rd 21201-B6-002-W19 RI(2) 50%

3/19/10 3rd/3rd 21201-B6-002-W19 RI(2) 80%

Item 10/6/94 1st/3rd 11201-B6-002-IR04 NRI 100%

C2.22 3/25/05 2nd/3rd 11201-B6-002-IR04 NRI 100%

10/7/09 3rd/1st 11201-B6-002-IR04 NRI 100%

9/27/96 1st/3rd 21201-B6-002-IR04 NRI 100%

10/1/05 2nd/3rd 21201-B6-002-IR04 NRI 100%

3/18/10 3rd/2nd 21201-B6-002-IR04 NRI 100%

(2) Subsurface Planer flaw acceptable per IWC-3510-1.

(3) The following relief requests address <90% inspection coverage for the 1st, 2nd, and 3rd Intervals: RR-29, RR-14, and VEGP-ISI-RR-05.

E1-17 to NL-20-1011 Proposed Alternative VEGP-ISI-ALT-04-04, Version 2.0, in Accordance with 10 CFR 50.55a(z)(1)

APPENDIX C RESULTS OF INDUSTRY SURVEY E1-18 to NL-20-1011 Proposed Alternative VEGP-ISI-ALT-04-04, Version 2.0, in Accordance with 10 CFR 50.55a(z)(1)

Overall Industry Inspection Summary The results of an industry survey of past inspections of SG MS and FW nozzles are summarized in Section 3 of Reference [1]. Table C1 provides a summary of the combined survey results for Item Nos. C2.22, C2.21, and C2.32(1). The results identify that SG MS and FW Nozzle-to-Shell Welds and Nozzle Inside Radius Section examinations adversely impact outage activities including worker exposure, personnel safety, and radwaste. A total of 74 domestic and international BWR and PWR units responded to the survey and provided information representing all PWR plant designs currently in operation in the U.S.

This included 2-loop, 3-loop, and 4-loop PWR designs from each of the PWR nuclear steam supply system (NSSS) vendors (i.e., Babcock and Wilcox (B&W),

Combustion Engineering (CE), and Westinghouse). A total of 727 examinations for Item Nos. C2.21, C2.22, and C2.32(1) components were conducted, with 563 of these specifically for PWR components. The majority of the PWR examinations were performed on SG MS and FW nozzles. Only one PWR examination identified two (2) flaws that exceeded ASME Code Section XI acceptance criteria. The flaws were linear indications of 0.3 and 0.5 in length and were detected in a MS nozzle-to-shell weld using magnetic particle examination techniques. The indications were dispositioned by light grinding (ADAMS Accession No. ML13217A093).

Table C1 - Summary of Survey Results Number of Number Number of Plant Type Reportable of Units Examinations Indications BWR 27 164 0 PWR 47 563 2 Totals 74 727 2 1 Item No. C2.32 is similar to Item No. C2.21 and was evaluated in the Reference [1] technical basis and included in the industry survey. Vogtle Units 1 & 2 have not performed any examinations on Item No. C2.32 components.

E1-19

Vogtle Electric Generating Plant, Units 1 & 2 Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 Version 2.0 Enclosure 2 SI Report No. 1900064.406.R0 to NL-20-1011 SI Report No. 1900064.406.R0 5215 Hellyer Ave.

Suite 210 San Jose, CA 95138-1025 Phone: 408-978-8200 Fax: 408-978-8964 www.structint.com schesworth@structint.com July 16, 2020 SI Report No. 1900064.406.R0 Mr. Robert Grizzi Program Manager, NDE PD Operations and Issue Program Support Nuclear Sector Electric Power Research Institute 1300 West WT Harris Blvd.

Charlotte, NC 28262

Subject:

Evaluations to Address Limited Examination Coverage of Vogtle Electric Generating Plant Units 1 and 2 Steam Generator Main Steam and Feedwater Nozzle-to-Shell Welds and Nozzle Inside Radius Sections

Dear Bob:

Per your request, Structural Integrity Associates, Inc. (SI) performed an evaluation to determine the failure probabilities (rupture and leakage) considering the plant specific examination coverage for the Vogtle Electric Generating Plant (VEGP) steam generator main steam and feedwater nozzle-to-shell welds and nozzle inside radius sections using the PROMISE software. The evaluation methodology and results are presented in Attachment A to this letter report.

We appreciate the opportunity to provide you with this service. Please do not hesitate to let me know if you have any questions.

Very truly yours, Scott Chesworth Senior Consultant cc: G. Stevens (EPRI)

E2-1 to NL-20-1011 SI Report No. 1900064.406.R0 ATTACHMENT A EVALUATIONS TO DETERMINE FAILURE PROBABILITIES CONSIDERING 50%

EXAMINATION COVERAGE FOR VOGTLE ELECTRIC GENERATING PLANT UNITS 1 AND 2 STEAM GENERATOR MAIN STEAM AND FEEDWATER NOZZLE-TO-SHELL WELDS AND NOZZLE INSIDE RADIUS SECTIONS Attachment A to 1900064.406.R0 A-1 E2-2 to NL-20-1011 SI Report No. 1900064.406.R0 BACKGROUND In Section 8.2.5 of EPRI Report 3002014590 (Reference [1]), the impacts of reduced examination coverage for steam generator (SG) main steam (MS) and feedwater (FW) nozzle-to-shell welds and nozzle inside radius sections were qualitatively evaluated. It was concluded that inservice inspection (ISI) examination coverage of any extent after the preservice inspection (PSI) examination (which has an assumed 100% coverage) is acceptable, since the probabilities of rupture and leakage with only a PSI examination (and no other follow-on ISI examinations for 80 years) are three orders of magnitude below the acceptance criteria for all but a single case, as shown in Table 1 (reproduction of Table 8-9 of Reference [1]).

The sole exception is the probability of leakage (but not rupture) for Case ID FEW-P3A, highlighted in Table 1. The probability of leakage after PSI and after 60 and 80 years of operation for this case exceeds the acceptance criteria by one order of magnitude (2.44x10-6 and 1.19x10-5 vs. an allowed value of 1x10-6). As explained in Sections 8.2.5 and 8.2.4.1.1 of Reference [1], this result was considered acceptable because (1) the increased likelihood of pressure boundary leakage is detectable by plant operators, (2) plant procedures allow for safe plant shutdown once any leakage is detected, and (3) the probability of rupture values are maintained three orders of magnitude below the acceptance criterion, thus eliminating the possibility of failure of the pressure boundary.

Southern Nuclear Operating Company (SNOC) is requesting an examination alternative [2] to that mandated in ASME Code,Section XI for Examination Category C-B, Item No. C2.21 and C2.22 components associated with the SG MS and FW nozzles for Vogtle Electric Generating Plant (VEGP) Units 1 and 2. The proposed alternative is to extend the ISI interval for these examinations to 30 years (from the current ASME Code,Section XI 10-year requirement). The examination history for VEGP Units 1 and 2 is provided in Appendix B of Reference [2] and is reproduced in Tables 2 and 3. As indicated in Tables 2 and 3, in addition to the PSI examination, four 10-year ISI examinations have been performed for VEGP Units 1 and 2 MS nozzle-to-shell welds, and three 10-year ISI examinations have been performed for VEGP Units 1 and 2 FW nozzle-to-shell welds, subsequent to the PSI examinations. All examinations had limited coverage ranging from 50% to 80%.

In the present study, additional probabilistic fracture mechanics (PFM) evaluations were performed to determine the probabilities of rupture and leakage based on the actual examination coverage obtained at VEGP Units 1 and 2. This evaluation only addresses limited coverage for the Item No. C2.21 nozzle-to-shell welds, since the Item No. C2.22 MS nozzle inner radii are exempt from ISI examinations and the FW nozzle inner radii had 100% coverage for all past ISI examinations (see Table 3).

Attachment A to SI Report 1900064.406.R0 A-2 E2-3 to NL-20-1011 SI Report No. 1900064.406.R0 Table 1 Probability of Rupture (per Year) and Probability of Leakage (per Year) for PSI Only (Table 8-9 from Ref. [1])

P(Leakage) at Case P(Rupture)

Component Identification at 80 yrs.

20 yrs. 40 yrs. 60 yrs. 80 yrs.

SGW-P1N 1.25E-12 5.00E-12 2.50E-12 1.67E-12 1.25E-12 Westinghouse Main Steam Nozzle SGW-P2C 1.25E-09 5.00E-09 2.50E-09 1.67E-09 1.25E-09 (SGW)

SGW-P2A 1.25E-09 5.00E-09 2.50E-09 1.67E-09 1.25E-09 SGB-P1N 1.25E-12 5.00E-12 2.50E-12 1.67E-12 2.50E-12 SGB-P2N 1.25E-12 5.00E-12 2.50E-12 1.67E-12 1.25E-12 B&W Main Steam SGB-P3C 1.25E-09 5.00E-09 2.50E-09 1.67E-09 1.25E-09 Nozzle (SGB) SGB-P3A 1.25E-09 5.00E-09 2.50E-09 1.67E-09 1.25E-09 SGB-P4A 1.25E-09 5.00E-09 2.50E-09 1.67E-09 1.25E-09 SGB-P4C 1.25E-09 5.00E-09 2.50E-09 1.67E-09 1.25E-09 FEW-P1N 1.25E-12 4.50E-11 9.88E-09 9.68E-08 3.20E-07 FEW-P2N 1.25E-12 5.00E-12 2.50E-12 1.67E-12 1.25E-12 Westinghouse FEW-P3C 1.25E-09 5.00E-09 2.50E-09 1.67E-09 1.25E-09 Feedwater Nozzle (FEW) FEW-P3A 1.25E-09 5.00E-09 2.08E-07 2.44E-06 1.19E-05 FEW-P4A 1.25E-09 5.00E-09 2.50E-09 1.67E-09 1.25E-09 FEW-P4C 1.25E-09 5.00E-09 2.50E-09 1.67E-09 1.25E-09 Note: The limiting case is displayed in bold red text highlighted in yellow.

Attachment A to SI Report 1900064.406.R0 A-3 E2-4 to NL-20-1011 SI Report No. 1900064.406.R0 Table 2 VEGP Units 1 and 2 Main Steam Nozzle Examination History Summary [2]

Table 3 VEGP Units 1 and 2 Feedwater Nozzle Examination History Summary [2]

Attachment A to SI Report 1900064.406.R0 A-4 E2-5 to NL-20-1011 SI Report No. 1900064.406.R0 EVALUATION As shown in Tables 2 and 3, examination coverage for Item No. C2.21 for both the VEGP Units 1 and 2 MS and FW nozzle-to-shell welds ranged from 50% to 80%. Item No. C2.21 was therefore evaluated using the minimum coverage of 50% achieved during these exams. Per Table 1, Case ID FEW-P3A has the highest probability of leakage value of 1.19x10-5, so the evaluation was performed for this limiting case.

Two ISI scenarios were considered:

1. The current ASME Code,Section XI examination requirement, which involves 10-year interval examinations after the PSI examination. For this case, the evaluation was performed assuming VEGP Units 1 and 2 will continue with the current 10-year inspection interval through 70 years of operation (i.e., ISI at 10, 20, 30, 40, 50, 60, and 70 years).
2. The alternative [2] requested for VEGP Units 1 and 2, where only the first three 10-year ISI examinations are performed after the PSI examination, followed by one examination on a 30-year interval (i.e., ISI at 10, 20, 30, and 60 years).

Plant records reflect that ASME Code,Section III PSI examinations involving radiographic testing (RT) were performed and found acceptable for the affected welds of the VEGP Units 1 and 2 MS and FW nozzles. The acceptability of the ASME Section III RT examinations indicates that 100% coverage was achieved during these examinations. PSI ultrasonic testing (UT) examinations were not performed on the vessel side of the MS or FW welds due to difficulties associated with the nozzle configurations. However, because of the success with the RT examinations, this is considered acceptable since as discussed in Section 8.2.4.1.1 of the Reference report [1], PSI refers to the collective initial ASME Code,Section III and Section XI examinations.

A comprehensive study performed in References [4] and [5] concluded that detection and sizing of flaws utilizing RT is as effective as UT. Figure 3.6 of Reference [4] provides theoretical probability of detection (POD) curves for RT examinations. The most conservative of these POD curves is compared to that from UT examinations used in the Reference [1] evaluations and is presented in Figure 1. As shown in this figure, except for extremely shallow flaws (less than 0.04 inches), the POD curve used in Reference [1] can be conservatively applied to RT examinations. It should be noted that the minimum flaw depth in all the simulations is 0.075 inches which is greater than the flaw depths at which RT governs.

The PROMISE software [3], which was used to perform the PFM evaluations in Reference [1],

was used to perform the present evaluation which covers 80 years of plant operation. The probabilities of rupture and leakage were determined for the two ISI scenarios discussed above (Section XI and the requested alternative) for the limiting case (Case ID FEW-P3A).

Attachment A to SI Report 1900064.406.R0 A-5 E2-6 to NL-20-1011 SI Report No. 1900064.406.R0 Figure 1 Comparison of POD Curve Used in Reference [1] (Based on UT) to that Based on RT

[4]

RESULTS The results of the evaluation are presented in Table 4. As shown in this table, the probabilities of rupture for 100% ISI coverage and 50% ISI coverage are identical and remain unchanged from the results shown in Table 1 for PSI examination only. The values of 1.25x10-9 for both scenarios are approximately three orders of magnitude less than the acceptance criteria of 1x10-6 specified in Reference [1].

The probabilities of leakage for the ASME Section XI (PSI+10+20+30+40+50+60+70) and alternative ISI (PSI+10+20+30+60) scenarios assuming 50% ISI coverage are both nearly equal at 5.9x10-6, which is above the acceptance criterion. However, these values are lower than the PSI only (100% coverage) value of 1.19x10-5 reported in Table 1 (Table 8-9 of Reference [1]).

Therefore, for the same reasons as for the rupture probabilities, the probability of leakage is decreased by performing ISI, regardless of coverage.

As indicated by comparison of the two scenarios in Table 4, the probability of leakage for 80 years for the alternative examination scenario (Scenario #2) is essentially identical to Scenario

  1. 1 where the ASME Code,Section XI 10-year examinations are performed through 70 years of Attachment A to SI Report 1900064.406.R0 A-6 E2-7 to NL-20-1011 SI Report No. 1900064.406.R0 operation. A further comparison of the cumulative probabilities of leakage vs. time for the two ISI scenarios is shown in Figure 2. The results presented in this figure show that the probability of leakage for the alternative ISI scenario is almost identical to the scenario where the ASME Code,Section XI 10-year ISI examinations are continued through 70 years of operation.

Regardless of which of the two ISI scenarios is considered, the probability of leakage remains the same after 80 years of operation. Therefore, changing the ISI schedule to the proposed alternative [2] does not alter the probability of leakage compared to the current ASME Code,Section XI schedule of repeated 10-year examinations.

Table 4 Sensitivity Study for ISI Coverage - Limiting Case FEW-P3A Probability of Rupture Probability of Leakage (per Year) at 80 years (per Year) at 80 years Case ID Scenario PSI/ISI Schedule 100% ISI 50% ISI 100% ISI 50% ISI Coverage Coverage Coverage Coverage 1 PSI+10+20+30+40+50+60+70 1.25E-09 1.25E-09 1.25E-09 5.93E-06 FEW-P3A 2 PSI+10+20+30+60 1.25E-09 1.25E-09 2.50E-09 5.95E-06 Figure 2 Comparison of Probability of Leakage for Two ISI Scenarios (Scenario #1 =

PSI+10+20+30+40+50+60+70 = Current ASME Code,Section XI Requirement) and (Scenario #2 = PSI+10+20+30+60 = VEGPs Alternative Request) for Limiting Case ID FEW-P3A Attachment A to SI Report 1900064.406.R0 A-7 E2-8 to NL-20-1011 SI Report No. 1900064.406.R0

SUMMARY

AND CONCLUSIONS PFM evaluations for the minimum 50% coverage achieved during examinations at VEGP Units 1 and 2 with various ISI scenarios were performed to determine the impact of reduced ISI coverage for the SG MS and FW nozzle-to-shell welds (ASME Code,Section XI, Examination Category C-B, Item No. C2.21). Because of the successful ASME Code,Section III RT examinations performed after fabrication of the welds, 100% coverage was used for the PSI examinations. The POD curve for UT used for ISI examinations in Reference [1] was conservatively applied to the PSI examinations.

It is concluded that limited coverage of as low as 50% for the VEGP Units 1 and 2 SG MS and FW nozzle-to-shell welds is acceptable for continued operation for the alternative requested by SNOC in Reference [2]. The probabilities of rupture for 50% coverage for the limiting case are three orders of magnitude below the acceptance criteria for 80 years of operation. The probability of leakage for the limiting case using the alternative ISI scenario (PSI+10+20+30+60) is slightly above the acceptance criterion (5.95x10-6 vs. 1x10-6); however, this probability of leakage is almost identical to the scenario where the ASME Code,Section XI 10-year ISI examinations are continued through 70 years of operation (5.93x10-6). Therefore, the examination interval associated with the VEGP Request for Alternative [2] does not increase the probabilities of rupture and does not significantly increase the probability of leakage from the currently required ASME Code,Section XI ISI examination interval.

As discussed in Sections 8.2.4.1.1 and 8.2.5 of Reference [1], the fact that the probability of leakage at location FEW-P3A slightly exceeds the acceptance criterion does not compromise plant safety. This is because pressure boundary leakage is detectable by plant operators, plant procedures allow for safe plant shutdown once any leakage is detected, and the probability of rupture values are maintained well below the acceptance criterion for 80 years of operation even under the scenario of PSI examination only.

REFERENCES

1. Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Nozzle Inside Radius Sections. EPRI, Palo Alto, CA: 2019. 3002014590.
2. Letter No. NL-19-0832 from Cheryl A. Gayheart (Southern Nuclear) to U. S. Nuclear Regulatory Commission, Vogtle Electric Generating Plant, Units 1 & 2, Proposed Inservice Inspection Alternate VEGP-ISI-ALT-04-04, December 11, 2019.
3. Structural Integrity Associates Report DEV1806.402, PROMISE 1.0 Theory and Users Manual, Revision 0.
4. T. L. Moran, P. Ramuhalli, A. F. Pardini, M. T. Anderson and S. R. Doctor, Replacement of Radiography with Ultrasonics for Nondestructive Inspection of Welds - Evaluation of Technical Gaps - An Interim Report, Pacific Northwest National Laboratory Report PNNL-19086, April 2010, ADAMS Accession No. ML101031254.
5. T. L. Moran, M. Prowant, C. A. Nove, A. F. Pardini, S. L. Crawford, A. D. Cinson and M.

T. Anderson, Applying Ultrasonic Testing in Lieu of Radiography for Volumetric Examination of Carbon Steel Piping, NUREG/CR-7204 (PNNL-24232), September 2015.

Attachment A to SI Report 1900064.406.R0 A-8 E2-9 to NL-20-1011 SI Report No. 1900064.406.R0 Prepared by: Verified by:

7/16/2020 7/16/2020 Dilip Dedhia Date Nathaniel G. Cofie Date Senior Associate Senior Associate Verified by: Approved by:

7/16/2020 7/16/2020 D. J. Shim Date Scott T. Chesworth Date Associate Senior Consultant Attachment A to SI Report 1900064.406.R0 A-9 E2-10

Vogtle Electric Generating Plant, Units 1 & 2 Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 Version 2.0 Enclosure 3 SI Report No. 1900064.407.R2 to NL-20-1011 SI Report No. 1900064.407.R2 5215 Hellyer Ave.

Suite 210 San Jose, CA 95138-1025 Phone: 408-978-8200 Fax: 408-978-8964 www.structint.com schesworth@structint.com August 5, 2020 SI Report No. 1900064.407.R2 Mr. Robert Grizzi Program Manager, NDE PD Operations and Issue Program Support Nuclear Sector Electric Power Research Institute 1300 West WT Harris Blvd.

Charlotte, NC 28262

Subject:

Evaluations to Address Benchmarking of the PROMISE Software to Include the Effects of Inspections

Dear Bob:

Per your request, Structural Integrity Associates, Inc. (SI) has performed evaluations to benchmark the PROMISE software against the VIPER-NOZ software to include the effect of inspections. The evaluation methodology and results are presented in Attachment A to this letter report. The updated results include an additional case with a different PSI/ISI combination.

Furthermore, probabilities of rupture are provided in addition to the probabilities of leakage, as requested by the U.S. NRC during the July 27th, 2020 PROMISE software audit.

We appreciate the opportunity to provide you with this service. Please do not hesitate to let me know if you have any questions.

Very truly yours, Scott Chesworth Senior Consultant cc: G. Stevens (EPRI)

E3-1 to NL-20-1011 SI Report No. 1900064.407.R2 ATTACHMENT A EVALUATIONS TO BENCHMARK THE PROMISE SOFTWARE TO INCLUDE THE EFFECT OF PRESERVICE AND INSERVICE INSPECTION Attachment A to 1900064.407.R2 A-1 E3-2 to NL-20-1011 SI Report No. 1900064.407.R2 BACKGROUND In Section 8.2.3.2.2 of EPRI Report 3002014590 (Reference [1]), the PROMISE probabilistic fracture mechanics (PFM) software [2] was benchmarked against the VIPER-NOZ software code [3]. The VIPER-NOZ software, used for performing PFM analyses of the BWR vessel nozzle-to-shell welds and the nozzle inner radii in BWRVIP-108-A [4], was chosen for the benchmarking because it was reviewed extensively by the NRC as a part of their Safety Evaluation (SE) approving BWRVIP-108-A, and subsequently BWRVIP-241-A [5]. A summary of the key inputs to the benchmarking exercise is provided in Table 8-4 of Reference [1], which is reproduced in Table 1. The results of the benchmarking are shown Table 8-5 of Reference

[1], which is reproduced in Table 2. As shown in Table 2, the probabilities of leakage are very similar.

Table 1 Benchmarking Inputs (Table 8-4 of Reference [1])

Input Value No. of cracks per inner radius section 1, constant Crack depth distribution PVRUF Fracture toughness (ksiin) Normal (200,5)

PSI None ISI None POD Curve Not applicable Fatigue crack growth law and threshold BWRVIP-108-A Uncertainties on transients None Residual stresses (ksi) None Table 2 Comparison of Cumulative Probability of Leakage Between PROMISE and VIPER-NOZ for Benchmarking (Table 8-5 of Reference [1])

Cyclic Stress (ksi) Cycles/year PROMISE VIPER-NOZ 25 500 2.8E-2 3.1E-2 15 500 1.7E-4 3.0E-4 As shown in Table 1, neither preservice inspection (PSI) nor inservice inspection (ISI) were considered in the benchmarking. Since one of the key features of the PROMISE software is its ability to evaluate the impacts of ISI on failure probabilities, the NRC requested during their July 1, 2020 audit of the PROMISE software that the benchmarking should include consideration of inspections to provide a benchmarking demonstration of the ISI capabilities of PROMISE.

Furthermore, during the July 27, 2020 follow-on audit, the NRC requested that in addition to the probability of leakage, a comparison of the probability of rupture be included in the benchmarking.

Attachment A to SI Report 1900064.407.R2 A-2 E3-3 to NL-20-1011 SI Report No. 1900064.407.R2 TECHNICAL APPROACH Two scenarios using different PSI/ISI scenarios were considered in the benchmarking. The input parameters for the two scenarios are provided in Tables 3 and 4. The changes to the inputs from the previous benchmarking exercise included in Reference [1] are shown in red italic text in Tables 3 and 4. As shown in these tables, a combination of PSI/ISI cases were considered in each of the two scenarios to determine the trending associated with ongoing inspections with both software codes. The two scenarios are defined as follows:

Scenario No. 1 considered four cases of PSI alone and PSI followed by 20-year ISI examinations up to 60 years as shown in Table 3.

Scenario No. 2 considered eight cases of PSI alone and PSI followed by 10-year ISI examinations up to 70 years as shown in Table 4.

The POD curve used in Reference [1] was also employed in this updated benchmarking exercise (i.e., Figure 8-2 of Reference [1] = the POD curve from BWRVIP-108-A and BWRIP-241-A).

Similar to the benchmarking exercise performed in Reference [1], the nozzle corner crack model was used to determine the stress intensity factors since the model is common to both software codes. A conservative combination of stress and fracture toughness was used to increase the likelihood of failure. For simplicity, a constant through-wall stress of 30 ksi was applied. The mean fracture toughness was lowered to 100 ksiin (from the original benchmarking value of 200 ksiin) with a standard deviation of 20 ksiin (compared to the original benchmarking value of 5 ksiin).

The BWRVIP-108-A fatigue crack growth (FCG) equation was used, along with a Weibull distribution for the coefficient of the FCG equation, and the FCG threshold was assumed to be zero. In addition, the PVRUF crack depth distribution was used for the initial crack size. All of these crack growth and crack depth distribution inputs remain identical to what was were used in the initial benchmarking exercise in Section 8.2.3.2.2 of Reference [1]. In both software codes, the inputs for the time increment for updating the crack growth was set to one-tenth of a year.

Attachment A to SI Report 1900064.407.R2 A-3 E3-4 to NL-20-1011 SI Report No. 1900064.407.R2 Table 3 Updated Benchmarking Inputs for First PSI/ISI Combination (Scenario No. 1)

Input Value No. of cracks per inner radius section 1, constant Crack depth distribution PVRUF Fracture toughness (ksiin) Normal (100, 20)

PSI Yes (at 0 years)

ISI Yes 4 cases:

1) No ISI
2) ISI at 20 yrs
3) ISI at 20, 40 yrs
4) ISI at 20, 40, 60 yrs POD Curve Figure 8-2 of Reference [1]

Fatigue crack growth law and threshold BWRVIP-108-A Applied Stress 30 ksi through thickness Transient stresses and uncertainties None Residual stresses (ksi) None Time increment for updating crack growth calculation One tenth of a year Cycles 500 per Year Attachment A to SI Report 1900064.407.R2 A-4 E3-5 to NL-20-1011 SI Report No. 1900064.407.R2 Table 4 Updated Benchmarking Inputs for Second PSI/ISI Combination (Scenario No. 2)

Input Value No. of cracks per inner radius section 1, constant Crack depth distribution PVRUF Fracture toughness (ksiin) Normal (100, 20)

PSI Yes (at 0 years)

ISI Yes 8 cases:

1) No ISI
2) ISI at 10 yrs
3) ISI at 10, 20 yrs
4) ISI at 10, 20, 30 yrs
5) ISI at 10, 20, 30, 40 yrs
6) ISI at 10, 20, 30, 40, 50 yrs
7) ISI at 10, 20, 30, 40, 50, 60 yrs
8) ISI at 10, 20, 30, 40, 50, 60, 70 yrs POD Curve Figure 8-2 of Reference [1]

Fatigue crack growth law and threshold BWRVIP-108 Applied Stress 30 ksi through thickness Transient stresses and uncertainties None Residual stresses (ksi) None Time increment for updating crack growth calculation One tenth of a year Cycles 1,000 per Year EVALUATION The probabilities of leakage and rupture were determined for multiple cases for each of the two PSI/ISI scenarios shown in Tables 3 and 4 using both the PROMISE and VIPER-NOZ software codes. The constant stress of 30 ksi applied to the nozzle corner was cycled from the constant value to zero with 500 cycles per year for Scenario No. 1 and 1,000 cycles per year for Scenario No. 2. The added number of cycles for Scenario No. 2 is to increase the likelihood of failure since it involves many more inspections.

RESULTS The probabilities of leakage and rupture as a function of the cases for each PSI/ISI scenario are presented in Figures 1 and 2 for Scenario No. 1 and Figures 3 and 4 for Scenario No. 2. As shown in these figures, there is very good agreement between the PROMISE and VIPER-NOZ software results for all cases for both scenarios. As expected, both software codes indicate that the probabilities of leakage and rupture decrease as more inspections are performed.

Attachment A to SI Report 1900064.407.R2 A-5 E3-6 to NL-20-1011 SI Report No. 1900064.407.R2 Figure 1 Comparison of Cumulative Probabilities of Leakage Between the PROMISE and the VIPER-NOZ Software Codes for Scenario No. 1 Figure 2 Comparison of Cumulative Probabilities of Rupture Between the PROMISE and the VIPER-NOZ Software Codes for Scenario No. 1 Attachment A to SI Report 1900064.407.R2 A-6 E3-7 to NL-20-1011 SI Report No. 1900064.407.R2 Figure 3 Comparison of Cumulative Probabilities of Leakage Between the PROMISE and the VIPER-NOZ Software Codes for Scenario No. 2 Figure 4 Comparison of Cumulative Probabilities of Rupture Between the PROMISE and the VIPER-NOZ Software Codes for Scenario No. 2 Attachment A to SI Report 1900064.407.R2 A-7 E3-8 to NL-20-1011 SI Report No. 1900064.407.R2

SUMMARY

AND CONCLUSION PFM evaluations were performed to benchmark the PROMISE software against the VIPER-NOZ software including the effect of inspections (both PSI and ISI) using multiple cases of different PSI/ISI combinations for two scenarios to determine the effect of ISI on failure probabilities for both leakage and rupture. Scenario No. 1 considered four cases of PSI alone and PSI followed by 20-year ISI examinations up to 60 years. Scenario No. 2 considered eight cases of PSI alone and PSI followed by 10-year ISI examinations up to 70 years. The results from the two software codes are in very good agreement for all cases for both scenarios, indicating that both software codes produce consistent probabilities of leakage and rupture using identical inputs.

REFERENCES

1. Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Nozzle Inside Radius Sections. EPRI, Palo Alto, CA: 2019. 3002014590.
2. Structural Integrity Associates Report DEV1806.402, PROMISE 1.0 Theory and Users Manual, Revision 0.
3. Structural Integrity Associates, VIPER-NOZ Version 1.1.
4. BWRVIP-108-A: BWR Vessels and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA 2018. 3002013092.
5. BWRVIP-241-A: BWR Vessel and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii. EPRI, Palo Alto, CA: 2018. 3002013093.

Attachment A to SI Report 1900064.407.R2 A-8 E3-9 to NL-20-1011 SI Report No. 1900064.407.R2 Prepared by: Verified by:

8/5/2020 8/5/2020 Dilip Dedhia Date Nathaniel G. Cofie Date Senior Associate Senior Associate Verified by: Approved by:

8/5/2020 8/5/2020 D. J. Shim Date Scott T. Chesworth Date Associate Senior Consultant Attachment A to SI Report 1900064.407.R2 A-9 E3-10

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