1CAN062301, Request for Alternative for Implementation of Extended Reactor Vessel Inservice Inspection Interval (ANO1-ISI-037)

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Request for Alternative for Implementation of Extended Reactor Vessel Inservice Inspection Interval (ANO1-ISI-037)
ML23159A269
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 06/08/2023
From: Couture P
Entergy Operations
To:
Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML23159A268 List:
References
1CAN062301
Download: ML23159A269 (1)
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NOTICE: Enclosure 1 to this letter contains Proprietary Information to be withheld from public disclosure per 10 CFR 2.390. Upon separation from Enclosure 1, this letter is DECONTROLLED Entergy Operations, Inc., 1340 Echelon Parkway, Jackson, MS 39213 1CAN062301 10 CFR 50.55a June 8, 2023 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Request for Alternative for Implementation of Extended Reactor Vessel lnservice Inspection Interval (ANO1-ISI-037)

Arkansas Nuclear One, Unit 1 NRC Docket No. 50-313 Renewed Facility Operating License No. DPR-51 Pursuant to 10 CFR 50.55a(z)(1), Entergy Operations, Inc. (Entergy) requests NRC approval to extend the inservice inspection interval for the Arkansas Nuclear One, Unit 1 (ANO-1) reactor pressure vessel (RPV) weld examinations from 2027 to 2034. Entergy proposes to implement an alternative to the requirement of American Society of Mechanical Engineers (ASME)

Section XI, IWB 2411, Inspection Program, that volumetric examination of RPV Examination categories B-A and B-D be performed once each 10-year Inservice Inspection (ISI) interval.

The enclosed request concludes that the current inspection interval of 10 years can be revised to 20 years with negligible change in risk. This is done by satisfying the risk criteria specified in Regulatory Guide 1.174, Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," U.S. Nuclear Regulatory Commission [ADAMS Accession Number ML023240437], dated November 2002. In addition, the requested alternative would provide an acceptable level of quality and safety.

Entergy requests approval of the requested alternative to the end of the current renewed ANO-1 Operating License (May 2034).

Although the analysis supports a longer period of extension, Entergy will perform the required B-A and B-D examinations prior to the end of the current operating license for ANO-1.

provides the basis for the proposed alternative.

Some information provided in Enclosure 1 is considered proprietary to Framatome and request it to be withheld from public disclosure in accordance with 10 CFR 2.390 of the Commissions Phil Couture Senior Manager Fleet Regulatory Assurance - Licensing Tel 601-368-5102

NOTICE: Enclosure 1 to this letter contains Proprietary Information to be withheld from public disclosure per 10 CFR 2.390. Upon separation from Enclosure 1, this letter is DECONTROLLED 1CAN062301 Page 2 of 2 regulations. The proprietary information is identified by text enclosed within double bolded brackets ((Example)). The non-proprietary version is provided in Enclosure 2.

Per the NRC's Safety Evaluation Report (SER), the results of WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval," [ADAMS Accession Number ML11306A084], dated October 2011, can only be used on Babcock &

Wilcox (B&W)-designed ANO-1 after the assumption that an equivalent of 12 heat-up/ cool-down cycles per year of operation can be validated to bound all of its design basis transients that contribute significantly to fatigue crack growth. Enclosure 3 provides the summary of the analysis validating this assumption.

This information is supported by an affidavit, signed by Philip A. Opsal, Manager, Product Licensing, for Framatome Inc. (Framatome, 3315 Old Forest Road, Lynchburg, VA 24501 ), the owner of the information. The affidavit sets forth the basis by which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b )(4) of 10 CFR 2.390 of the Commission's regulations. The affidavit is included in Enclosure 4.

There are no new regulatory commitments established in this submittal.

If there are any questions or if additional information is needed, please contact Riley Keele, Manager, Regulatory Assurance, Arkansas Nuclear One, at 479-858-7826.

Respectfully, Philip Couture Phil Couture PC/rwc Digitally signed by Philip Couture Date: 2023.06.08 15:42:44 -05'00'

Enclosures:

1.

Request for Alternative ANO1-ISl-037 (PROPRIETARY)

2.

Request for Alternative ANO1-ISl-037 (NON-PROPRIETARY)

3.

Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location (Framatome Document 86-9352400-000)

4.

Affidavit cc:

NRC Region IV Regional Administrator NRC Senior Resident Inspector - Arkansas Nuclear One NRC Project Manager - Arkansas Nuclear One

ENCLOSURE 2 1CAN062301 REQUEST FOR ALTERNATIVE ANO1-ISI-037 (NON-PROPRIETARY) 1CAN062301 Page 1 of 8 Request for Alternative for Implementation of Extended Reactor Vessel lnservice Inspection Interval (ANO1-ISI-037)

1. ASME Code Component(s) Affected The affected component is the Arkansas Nuclear One Unit 1 (ANO-1) reactor vessel (RV),

specifically, the following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI (Reference 1) examination categories and item numbers covering examinations of the RV. These examination categories and item numbers are from IWB-2500 and Table IWB-2500-1 of the ASME BPV Code,Section XI.

Category B-A welds are defined as "Pressure Retaining Welds in Reactor Vessel."

Category B-D welds are defined as "Full Penetration Welded Nozzles in Vessels."

Examination Category Item No.

Description B-A B1.10 Shell Welds B-A B1.11 Circumferential Shell Welds B-A B1.12 Longitudinal Shell Welds B-A B1.20 Head Welds B-A B1.21 Circumferential Head Welds B-A B1.30 Shell-to-Flange Weld B-A B1.50 Repair Welds B-A B1.51 Beltline Welds B-D B3.90 Nozzle-to-Vessel Welds B-D B3.100 Nozzle Inside Radius Section (Throughout this request, the above examination categories are referred to as "the subject examinations" and the ASME BPV Code,Section XI, is referred to as "the Code.")

2. Applicable Code Edition and Addenda ASME Code Section XI, 'Rules for Inservice Inspection of Nuclear Power Plant Components,"

2007 Edition though 2008 Addenda (Reference 1).

3. Applicable Code Requirement IWB-2411, Inspection Program, requires volumetric examination of essentially 100% of reactor vessel pressure-retaining welds identified in Table IWB-2500-1 once each 10-year interval. The fifth 10-year inservice inspection (ISI) interval for ANO Unit 1 is scheduled to end on May 30, 2027. The applicable Code for the sixth 10-year ISI interval will be selected in accordance with the requirements of 10 CFR 50.55a.
4. Reason for Request An alternative is requested from the requirement of the IWB-2411 Inspection Program, that volumetric examination of essentially 100% of reactor vessel pressure-retaining Examination Category B-A and B-D welds be performed once each 10-year interval. Extension of the interval between examinations of Category B-A and B-D welds from 10 years to up to 20 years will result in a reduction in man-rem exposure and examination costs.

1CAN062301 Page 2 of 8

5. Proposed Alternative and Basis for Use Entergy proposes not to perform the ASME Code required volumetric examination of the ANO-1 reactor vessel full penetration pressure-retaining Examination Category B-A and B-D welds for the fifth inservice inspection, currently scheduled for 2027. Entergy will perform the fifth ASME Code required volumetric examination of the ANO-1 reactor vessel full penetration pressure-retaining Examination Category B-A and B-D welds in 2036.

The proposed inspection date is a deviation from the latest revised implementation plan, OG-10-238 (Reference 2), since the implementation originally showed the next inspection being performed in 2028 for ANO-1. The impact to the implementation plan in OG-10-238 would increase the number of inspections in 2036 (from one to two) and decrease the number of inspections in 2028 (from five to four). Based on Figure 3 and Figure 4 of OG-10-238, this proposed inspection schedule is considered to have a minor impact on the future inspection plan and the distribution of inspections over time.

In accordance with 10 CFR 50.55a(z)(1), an alternate inspection interval is requested on the basis that the current interval can be revised with negligible change in risk by satisfying the risk criteria specified in Regulatory Guide (RG) 1.174 (Reference 3). The methodology used to conduct this analysis is based on that defined in Reference 4. This study focuses on risk assessments of materials within the beltline and extended beltline regions of the RV wall. The results of the calculations for ANO-1 were compared to those obtained from the Babcock and Wilcox (B&W) pilot plant evaluated in WCAP-16168-NP-A, Revision 3. Appendix A of WCAP-16168-NP-A identifies the parameters to be compared. Demonstrating that the parameters for ANO-1 are bounded by the results of the B&W pilot plant qualifies ANO-1 for an ISI interval extension.

Table 1 below lists the critical parameters investigated in WCAP-16168-NP-A and compares the results of the B&W pilot plant to those of ANO-1. Tables 2 and 3 provide additional information that was requested by the NRC and included in Appendix A of Reference 4.

1CAN062301 Page 3 of 8 Table 1: Critical Parameters for the Application of Bounding Analysis for ANO-1 Parameter Pilot Plant Basis Plant-Specific Basis Additional Evaluation Required?

Dominant Pressurized Thermal Shock (PTS) Transients in the NRC PTS Risk Study are Applicable NRC PTS Risk Study (Reference 5)

PTS Generalization Study (Reference 6)

No Through-Wall Cracking Frequency (TWCF) 4.42E-07 Events per year (Reference 4) 1.88E-11 Events per year (Calculated per Reference 4)

No Frequency and Severity of Design Basis Transients 12 heatup/cooldown cycles per year (Reference 4)

Bounded by 12 heatup/cooldown cycles per year Yes (as required by Reference 4 and summarized in (Reference 10))

Cladding Layers (Single/Multiple)

Single Layer (Reference 4)

Single Layer No Table 2 below provides a summary of the latest reactor vessel inspection for ANO-1 and an evaluation of the recorded indications. This information confirms that satisfactory examinations have been performed on the ANO-1 reactor vessel.

1CAN062301 Page 4 of 8 Table 2: Additional Information Pertaining to Reactor Vessel Inspection for ANO-1 Inspection methodology:

The latest RV ISI for ANO-1 (1R26, 2016) was conducted in accordance with the requirements of Appendix VIII of the ASME Code,Section XI, 2001 Edition through 2003 Addenda (Reference 11). Examinations of Category B-A and B-D welds were performed to the acceptance standards of Section XI, Appendix VIII, 2001 Edition through the 2003 Addenda as modified by 10 CFR 50.55a(b)(2)(xiv, xv and xvi). Future inservice inspections will continue to be performed to ASME Section XI, Appendix VIII methodology Number of past inspections:

Four 10-Year inservice inspections and a pre-service inspection have been performed.

Number of indications found:

There was one indication identified in the beltline and extended beltline regions during the most recently completed inservice inspection. This subsurface indication is located in the lower shell longitudinal weld (Item 12/13 in Table 3). The indication is acceptable per Table IWB-3510-1 of Section XI of the ASME Code. The indication is not within the inner 1/10th or 1 inch of the reactor vessel thickness; therefore, it is inherently acceptable per the requirements of the Alternate PTS Rule, 10 CFR 50.61a (Reference 7).

The fourth 10-year inspection was the first ISI examination that detected the one indication described above. There is no site-specific flaw growth data since this indication was evaluated as acceptable per ASME Section XI Table IWB-3510-1.

Proposed inspection schedule for balance of plant life:

The fifth inservice inspection is scheduled for 2027 (The fifth 10-year ISI interval for ANO Unit 1 is scheduled to end on May 30, 2027). This inspection will instead be performed in 2036 plus or minus one refueling outage. The proposed inspection date is a deviation from the latest revised implementation plan, OG-10-238 (Reference 2), since the implementation plan reflects the next inspection being performed in 2028 for ANO-1. The date for the next ANO-1 vessel ISI is within plus or minus one refueling outage from the date provided in Reference 2, as discussed in Appendix B of the NRC's Safety Evaluation of Reference 4. The impact to the implementation plan in OG-10-238 would increase the number of inspections in 2036 (from one to two) and decrease the number of inspections in 2028 (from five to four). Based on Figure 3 and Figure 4 of OG-10-238, this proposed inspection schedule is considered to have a minor impact on the future inspection plan and the distribution of inspections over time.

1CAN062301 Page 5 of 8 Table 3 summarizes the inputs and outputs for the calculation of through-wall cracking frequency (TWCF).

Table 3: Details of TWCF Calculation for ANO-1 at 54 Effective Full Power Years (EFPY)

Inputs(1)

Above LNBF taper Twall [inches]:

12 Below LNBF taper Twall [inches]:

8.44 No.

Region and Component Description Material Heat No.

Identification Copper

[weight %]

Nickel

[weight %]

R.G. 1.99 Position Chemistry Factor [ºF]

RTNDT(u)

[ºF]

Fluence

[Neutron/cm2, E > 1.0 MeV]

1 Lower Nozzle Beltline Forging (LNBF)

AYN 131 0.03 0.70 1.1 20.0 27.5 1.21E+19 2

Upper Shell Plate 1 C5120-2 0.17 0.55 1.1 122.75

-10 1.33E+19 3

Upper Shell Plate 2 C5114-2 0.15 0.52 1.1 105.6

-10 4

Lower Shell Plate 1 C5120-1 0.17 0.55 1.1 122.75

-10 1.31E+19 5

Lower Shell Plate 2 C5114-1 0.15 0.52 1.1 105.6 0

6 Dutchman Forging (Transition Piece) 125W609VA1

(( ))

(( ))

1.1

(( ))

(( ))

1.61E+17 7

Lower Nozzle Beltline Forging to Upper Shell Circumferential. Weld WF-182-1 0.24 0.63 1.1 177.95

-84.2 1.21E+19 8

Upper Shell to Lower Shell Circumferential Weld WF-112 0.27 0.59 1.1 182.55

-98.0 1.28E+19 9

Lower Shell to Dutchman Forging (Transition Piece) Weld SA-1788

(( ))

(( ))

1.1

(( ))

(( ))

1.61E+17 10 Upper Shell Longitudinal Weld 1 WF-18 0.19 0.57 1.1 167.0

-48.6 1.03E+19 11 Upper Shell Longitudinal Weld 2 WF-18 0.19 0.57 1.1 167.0

-48.6 12 Lower Shell Longitudinal Weld 1 WF-18 0.19 0.57 1.1 167.0

-48.6 1.09E+19 13 Lower Shell Longitudinal Weld 2 WF-18 0.19 0.57 1.1 167.0

-48.6 Outputs Methodology Used to Calculate T30: Regulatory Guide 1.99, Revision 2 (Reference 8)

Controlling Material Region No.

XX RTMAX-XX [°R]

Fluence

[Neutron/cm2, E >1.0 MeV]

FF (Fluence Factor)

T30 [ºF]

TWCF95-XX Limiting Axial Weld - AW 12/13 2.5000 582.09 1.09E+19 1.0241 171.02 0.000E+00 Limiting Plate - PL 2

2.5000 582.15 1.33E+19 1.0793 132.48 7.532E-12 Limiting Forging - FO 1

2.5000 508.23 1.21E+19 1.0532 21.06 5.561E-15 Limiting Circumferential Weld - CW 2/4 2.5000 580.86 1.28E+19 1.0687 131.19 0.000E+00 TWCF95-TOTAL(AWTWCF95-AW + PLTWCF95-PL + FOTWCF95-FO + CWTWCF95-CW):

1.88E-11 Note 1: Material properties are based on ANP-3300 (Reference 9). Fluence projections and material properties not included in Reference 9 were provided by Framatome.

1CAN062301 Page 6 of 8

6. Duration of Proposed Alternative This request is applicable to the ANO-1 inservice inspection program for the fifth and sixth 10-year inspection intervals.
7. Precedents U.S. Nuclear Regulatory Commission (NRC) to Virginia Electric and Power Company, "Surry Power Station Units 1 and 2 - Relief Implementing Extended Reactor Vessel Inspection Interval (TAC Nos. ME8573 and ME8574)," Agencywide Document Access and Management System (ADAMS) Accession Number ML13106A140, dated April 30, 2013.

NRC to Southern Nuclear Operating Company, Inc., "Vogtle Electric Generating Plant, Units 1 and 2 - Request for Alternatives VEGP-ISI-ALT-05 and VEGP-ISI-ALT-06 (TAC Nos. MF2596 and MF2597)," ADAMS Accession Number ML14030A570, dated March 20, 2014.

Duke to NRC, "Catawba Nuclear Station Units 1 and 2: Proposed Relief Request 13-CN-003, Request for Alternative to the Requirement of IWB-2500, Table IWB-2500-1, Category B-A and Category B-D for Reactor Pressure Vessel Welds (TAC Nos. MF1922 and MF1923)," ADAMS Accession Number ML14079A546, dated March 26, 2014.

NRC to Tennessee Valley Authority, "Sequoyah Nuclear Plant, Units 1 and 2 - Requests for Alternatives 13-ISI-1 and 13-ISI-2 to Extend the Reactor Vessel Weld Inservice Inspection Interval (TAC Nos. MF2900 and MF2901)," ADAMS Accession Number ML14188B920, dated August 1, 2014.

NRC to Exelon Generation Company, LLC, "Byron Station, Unit No. 1 - Relief from Requirements of the ASME Code to Extend the Reactor Vessel Inservice Inspection Interval (TAC No. MF3596)," ADAMS Accession Number ML14303A506, dated December 10, 2014.

NRC to Wolf Creek Nuclear Operating Corporation, "Wolf Creek Generating Station -

Request for Relief Nos. I3R-08 and I3R-09 for the Third 10-Year Inservice Inspection Program Interval (TAC Nos. MF3321 and MF3322)," ADAMS Accession Number ML14321A864, dated December 10, 2014.

NRC to Union Electric Company, "Callaway Plant, Unit 1 - Request for Relief I3R-17, Alternative to ASME Code Requirements Which Extends the Reactor Vessel Inspection Interval from 10 to 20 Years (TAC No. MF3876)," ADAMS Accession Number ML15035A148, dated February 10, 2015.

NRC to Exelon Generation Company, LLC, "Braidwood Station, Units 1 and 2 - Request for Relief from the Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) (CAC Nos. MF8191 and MF8192)," ADAMS Accession Number ML17054C255, dated March 15, 2017.

NRC to STP Nuclear Operating Company, "South Texas Project, Units 1 and 2 - Relief from the Requirements of the ASME Code Regarding the Third 10-Year Inservice Inspection Program Interval (EPID L-2018-LLR-0010)," ADAMS Accession Number ML18177A425, dated July 24, 2018.

NRC to Indiana Michigan Power Company, "Donald C. Cook Nuclear Plant, Unit No. 1 -

Approval of Alternative to the ASME Code Regarding Reactor Vessel Weld Examination -

Relief Request ISIR-4-08 (EPID: L-2018-LLR-0106)," ADAMS Accession Number ML18284A310, dated October 26, 2018.

1CAN062301 Page 7 of 8 NRC to Exelon Generation Company, LLC, "R. E. Ginna Nuclear Power Plant - Issuance of Relief Request ISI-18 Regarding Fifth 10-year Inservice Inspection Program Interval (EPID L-2018-LLR-0104)," ADAMS Accession Number ML19100A004, dated April 22, 2019.

NRC to Florida Power & Light Company, "Point Beach Nuclear Plant, Units 1 and 2 -

Approval of Relief Requests 1-RR-13 and 2-RR-13 Regarding Extension of Inspection Interval for Reactor Pressure Welds from 10 to 20 years (EPID L-2019-LLR-0060)," ADAMS Accession Number ML20036F261, dated March 4, 2020.

NRC to Florida Power & Light Company, "St. Lucie Plant, Unit 2 - Authorization of RR#15 Regarding Extension of ASME Requirements Related to Reactor Pressure Vessel Weld Examinations from 10 to 20 Years (EPID L-2020-LLR-0283)," ADAMS Accession Number ML21236A131, dated September 30, 2021.

NRC to Duke Energy Company, "Oconee Nuclear Station, Units 1, 2, and 3 - Authorization and Safety Evaluation for Alternative Reactor Pressure Vessel Inservice Inspection Intervals (EPID L-2021-LLR-0004)," ADAMS Accession Number ML21281A141, dated November 19, 2021.

NRC to Florida Power & Light Company, "Turkey Point Nuclear Generating Unit Nos. 3 and 4 - Authorization of Relief Request Nos. 8 and 9 Regarding Extension of Inspection Interval for Reactor Pressure Vessel Welds (EPID L-2021-LLR-0038)," ADAMS Accession Number ML22123A192, dated May 10, 2022.

8. References
1. ASME Boiler and Pressure Vessel Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components" 2007 Edition with 2008 Addenda, ASME International.
2. OG-10-238, "Revision to the Revised Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP-16168-NP, Revision 1, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval.' PA-MSC-0120," [ADAMS Accession Number ML11153A033], dated July 12, 2010.
3. NRC Regulatory Guide 1.174, Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," U.S. Nuclear Regulatory Commission, [ADAMS Accession Number ML023240437], dated November 2002.
4. Westinghouse Report, WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval," [ADAMS Accession Number ML11306A084], dated October 2011.
5. NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS),"

U.S. Nuclear Regulatory Commission, [ADAMS Accession Number ML15222A848],

dated March 2010.

6. NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS)

Risk Results to Additional Plants," U.S. Nuclear Regulatory Commission, [ADAMS Accession Number ML042880482], dated December 14, 2004.

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

1CAN062301 Page 8 of 8

8. NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, [ADAMS Accession Number ML003740284], dated May 1988.
9. Areva Report ANP-3300, Revision 0, "Arkansas Nuclear One (ANO) Unit 1 Pressure-Temperature Limits at 54 EFPY," [ADAMS Accession Number ML14241A241], dated June 2014.
10. Framatome Document, 86-9352400-000, "Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location."
11. ASME Boiler and Pressure Vessel Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components" 2001 Edition with 2003 Addenda, ASME International.

ENCLOSURE 3 1CAN062301 EQUIVALENT FATIGUE CRACK GROWTH FOR ARKANSAS NUCLEAR ONE UNIT 1 BELTLINE SHELL LOCATION (FRAMATOME DOCUMENT 86-9352400-000)

Page 1 of 10 0402-01-F01 (Rev. 021, 03/12/2018)

PROPRIETARY CALCULATION

SUMMARY

SHEET (CSS)

Document No.

86 9352400 000 Safety Related: Yes No Title Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location PURPOSE AND

SUMMARY

OF RESULTS:

Purpose:

Per the NRCs Safety Evaluation Report (SER), the results of WCAP-16168-NP-A, Revision 3 (Reference [1]) can only be used on B&W-designed Arkansas Nuclear One Unit 1 (ANO-1) after the assumption that an equivalent of 12 heat-up / cool-down cycles per year of operation can be validated to bound all of its design basis transients that contribute significantly to fatigue crack growth. This document provides the summary of the analysis validating this assumption.

Summary of Results: Based on the results of the analysis summarized herein, the equivalent fatigue crack growth (Table 4-2) using 12 equivalent heat-up and cool-down cycles per year is larger than the detailed transient fatigue crack growth from the detailed design transients (Table 4-1);

therefore, the assumption of WCAP-16168-NP-A, Reference [1], for Arkansas Nuclear One Unit 1 that 12 equivalent heat-up and cool-down cycles bound the fatigue crack growth from all the other test, normal / upset and emergency / faulted service level transients is valid.

The maximum RTMAX-FO for ANO-1 forging = 52.3°F (512.0°R).

FRAMATOME INC. PROPRIETARY This document and any information contained herein is the property of Framatome Inc. (Framatome) and is to be considered proprietary and may not be reproduced or copied in whole or in part. This document shall not be furnished to others without the express written consent of Framatome and is not to be used in any way which is or may be detrimental to Framatome. This document and any copies that may have been made must be returned to Framatome upon request.

If the computer software used herein is not the latest version per the EASI list, AP 0402-01 requires that justification be provided.

THE DOCUMENT CONTAINS ASSUMPTIONS THAT SHALL BE VERIFIED PRIOR TO USE THE FOLLOWING COMPUTER CODES HAVE BEEN USED IN THIS DOCUMENT:

CODE/VERSION/REV CODE/VERSION/REV Yes No N / A Controlled Document

Document No. 86-9352400-000 0402-01-F01 (Rev. 021, 03/12/2018)

PROPRIETARY Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location Page 2 Review Method: Design Review (Detailed Check)

Alternate Calculation Does this document establish design or technical requirements? YES NO Does this document contain Customer Required Format? YES NO Signature Block Name and Title (printed or typed)

Signature P/R/A/M and LP/LR Date Pages/Sections Prepared/Reviewed/Approved Martin Kolar, Principal Engineer P

All Pages / All Sections Luziana Matte, Advisory Engineer R

All Pages / All Sections Ryan Hosler, Supervisory Engineer A

All Pages / All Sections Notes: P/R/A designates Preparer (P), Reviewer (R), Approver (A);

LP/LR designates Lead Preparer (LP), Lead Reviewer (LR);

M designates Mentor (M)

In preparing, reviewing and approving revisions, the lead preparer/reviewer/approver shall use All or All except

___ in the pages/sections reviewed/approved. All or All except ___ means that the changes and the effect of the changes on the entire document have been prepared/reviewed/approved. It does not mean that the lead preparer/reviewer/approver has prepared/reviewed/approved all the pages of the document.

With Approver permission, calculations may be revised without using the latest CSS form. This deviation is permitted when expediency and/or cost are a factor. Approver shall add a comment in the right-most column that acknowledges and justifies this deviation.

Project Manager Approval of Customer References and/or Customer Formatting (N/A if not applicable)

Name (printed or typed)

Title (printed or typed)

Signature Date Comments N / A M KOLAR 12/1/2022 LR MATTE 12/1/2022 RS HOSLER 12/1/2022 Controlled Document

Document No. 86-9352400-000 0402-01-F01 (Rev. 021, 03/12/2018)

PROPRIETARY Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location Page 3 Record of Revision Revision No.

Pages/Sections/Paragraphs Changed Brief Description / Change Authorization 000 All Pages / All Sections Initial Issue of the Document Controlled Document

Document No. 86-9352400-000 PROPRIETARY Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location Page 4 Table of Contents Page SIGNATURE BLOCK................................................................................................................................ 2 RECORD OF REVISION.......................................................................................................................... 3 LIST OF TABLES..................................................................................................................................... 5 1.0 PURPOSE..................................................................................................................................... 6 2.0 ASSUMPTIONS............................................................................................................................ 6 2.1 Unverified Assumptions..................................................................................................................... 6 2.2 Justified Assumptions........................................................................................................................ 6 3.0 INPUTS......................................................................................................................................... 7 3.1 Temperature / Pressure Transients.................................................................................................. 7 4.0 RESULTS...................................................................................................................................... 8 4.1 Transient Fatigue Crack Growth Summary....................................................................................... 8 4.2 Equivalent Crack Growth................................................................................................................... 8 4.3 Transient Crack Growth Contribution Ratio...................................................................................... 8

5.0 CONCLUSION

.............................................................................................................................. 9 5.1 Fatigue Crack Growth....................................................................................................................... 9 5.2 RTMAX Temperature........................................................................................................................... 9

6.0 REFERENCES

............................................................................................................................ 10 Controlled Document

Document No. 86-9352400-000 PROPRIETARY Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location Page 5 List of Tables Page Table 3-1: List of Transient Events Considered for ANO-1....................................................................... 7 Table 4-1: Transient Fatigue Crack Growth Summary............................................................................. 8 Table 4-2: Equivalent Fatigue Crack Growth............................................................................................ 8 Table 4-3: Crack Growth Relative Contribution........................................................................................ 9 Controlled Document

Document No. 86-9352400-000 PROPRIETARY Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location Page 6 1.0 PURPOSE This document is part of the Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval PWR Owners Group project PA-MSC-0943, B&W Site Specific Fatigue Crack Growth Evaluation.

Westinghouse Electric Company (WEC) performed a Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval analysis documented in WCAP-16168-NP-A, Revision 3 and US NRC approved this work in 2011 (Reference [1]).

Per the NRCs Safety Evaluation Report (SER), the results of WCAP-16168-NP-A, Revision 3 (Reference [1])

can only be used on B&W designed Arkansas Nuclear One Unit 1 if the following requirements outlined in Appendix B (items 3 and 4) of Reference [1] are addressed:

Item 3:

Licensees must verify that the fatigue crack growth of 12 heat-up/cool-down transients per year bound the fatigue crack growth for all of its design basis transients.

Licensees must identify the design basis transients that contribute to significant fatigue crack growth.

Item 4:

If the subject plant has reactor vessel forgings that are susceptible to underclad cracking with RTMAX-FO values exceeding 240°F, then the WCAP analyses are not applicable. The licensees must submit a plant-specific evaluation for any extension to the 10-year inspection interval for ASME Code,Section XI, Category B-A and B-D RPV welds.

This document is a summary of an analysis validating item 3 and item 4 for a 20 year in-service inspection (ISI) interval.

This current summary document is a non-proprietary document in support of a relief request being prepared by Westinghouse.

2.0 ASSUMPTIONS 2.1 Unverified Assumptions No unverified assumption was made during the preparation of this document.

2.2 Justified Assumptions

1. For the purpose of this document, based on the chemical composition, the cladding weld material is taken as 18Cr-8Ni (Type 304 stainless steel).
2. Due to the similarities in the base metal material chemical compositions, the reactor vessel shell material is assumed as low-alloy steel SA-508 Class 2 for the entire region of interest.

Controlled Document

Document No. 86-9352400-000 PROPRIETARY Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location Page 7 3.0 INPUTS 3.1 Temperature / Pressure Transients The inside surface of the reactor vessel is subjected to transient loads in the form of primary coolant cold leg temperatures and pressures as defined by the reactor coolant system functional specification. The number of applicable transient cycles corresponds to 20 years of operation. In order to form complete stress cycles, these individual transients are combined into transient groups. These combined transients are used to calculate the cyclic variations of stress intensity factor that produce fatigue crack growth over the life of the plant.

Table 3-1 lists transient events considered in this evaluation.

Table 3-1: List of Transient Events Considered for ANO-1 Transient Number(1)

TransientName/Description Category

1A Heatupfrom0%to8%Power normal

1B Cooldownfrom8%to0%Power normal

1C TechnicalSpecificationCooldown emergency

2A PowerChangefrom0%to15%

normal

2B PowerChangefrom15%to0%

normal

3 PowerLoadingfrom8%to100%

normal

4 PowerUnloadingfrom100%to8%

normal

5 10%StepLoadIncrease normal

6 10%StepLoadDecrease normal

7 StepLoadReductionfrom100%to8%Power upset

8A ReactorTripTypeA(LossofRCFlow) upset

8B ReactorTripTypeB(TurbineTrip) upset

8C ReactorTripTypeC(LossofMFWFlow) upset

9 RapidDepressurization upset

10 ChangeofFlow upset

11(2)

RodWithdrawalAccident upset

12 Hydrotest test

14 ControlRodDrop upset

15(2)

LossofStationPower upset

16 SteamLineFailure faulted

17A(2)

LossofFeedwatertoOneSteamGenerator upset

17B(2)

StuckOpenTurbineBypassValve emergency

21(2)

LossofCoolant faulted

22A HighPressureInjectionTest normal

22B CoreFloodCheckValveTest normal

23 SteamGeneratorFilling,Draining,FlushingandCleaning normal

24 RCPRestartwithVoidsintheRCS emergency

25 RefillofHot,Dry,DepressurizedSG upset

26 PressurizerBypassSprayInterruption/Restoration test

27 NaturalCirculationCooldown

Note(1):Transientsthatdonotcontributetothegrowthofthepostulatedflawarenotconsidered.

Note(2):partofReactorTripgroup.

ThecalculationoffinalcracksizefollowsmethodologyofSectionA5200,Reference[2].

CrackgrowthratesarebasedonSectionA4300ofReference[2].

MFW=MainFeedWater,RC(S)=ReactorCoolant(System),RCP=ReactorCoolantPump

SG=SteamGenerator

Controlled Document

Document No. 86-9352400-000 PROPRIETARY Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location Page 8 4.0 RESULTS 4.1 Transient Fatigue Crack Growth Summary The initial (3.7% of the shell wall thickness including cladding) and final (for 20 years of operation) flaw depths, considering all the design transients listed under Table 3-1 are listed in Table 4-1. The initial flaw size bounds the as-found flaw from the latest in-service inspection (ISI) for ANO-1. The analysis considers a flaw length over depth aspect ratio l/a = 10 and the flaw aspect ratio is maintained during the entire fatigue crack growth.

Table 4-1: Transient Fatigue Crack Growth Summary

time a

[year]

[inch]

0 0.3191

20 0.3272

4.2 Equivalent Crack Growth The result of the calculation shown in this Section is an acceptance criterion used to compare with the transient fatigue crack growth calculated in Section 4.1. The final crack size is calculated for 20 years of operation with 12 equivalent heat-up / cool-down cycles per one year.

Table 4-2: Equivalent Fatigue Crack Growth time cycles a

[year]

[inch]

0 0

0.3191 20 240 0.3273 4.3 Transient Crack Growth Contribution Ratio This section presents results from the detailed transient fatigue crack growth calculation summarized in Table 4-1 for 20 years of operation. The lines in Table 4-3 are sorted by their relative contribution to total fatigue crack growth.

Table 4-3 lists the order of the transients from the highest to the lowest contributors. The highest contributor to crack propagation is from the power change transient, followed by power loading transient, and followed by heat-up / cool-down transient, which together contribute to nearly 97% of the total crack growth and this fatigue crack growth occurs in water environment.

Remaining transients are considered negligible with a total contribution of approximately 3% of the total crack growth.

Controlled Document

Document No. 86-9352400-000 PROPRIETARY Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location Page 9 Table 4-3: Crack Growth Relative Contribution

transientname participation

%ratio

powerchange

~97%

powerloading

heatup/cooldown

rodwithdrawal

~3%

flowchange

powerunloading

trip

steploadincrease/decrease

feedwaterloss

5.0 CONCLUSION

5.1 Fatigue Crack Growth The equivalent fatigue crack growth (Table 4-2) using 12 equivalent heat-up / cool-down cycles per year is larger than the fatigue crack growth from the detailed design transients (Table 4-1). Therefore, the requirements outlined in Item 3 of Appendix B of WCAP-16168-NP-A, Reference [1], for Arkansas Nuclear One Unit 1 are met.

Furthermore, power change, power loading and heat-up / cool-down transients were identified as major contributors to the fatigue crack growth.

This fulfills the intent of Item 3, Appendix B of WCAP-16168-NP-A, Reference [1] for Arkansas Nuclear One Unit 1.

5.2 RTMAX Temperature The maximum RTMAX-FO value for ANO-1 forgings = 52.3°F (512.0°R). This fulfills the intent of Item 4, Appendix B of WCAP-16168-NP-A, Reference [1] for Arkansas Nuclear One Unit 1.

Controlled Document

Document No. 86-9352400-000 PROPRIETARY Equivalent Fatigue Crack Growth for Arkansas Nuclear One Unit 1 Beltline Shell Location Page 10

6.0 REFERENCES

1.

Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval, WCAP-16168-NP-A, Revision 3, October 2011, Westinghouse Electric Company LLC

2.

ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2013 Edition Controlled Document

ENCLOSURE 4 1CAN062301 AFFIDAVIT (3 Pages)

A F F I D A V I T

1.

My name is Philip A. Opsal. I am Manager, Product Licensing for Framatome Inc. (formally known as AREVA Inc.), and as such I am authorized to execute this Affidavit.

2.

I am familiar with the criteria applied by Framatome to determine whether certain Framatome information is proprietary. I am familiar with the policies established by Framatome to ensure the proper application of these criteria.

3. I am familiar with the Framatome information contained in Enclosure 1 to Entergy Letter to USNRC 1CAN062301, Request for Alternative for Implementation of Extended Reactor Vessel Inservice Inspection Interval (ANO1-ISI-037), referred to herein as this Document. Information contained in Table 3 of this Document, identified in bolded text in double square brackets, has been classified by Framatome as proprietary in accordance with the policies established by Framatome for the control and protection of proprietary and confidential information.
4.

This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by Framatome and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5.

This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is

made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) Trade secrets and commercial or financial information.

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The following criteria are customarily applied by Framatome to determine whether information should be classified as proprietary:

(a)

The information reveals details of Framatomes research and development plans and programs or their results.

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Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

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The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for Framatome.

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The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for Framatome in product optimization or marketability.

(e)

The information is vital to a competitive advantage held by Framatome, would be helpful to competitors to Framatome, and would likely cause substantial harm to the competitive position of Framatome.

The information in this Document is considered proprietary for the reasons set forth in paragraphs 6(d), and 6(e) above.

7.

In accordance with Framatomes policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside Framatome only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8.

Framatome policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9.

The foregoing statements are true and correct to the best of my knowledge, information, and belief.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on June 8, 2023.

Philip A. Opsal

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