RA-20-0328, Request for Alternative for Implementation of Extended Reactor Vessel Inservice Inspection Intervals

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Request for Alternative for Implementation of Extended Reactor Vessel Inservice Inspection Intervals
ML21019A276
Person / Time
Site: Oconee  Duke Energy icon.png
Issue date: 01/19/2021
From: Burchfield J
Duke Energy Carolinas
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-20-0328 86-931520-00
Download: ML21019A276 (35)
v • d • e


Text

( ., DUKE J. Ed Burchfield, Jr.

Vice President ENERGY Oconee Nuclear Station Duke Energy ON01 VP I 7800 Rochester Hwy Seneca, SC 29672 0, 864.873.3478 f 864.873.4208 Ed.Burchlielcl@duke-energy.com Serial: RA-20-0328 10 CFR 50.55a January 19, 2021 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 OCONEE NUCLEAR STATION, UNITS NO. 1, 2 AND 3 DOCKET NO. 50-269, 270 AND 287 I RENEWED LICENSE NO. DPR-38, DPR-47 AND DPR-55

SUBJECT:

Request for Alternative for Implementation of Extended Reactor Vessel lnservice Inspection Intervals for Oconee Units 1, 2, and 3 Pursuant to 10 CFR 50.55a(z)(1}, Duke Energy Carolinas, LLC (Duke Energy) requests NRC approval to extend the inservice inspection interval for the Oconee Nuclear Station (ONS) Units 1, 2, and 3 reactor pressure vessel (RPV) weld examinations from 2024 to 2034.

Duke Energy proposes to implement an alternative to the requirement of ASME Section XI IWB-2411, Inspection Program, that volumetric examination of RPV Examination categories 8-A and 8-D be performed once each 10-year ISi interval. The current fifth ISi interval ends on July 15, 2024. Duke Energy proposes to perform the fifth ASME Section XI Category B-A and 8-D examinations in the sixth ISi interval by no later than 2034. The Enclosure to this letter provides the basis for the proposed alternative. The Attachment to this letter provides supporting documentation for the proposed alternative.

Duke Energy requests NRC review and approval of the proposed alternative by January 2022.

This letter contains no new regulatory commitments. If you have questions concerning this request, please contact Art Zaremba, Director - Fleet Licensing, at (980) 373-2062.

Sincerely, Je1A J. Ed Burchfield, Jr.

Vice President Oconee Nuclear Station

RA-20-0328 Page 2

Enclosure:

1. Request for Alternative for Implementation of Extended Reactor Vessel Inservice Inspection Intervals for Oconee Units 1, 2, and 3.

Attachment:

1. Framatome Document 86-9315280-000 Equivalent Fatigue Crack Growth for Oconee Units 1, 2 and 3 Beltline Shell Location cc : L. Dudes, Regional Administrator USNRC Region II J. Nadel, USNRC Senior Resident Inspector - ONS S. Williams, NRR Project Manager - ONS

Enclosure 1 Duke Energy Carolinas, LLC Oconee Nuclear Station, Units 1, 2, and 3 Request for Alternative RA-20-0328 Request for Alternative in Accordance with 10 CFR 50.55a(z)(1)

Request for Alternative for Implementation of Extended Reactor Vessel Inservice Inspection Intervals for Oconee Units 1, 2, and 3

10 CFR 50.55a Relief Request RA-20-0328 Enclosure 1 Page 1 of 21 Proposed Alternative for Oconee Unit 1 In Accordance with 10 CFR 50.55a(z)(1)

-Alternative Provides Acceptable Level of Quality and Safety-

1. ASME Code Component(s) Affected The affected component is the Oconee Unit 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 B1.11 Circumferential Shell Welds B-A B1.12 Longitudinal Shell Welds B-A B1.21 Circumferential Head Welds B-A B1.30 Shell-to-Flange Weld 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 through 2008 Addenda (Reference 1) is applicable for Oconee Unit 1 Fifth Inservice Inspection Interval (ISI), which started on July 15, 2014 and is scheduled to end on July 15, 2024. Volumetric examinations shall be performed for the Sixth ISI interval in accordance with requirements of 10 CFR 50.55a and the Edition and Addenda of the ASME Code,Section XI, applicable at the time of the examination.

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 ISI interval.

10 CFR 50.55a Relief Request RA-20-0328 Enclosure 1 Page 2 of 21

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 ISI 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 radiological exposure and examination costs.

5. Proposed Alternative and Basis for Use

Duke Energy proposes not to perform the ASME Code required volumetric examination of the Oconee Unit 1 reactor vessel full penetration pressure-retaining Examination Category B-A and B-D welds for the fifth inservice inspection, currently scheduled for Fall 2022. Duke Energy will perform the fifth volumetric examination of the Oconee Unit 1 reactor vessel full penetration pressure-retaining Examination Category B-A and B-D welds during the Sixth ISI interval in 2032. The proposed inspection date is consistent with latest revised implementation plan, OG-10-238 (Reference 2).

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 1.174 (Reference 3).

The methodology used to conduct this analysis is based on that defined in the study WCAP-16168-NP-A, Revision 3, Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval (Reference 4).

This study focuses on risk assessments of materials within the beltline region of the RV wall. The results of the calculations for Oconee Unit 1 were compared to those obtained from the Babcock & Wilcox pilot plant evaluated in WCAP-16168-NP-A, Revision 3. Appendix A of the WCAP identifies the parameters to be compared. Demonstrating that the parameters for Oconee Unit 1 are bounded by the results of the Babcock & Wilcox pilot plant qualifies Oconee Unit 1 for an ISI interval extension.

10 CFR 50.55a Relief Request RA-20-0328 Enclosure 1 Page 3 of 21 Table 1 below lists the critical parameters investigated in the WCAP and compares the results of the Babcock & Wilcox pilot plant to those of Oconee Unit 1. Tables 2 and 3 provide additional information that was requested by the NRC and included in Appendix A of Reference 4.

Table 1: Critical Parameters for the Application of Bounding Analysis for Oconee Unit 1 Additional Pilot Plant Parameter Plant-Specific Basis Evaluation Basis Required?

Dominant Pressurized Thermal Shock NRC PTS Risk Study PTS Generalization (PTS) Transients in the NRC PTS Risk No (Reference 5) Study (Reference 6)

Study are Applicable 1.40E-10 Events per Through-Wall Cracking Frequency 4.42E-07 Events per year year (Calculated per No (TWCF) (Reference 4)

Reference 4)

Yes Bounded by 12 (as required by Frequency and Severity of Design Basis 12 heatup/cooldown cycles heatup/cooldown Reference 4 and Transients per year (Reference 4) cycles per year 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 Oconee Unit 1 and an evaluation of the recorded indications. This information confirms that satisfactory examinations have been performed on the Oconee Unit 1 reactor vessel.

10 CFR 50.55a Relief Request RA-20-0328 Enclosure 1 Page 4 of 21 Table 2: Additional Information Pertaining to Reactor Vessel Inspection for Oconee Unit 1 The latest ISI for Oconee Unit 1 was conducted in accordance with the ASME Code,Section XI, 1998 Edition, with 2000 Addenda. Examinations of Category B-A and B-D Inspection welds were performed to ASME Section XI, Appendix VIII, 1998 Edition with 2000 methodology: 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 Four inservice inspections have been performed.

inspections:

There were 39 indications identified in the beltline and extended beltline region of the RV during the last ISI. The subsurface indications are located in the Intermediate to Upper Shell, Upper Shell to Lower Shell, Lower Shell to Dutchman circumferential welds and the Intermediate Shell longitudinal welds (Items 9, 10, 11 and 12, respectively in Table 3). All indications are acceptable per Table IWB-3510-1 of Section XI of the ASME Code. Ten of these indications are within the inner 1/10th or 1 inch of the reactor vessel thickness and required further evaluation. These indications are acceptable per the requirements of the Alternate PTS Rule, 10 CFR 50.61a (Reference 7).

A disposition of the ten flaws against the limits of the Alternate PTS Rule is shown in the tables below. Four flaws are located within the weld material and six flaws are located within the plate material of the reactor vessel.

Scaled Maximum number of flaws per 1899 inches of inside Number of Through-Wall Extent, TWE surface weld length of the inspection volume that are greater Oconee Unit 1 (in) than or equal to TWEMIN and less than TWEMAX. This flaw Flaws Evaluated I TWEMIN I TWEMAX I density does not include underclad cracks in forgings.

I (Axial/Circ.)

I 0 0.075 No Limit 0 0.075 0.475 316 4 (1/3)

Number of 0.125 0.475 172 3 (1/2) 0.175 0.475 43 1 (1/0) indications found: 0.225 0.475 16 0 0.275 0.475 7 0 0.325 0.475 5 0 0.375 0.475 2 0 0.425 0.475 1 0 0.475 Infinite 0 0 Scaled Maximum number of flaws per 16,562 square-inches Through-Wall Extent, Number of of inside surface area in the inspection volume that are TWE (in) Oconee Unit 1 greater than or equal to TWEMIN and less than TWEMAX.

Flaws Evaluated This flaw density does not include underclad cracks in TWEMIN TWEMAX (Axial/Circ.)

forgings.

0 0.075 No Limit 1 (0/1) 0.075 0.375 133 5 (1/4) 0.125 0.375 52 3 (1/2) 0.175 0.375 14 1 (1/0) 0.225 0.375 4 1 (1/0) 0.275 0.375 1 1 (1/0) 0.325 0.375 0 0 0.375 Infinite 0 0 Proposed inspection The fifth inservice inspection is currently scheduled for Fall 2022. This inspection will schedule for instead be performed during the 2032 refueling outage. The proposed inspection date is balance of plant consistent with the latest revised implementation plan, OG-10-238 (Reference 2).

life:

10 CFR 50.55a Relief Request RA-20-0328 Enclosure 1 Page 5 of 21 Table 3 summarizes the inputs and outputs for the calculation of through-wall cracking frequency (TWCF).

Table 3: Details of TWCF Calculation for Oconee Unit 1 at 60 Years of Operation or 54 Effective Full Power Years (EFPY)

Inputs(1)

Twall [inches]: 8.628 Copper Nickel Fluence Material R.G. 1.99 Chemistry RTNDT(u)

No. Region and Component Description Material Heat No. [weight [weight [n/cm2, Identification Position Factor [ºF] [ºF]

%]  %] E > 1.0 MeV]

1 Lower Nozzle Belt Forging AHR-54 ZV-2861 0.155 0.65 1.1 119.3 3 1.72E+18 2 Intermediate Shell Plate C2197-2 C2197-2 0.15 0.50 1.1 104.5 1 1.39E+19 3 Upper Shell Plate C3265-1 C3265-1 0.10 0.50 1.1 65.0 1 1.58E+19 4 Upper Shell Plate C3278-1 C3278-1 0.12 0.60 1.1 83.0 1 1.58E+19 5 Lower Shell Plate C2800-1 C2800-1 0.11 0.63 1.1 74.5 1 1.56E+19 6 Lower Shell Plate C2800-2 C2800-2 0.11 0.63 1.1 74.5 1 1.56E+19 7 Dutchman Forging 122S347VA1 122S347VA1 0.11 0.66 1.1 74.9 3 1.33E+17 Lower Nozzle Belt to Intermediate Shell Circ.

8 SA-1135 61782, Flux Lot 8457 0.23 0.52 1.1 157.4 -5 1.72E+18 Weld (100%)

(2) 9a Intermediate to Upper Shell Circ. Weld (61%) SA-1229 71249, Flux Lot 8492 0.23 0.59 1.1 167.6 10 1.41E+19 9b Intermediate to Upper Shell Circ. Weld (39%) WF-25 299L44, Flux Lot 8650 0.34 0.68 1.1 220.6(2) -7 1.41E+19 10 Upper to Lower Shell Circ. Weld (100%) SA-1585 72445, Flux Lot 8597 0.22 0.54 1.1 158.0 -5 1.53E+19 11 Lower Shell to Dutchman Circ. Weld (100%) WF-9 72445, Flux Lot 8632 0.22 0.54 1.1 158.0 -5 1.33E+17 12 Intermediate Shell Long. Welds (100%) SA-1073 1P0962, Flux Lot 8445 0.21 0.64 1.1 170.6 -5 1.07E+19 13 Upper Shell Long. Welds (100%) SA-1493 8T1762, Flux Lot 8578 0.19 0.57 1.1 152.4 -5 1.14E+19 14 Lower Shell Long. Weld (100%) SA-1426 8T1762, Flux Lot 8553 0.19 0.57 1.1 152.4 -5 1.27E+19 15 Lower Shell Long. Weld (100%) SA-1430 8T1762, Flux Lot 8553 0.19 0.57 1.1 152.4 -5 1.27E+19 Outputs Methodology Used to Calculate T30: Regulatory Guide 1.99, Revision 2 (Reference 8)

Fluence FF RTMAX-XX Controlling Material Region No. XX [n/cm2, E (Fluence T30 [ºF] TWCF95-XX

[°R]

>1.0 MeV] Factor)

Limiting Axial Weld - AW 12 2.4790 628.50 1.0700E+19 1.0189 173.83 2.800E-12 Limiting Plate - PL 2 2.5000 574.73 1.3900E+19 1.0914 114.06 2.352E-13 Limiting Circumferential Weld - CW 9a, 9b 1.9822 711.31 1.4100E+19 1.0954 241.64 6.668E-11 Limiting Forging - FO 1 2.5000 526.37 1.7200E+18 0.5340 63.70 2.375E-15 TWCF95-TOTAL = (AWTWCF95-AW + PLTWCF95-PL + CWTWCF95-CW + FOTWCF95-FO): 1.40E-10 Note 1: Material properties are based on WCAP-17571-NP (Reference 9). Fluence projections and material properties not included in Reference 9 were provided by Framatome.

Note 2: The intermediate to upper shell circumferential weld seam was fabricated using two weld heats. The more limiting initial RTNDT value and Chemistry Factor (CF) values between the two weld heats were used for these materials in the subsequent TWCF calculations.

10 CFR 50.55a Relief Request RA-20-0328 Enclosure 1 Page 6 of 21

6. Duration of Proposed Alternative

This request is applicable to the Oconee Unit 1 inservice inspection program for the fifth and sixth inspection intervals.

7. Precedents

  • Surry Power Station Units 1 and 2 - Relief Implementing Extended Reactor Vessel Inspection Interval (TAC Nos. ME8573 and ME8574), dated April 30, 2013, Agencywide Document Access and Management System (ADAMS) Accession Number ML13106A140.
  • 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), dated March 20, 2014, ADAMS Accession Number ML14030A570.
  • 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), dated August 1, 2014, ADAMS Accession Number ML14188B920.
  • Byron Station, Unit No. 1 - Relief from Requirements of the ASME Code to Extend the Reactor Vessel Inservice Inspection Interval (TAC No. MF3596), dated December 10, 2014, ADAMS Accession Number ML14303A506.
  • 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), dated December 10, 2014, ADAMS Accession Number ML14321A864.
  • 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), dated February 10, 2015, ADAMS Accession Number ML15035A148.
  • 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), dated March 15, 2017, ADAMS Accession Number ML17054C255.
  • 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), dated July 24, 2018, ADAMS Accession Number ML18177A425.
  • 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), dated October 26, 2018, ADAMS Accession Number ML18284A310.
  • 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), dated April 22, 2019, ADAMS Accession Number ML19100A004.

10 CFR 50.55a Relief Request RA-20-0328 Enclosure 1 Page 7 of 21

  • 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), dated March 4, 2020, ADAMS Accession Number ML20036F261.
8. References
1. ASME Boiler and Pressure Vessel Code,Section XI, 2007 Edition through 2008 Addenda, American Society of Mechanical Engineers, New York.
2. PWROG Letter 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, July 12, 2010, (ADAMS Accession Number ML11153A033).
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, November 2002, (ADAMS Accession Number ML023240437).
4. Westinghouse Report, WCAP-16168-NP-A, Revision 3, Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval, October 2011, (ADAMS Accession Number ML11306A084).
5. NUREG-1874, Recommended Screening Limits for Pressurized Thermal Shock (PTS), U.S.

Nuclear Regulatory Commission, March 2010, (ADAMS Accession Number ML15222A848).

6. NRC Letter Report, Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants, U.S. Nuclear Regulatory Commission, December 14, 2004 (ADAMS Accession Number ML042880482).
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.
8. NRC Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S. Nuclear Regulatory Commission, May 1988, (ADAMS Accession Number ML003740284).
9. Westinghouse Report, WCAP-17571-NP, Revision 0, Oconee Units 1, 2, and 3 Reactor Pressure Vessel Integrity Program Plans, June 2012.
10. Framatome Document, 86-9315280-000, Equivalent Fatigue Crack Growth for Oconee 1, 2 and 3 Units Beltline Shell Location, September 16, 2020.

10 CFR 50.55a Relief Request RA-20-0328 Enclosure 1 Page 8 of 21 Proposed Alternative for Oconee Unit 2 In Accordance with 10 CFR 50.55a(z)(1)

-Alternative Provides Acceptable Level of Quality and Safety-

1. ASME Code Component(s) Affected The affected component is the Oconee Unit 2 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 B1.11 Circumferential Shell Welds B-A B1.21 Circumferential Head Welds B-A B1.30 Shell-to-Flange Weld 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 through 2008 Addenda (Reference 1) is applicable for Oconee Unit 2 Fifth Inservice Inspection Interval (ISI), which started on July 15, 2014 and is scheduled to end on July 15, 2024. Volumetric examinations shall be performed for the Sixth ISI interval in accordance with requirements of 10 CFR 50.55a and the Edition and Addenda of the ASME Code,Section XI, applicable at the time of the examination.

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 ISI interval.

10 CFR 50.55a Relief Request RA-20-0328 Enclosure 1 Page 9 of 21

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 ISI 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 radiological exposure and examination costs.

5. Proposed Alternative and Basis for Use

Duke Energy proposes not to perform the ASME Code required volumetric examination of the Oconee Unit 2 reactor vessel full penetration pressure-retaining Examination Category B-A and B-D welds for the fifth inservice inspection, currently scheduled for Fall 2023. Duke Energy will perform the fifth volumetric examination of the Oconee Unit 2 reactor vessel full penetration pressure-retaining Examination Category B-A and B-D welds during the Sixth ISI interval in 2033. The proposed inspection date is consistent with the latest revised implementation plan, OG-10-238 (Reference 2).

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 1.174 (Reference 3).

The methodology used to conduct this analysis is based on that defined in the study WCAP-16168-NP-A, Revision 3, Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval (Reference 4).

This study focuses on risk assessments of materials within the beltline region of the RV wall. The results of the calculations for Oconee Unit 2 were compared to those obtained from the Babcock & Wilcox pilot plant evaluated in WCAP-16168-NP-A, Revision 3. Appendix A of the WCAP identifies the parameters to be compared. Demonstrating that the parameters for Oconee Unit 2 are bounded by the results of the Babcock & Wilcox pilot plant qualifies Oconee Unit 2 for an ISI interval extension.

10 CFR 50.55a Relief Request RA-20-0328 Enclosure 1 Page 10 of 21 Table 1 below lists the critical parameters investigated in the WCAP and compares the results of the Babcock & Wilcox pilot plant to those of Oconee Unit 2. Tables 2 and 3 provide additional information that was requested by the NRC and included in Appendix A of Reference 4.

Table 1: Critical Parameters for the Application of Bounding Analysis for Oconee Unit 2 Additional Pilot Plant Parameter Plant-Specific Basis Evaluation Basis Required?

Dominant Pressurized Thermal Shock NRC PTS Risk Study PTS Generalization (PTS) Transients in the NRC PTS Risk No (Reference 5) Study (Reference 6)

Study are Applicable 2.82E-11 Events per Through-Wall Cracking Frequency 4.42E-07 Events per year (Calculated per No (TWCF) year (Reference 4)

Reference 4)

Yes 12 heatup/cooldown Bounded by 12 (as required by Frequency and Severity of Design Basis cycles per year heatup/cooldown Reference 4 and Transients (Reference 4) cycles per year summarized in Reference 10)

Single Layer Cladding Layers (Single/Multiple) Single Layer No (Reference 4)

Table 2 below provides a summary of the latest reactor vessel inspection for Oconee Unit 2 and an evaluation of the recorded indications. This information confirms that satisfactory examinations have been performed on the Oconee Unit 2 reactor vessel.

10 CFR 50.55a Relief Request RA-20-0328 Enclosure 1 Page 11 of 21 Table 2: Additional Information Pertaining to Reactor Vessel Inspection for Oconee Unit 2 The latest ISI for Oconee Unit 2 was conducted in accordance with the ASME Code,Section XI, 1998 Edition, with 2000 Addenda. Examinations of Category B-A and B-D welds were Inspection performed to ASME Section XI, Appendix VIII, 1998 Edition with 2000 Addenda as methodology:

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 Four inservice inspections have been performed.

inspections:

24 indications were identified in the beltline and extended beltline region of the RV during the last inservice inspection. Two of the indications are surface indications (one on the inside surface and one on the outside surface of the RV) and 22 of the indications are subsurface indications. The indications are located in the Lower Nozzle Belt to Upper Shell, Upper Shell to Lower Shell and Lower Shell to Dutchman circumferential welds (Items 5, 6 and 7, respectively in Table 3). All indications are acceptable per Table IWB-3510-1 of Section XI of the ASME Code. Four of these indications are within the inner 1/10th or 1 inch of the reactor vessel thickness and required further evaluation. These indications are acceptable per the requirements of the Alternate PTS Rule, 10 CFR 50.61a (Reference 7).

A disposition of the four flaws against the limits of the Alternate PTS Rule is shown in the tables below. The four flaws evaluated are located within the forging material of the reactor vessel.

Number of indications found: Scaled Maximum number of flaws per 9691 square-inches Through-Wall Extent, Number of of inside surface area in the inspection volume that are TWE (in) Oconee Unit 2 greater than or equal to TWEMIN and less than TWEMAX.

Flaws Evaluated This flaw density does not include underclad cracks in TWEMIN TWEMAX (Axial/Circ.)

forgings.

0 0.075 No Limit 0 0.075 0.375 78 4 (2/2) 0.125 0.375 30 4 (2/2) 0.175 0.375 8 3 (2/1) 0.225 0.375 2 2 (1/1) 0.275 0.375 0 0 0.325 0.375 0 0 0.375 Infinite 0 0 Proposed inspection The fifth inservice inspection is currently scheduled for Fall 2023. This inspection is schedule for proposed to be performed during the 2033 refueling outage. The proposed inspection date is balance of plant consistent with the latest revised implementation plan, OG-10-238 (Reference 2).

life:

10 CFR 50.55a Relief Request RA-20-0328 Enclosure 1 Page 12 of 21 Table 3 summarizes the inputs and outputs for the calculation of through-wall cracking frequency (TWCF).

Table 3: Details of TWCF Calculation for Oconee Unit 2 at 60 Years of Operation or 54 Effective Full Power Years (EFPY)

Inputs(1)

Twall [inches]: 8.628 Region and Component Material Material Copper Nickel R.G. 1.99 Chemistry RTNDT(u) Fluence [n/cm2, No.

Description Identification Heat No. [weight %] [weight %] Position Factor [ºF] [ºF] E > 1.0 MeV]

1 Lower Nozzle Belt Forging AMX 77 123T382 0.13 0.76 1.1 95.0 3 1.35E+19 2 Upper Shell Forging AAW 163 3P2359 0.04 0.75 1.1 26.0 20 1.50E+19 3 Lower Shell Forging AWG 164 4P1885 0.02 0.80 1.1 20.0 20 1.49E+19 4 Dutchman Forging 122T293VA1 122T293VA1 0.13 0.72 1.1 94.0 3 1.26E+17 5 Lower Nozzle Belt to Upper Shell WF-154 406L44, Flux 0.27 0.59 1.1 182.6 -5 1.35E+19 Circumferential Weld (100%) Lot 8720 6 Upper Shell to Lower Shell WF-25 299L44, Flux 0.34 0.68 1.1 220.6 -7 1.45E+19 Circumferential Weld (100%) Lot 8650 7 Lower Shell to Dutchman WF-112 406L44, Flux 0.27 0.59 1.1 182.6 -5 1.26E+17 Circumferential Weld (100%) Lot 8688 Outputs Methodology Used to Calculate T30: Regulatory Guide 1.99, Revision 2 (Reference 8)

FF Controlling Material Fluence [n/cm2, XX RTMAX-XX [°R] (Fluence T30 [ºF] TWCF95-XX Region No. (From Above) E >1.0 MeV]

Factor)

Limiting Circumferential Weld - CW 6 2.0740 695.99 1.4500E+19 1.1030 243.32 1.346E-11 Limiting Forging - FO 1 2.5000 565.59 1.3500E+19 1.0834 102.92 1.054E-13 TWCF95-TOTAL = (CWTWCF95-CW + FOTWCF95-FO): 2.82E-11 Note 1: Material properties are based on WCAP-17571-NP (Reference 9). Fluence projections and material properties not included in Reference 9 were provided by Framatome.

10 CFR 50.55a Relief Request RA-20-0328 Enclosure 1 Page 13 of 21

6. Duration of Proposed Alternative

This request is applicable to the Oconee Unit 2 inservice inspection program for the fifth and sixth inspection intervals.

7. Precedents

  • Surry Power Station Units 1 and 2 - Relief Implementing Extended Reactor Vessel Inspection Interval (TAC Nos. ME8573 and ME8574), dated April 30, 2013, Agencywide Document Access and Management System (ADAMS) Accession Number ML13106A140.
  • 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), dated March 20, 2014, ADAMS Accession Number ML14030A570.
  • 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), dated August 1, 2014, ADAMS Accession Number ML14188B920.
  • Byron Station, Unit No. 1 - Relief from Requirements of the ASME Code to Extend the Reactor Vessel Inservice Inspection Interval (TAC No. MF3596), dated December 10, 2014, ADAMS Accession Number ML14303A506.
  • 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), dated December 10, 2014, ADAMS Accession Number ML14321A864.
  • 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), dated February 10, 2015, ADAMS Accession Number ML15035A148.
  • 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), dated March 15, 2017, ADAMS Accession Number ML17054C255.
  • 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), dated July 24, 2018, ADAMS Accession Number ML18177A425.
  • 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), dated October 26, 2018, ADAMS Accession Number ML18284A310.
  • 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), dated April 22, 2019, ADAMS Accession Number ML19100A004.

10 CFR 50.55a Relief Request RA-20-0328 Enclosure 1 Page 14 of 21

  • 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), dated March 4, 2020, ADAMS Accession Number ML20036F261.
8. References
1. ASME Boiler and Pressure Vessel Code,Section XI, 2007 Edition through 2008 Addenda, American Society of Mechanical Engineers, New York.
2. PWROG Letter 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, July 12, 2010, (ADAMS Accession Number ML11153A033).
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, November 2002, (ADAMS Accession Number ML023240437).
4. Westinghouse Report, WCAP-16168-NP-A, Revision 3, Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval, October 2011, (ADAMS Accession Number ML11306A084).
5. NUREG-1874, Recommended Screening Limits for Pressurized Thermal Shock (PTS), U.S.

Nuclear Regulatory Commission, March 2010, (ADAMS Accession Number ML15222A848).

6. NRC Letter Report, Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants, U.S. Nuclear Regulatory Commission, December 14, 2004 (ADAMS Accession Number ML042880482).
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.
8. NRC Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S. Nuclear Regulatory Commission, May 1988, (ADAMS Accession Number ML003740284).
9. Westinghouse Report, WCAP-17571-NP, Revision 0, Oconee Units 1, 2, and 3 Reactor Pressure Vessel Integrity Program Plans, June 2012.
10. Framatome Document, 86-9315280-000, Equivalent Fatigue Crack Growth for Oconee 1, 2 and 3 Units Beltline Shell Location, September 16, 2020.

10 CFR 50.55a Relief Request RA-20-0328 Enclosure 1 Page 15 of 21 Proposed Alternative for Oconee Unit 3 In Accordance with 10 CFR 50.55a(z)(1)

-Alternative Provides Acceptable Level of Quality and Safety-

1. ASME Code Component(s) Affected The affected component is the Oconee Unit 3 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 B1.11 Circumferential Shell Welds B-A B1.21 Circumferential Head Welds B-A B1.30 Shell-to-Flange Weld 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 through 2008 Addenda (Reference 1) is applicable for Oconee Unit 3 Fifth Inservice Inspection Interval (ISI), which started on July 15, 2014 and is scheduled to end on July 15, 2024. Volumetric examinations shall be performed for the Sixth ISI interval in accordance with requirements of 10 CFR 50.55a and the Edition and Addenda of the ASME Code,Section XI, applicable at the time of the examination.

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 ISI interval.

10 CFR 50.55a Relief Request RA-20-0328 Enclosure 1 Page 16 of 21

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 ISI 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 radiological exposure and examination costs.

5. Proposed Alternative and Basis for Use

Duke Energy proposes not to perform the ASME Code required volumetric examination of the Oconee Unit 3 reactor vessel full penetration pressure-retaining Examination Category B-A and B-D welds for the fifth inservice inspection, currently scheduled for Spring 2024. Duke Energy will perform the fifth volumetric examination of the Oconee Unit 3 reactor vessel full penetration pressure-retaining Examination Category B-A and B-D welds during the Sixth ISI interval in 2034. The proposed inspection date is consistent with the latest revised implementation plan, OG-10-238 (Reference 2).

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 1.174 (Reference 3).

The methodology used to conduct this analysis is based on that defined in the study WCAP-16168-NP-A, Revision 3, Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval (Reference 4).

This study focuses on risk assessments of materials within the beltline region of the RV wall. The results of the calculations for Oconee Unit 3 were compared to those obtained from the Babcock & Wilcox pilot plant evaluated in WCAP-16168-NP-A, Revision 3. Appendix A of the WCAP identifies the parameters to be compared. Demonstrating that the parameters for Oconee Unit 3 are bounded by the results of the Babcock & Wilcox pilot plant qualifies Oconee Unit 3 for an ISI interval extension.

10 CFR 50.55a Relief Request RA-20-0328 Enclosure 1 Page 17 of 21 Table 1 below lists the critical parameters investigated in the WCAP and compares the results of the Babcock & Wilcox pilot plant to those of Oconee Unit 3. Tables 2 and 3 provide additional information that was requested by the NRC and included in Appendix A of Reference 4.

Table 1: Critical Parameters for the Application of Bounding Analysis for Oconee Unit 3 Additional Pilot Plant Parameter Plant-Specific Basis Evaluation Basis Required?

Dominant Pressurized Thermal Shock NRC PTS Risk Study PTS Generalization (PTS) Transients in the NRC PTS Risk No (Reference 5) Study (Reference 6)

Study are Applicable 2.43E-12 Events per Through-Wall Cracking Frequency 4.42E-07 Events per year (Calculated per No (TWCF) year (Reference 4)

Reference 4)

Yes 12 heatup/cooldown Bounded by 12 (as required by Frequency and Severity of Design Basis cycles per year heatup/cooldown Reference 4 and Transients (Reference 4) cycles per year summarized in Reference 10)

Single Layer Cladding Layers (Single/Multiple) Single Layer No (Reference 4)

Table 2 below provides a summary of the latest reactor vessel inspection for Oconee Unit 3 and an evaluation of the recorded indications. This information confirms that satisfactory examinations have been performed on the Oconee Unit 3 reactor vessel.

10 CFR 50.55a Relief Request RA-20-0328 Enclosure 1 Page 18 of 21 Table 2: Additional Information Pertaining to Reactor Vessel Inspection for Oconee Unit 3 The latest ISI for Oconee Unit 3 was conducted in accordance with the ASME Code,Section XI, 1998 Edition, with 2000 Addenda. Examinations of Category B-A and B-D welds were Inspection performed to ASME Section XI, Appendix VIII, 1998 Edition with 2000 Addenda as methodology:

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 Four inservice inspections have been performed.

inspections:

There were 47 indications identified in the beltline and extended beltline region of the RV during the last ISI. The indications are located in the Lower Nozzle Belt to Upper Shell, Upper Shell to Lower Shell and Lower Shell to Dutchman circumferential welds (Items 5, 6 and 7, respectively in Table 3). All indications are subsurface and are acceptable per Table IWB-3510-1 of Section XI of the ASME Code. Five of these indications are within the inner 1/10th or 1 inch of the reactor vessel thickness and required further evaluation. These indications are acceptable per the requirements of the Alternate PTS Rule, 10 CFR 50.61a (Reference 7).

A disposition of the five flaws against the limits of the Alternate PTS Rule is shown in the tables below. The five flaws evaluated are located within the forging material of the reactor vessel.

Number of indications Scaled Maximum number of flaws per 9691 square-inches Through-Wall Extent, Number of found: of inside surface area in the inspection volume that are Oconee Unit 3 TWE (in) greater than or equal to TWEMIN and less than TWEMAX.

Flaws Evaluated This flaw density does not include underclad cracks in I

(Axial/Circ.)

I TWEMIN TWEMAX I forgings. I I 0 0.075 No Limit 0 0.075 0.375 78 5 (2/3) 0.125 0.375 30 4 (2/2) 0.175 0.375 8 0 0.225 0.375 2 0 0.275 0.375 0 0 0.325 0.375 0 0 0.375 Infinite 0 0 Proposed inspection The fifth inservice inspection is currently scheduled for Spring 2024. This inspection is schedule for proposed to be performed during the 2034 refueling outage. The proposed inspection date is balance of plant consistent with the latest revised implementation plan, OG-10-238 (Reference 2).

life:

10 CFR 50.55a Relief Request RA-20-0328 Enclosure 1 Page 19 of 21 Table 3 summarizes the inputs and outputs for the calculation of through-wall cracking frequency (TWCF).

Table 3: Details of TWCF Calculation for Oconee Unit 3 at 60 Years of Operation or 54 Effective Full Power Years (EFPY)

Inputs(1)

I Twall [inches]: 8.628 Region and Component Material Material Copper Nickel R.G. 1.99 Chemistry Fluence [n/cm2, No. RTNDT(u) [ºF]

Description Identification Heat No. [weight %] [weight %] Position Factor [ºF] E > 1.0 MeV]

1 Lower Nozzle Belt Forging 4680 4680 0.13 0.91 1.1 96 3 1.38E+19 2 Upper Shell Forging AWS 192 522314 0.01 0.73 2.1 36 40 1.54E+19 3 Lower Shell Forging ANK 191 522194 0.02 0.76 1.1 20 40 1.54E+19 4 Dutchman Forging 417543-1 417543-1 0.13 0.85 1.1 96 3 1.33E+17 Lower Nozzle Belt to Upper Shell 821T44, Flux 5 WF-200 0.24 0.63 1.1 178 -5 1.38E+19 Circumferential Weld (100%) Lot 8773 Upper Shell to Lower Shell 72442, Flux 6a WF-67 0.26 0.60 1.1 180 -5(2) 1.49E+19 Circumferential Weld (75%) Lot 8669 Upper Shell to Lower Shell 72105, Flux 6b WF-70 0.32 0.58 1.1 199.3(2) -26 1.49E+19 Circumferential Weld (25%) Lot 8669 Lower Shell to Dutchman 8T1554, Flux 7 WK-169-1 0.16 0.57 1.1 143.9 -5 1.33E+17 Circumferential Weld (100%) Lot 8754 Outputs Methodology Used to Calculate T30: Regulatory Guide 1.99, Revision 2 (Reference 8)

Controlling Fluence [n/cm2, FF (Fluence Material Region No. XX RTMAX-XX [°R] T30 [ºF] TWCF95-XX E >1.0 MeV] Factor)

(From Above)

Limiting Circumferential Weld - CW 6a, 6b 2.1941 675.98 1.4900E+19 1.1104 221.31 9.695E-13 Limiting Forging - FO 1 2.5000 567.26 1.3800E+19 1.0895 104.59 1.223E-13 TWCF95-TOTAL = (CWTWCF95-CW + FOTWCF95-FO): 2.43E-12 Note 1: Material properties are based on WCAP-17571-NP (Reference 9). Fluence projections and material properties not included in Reference 9 were provided by Framatome.

Note 2: The Upper Shell to Lower Shell weld seam was fabricated using two weld heats. The more limiting initial RTNDT value and Chemistry Factor (CF) values between the two weld head were used for these materials in the subsequent TWCF calculations.

10 CFR 50.55a Relief Request RA-20-0328 Enclosure 1 Page 20 of 21

6. Duration of Proposed Alternative

This request is applicable to the Oconee Unit 3 inservice inspection program for the fifth and sixth inspection intervals.

7. Precedents

  • Surry Power Station Units 1 and 2 - Relief Implementing Extended Reactor Vessel Inspection Interval (TAC Nos. ME8573 and ME8574), dated April 30, 2013, Agencywide Document Access and Management System (ADAMS) Accession Number ML13106A140.
  • 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), dated March 20, 2014, ADAMS Accession Number ML14030A570.
  • 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), dated August 1, 2014, ADAMS Accession Number ML14188B920.
  • Byron Station, Unit No. 1 - Relief from Requirements of the ASME Code to Extend the Reactor Vessel Inservice Inspection Interval (TAC No. MF3596), dated December 10, 2014, ADAMS Accession Number ML14303A506.
  • 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), dated December 10, 2014, ADAMS Accession Number ML14321A864.
  • 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), dated February 10, 2015, ADAMS Accession Number ML15035A148.
  • 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), dated March 15, 2017, ADAMS Accession Number ML17054C255.
  • 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), dated July 24, 2018, ADAMS Accession Number ML18177A425.
  • 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), dated October 26, 2018, ADAMS Accession Number ML18284A310.
  • 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), dated April 22, 2019, ADAMS Accession Number ML19100A004.

10 CFR 50.55a Relief Request RA-20-0328 Enclosure 1 Page 21 of 21

  • 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), dated March 4, 2020, ADAMS Accession Number ML20036F261.
8. References
1. ASME Boiler and Pressure Vessel Code,Section XI, 2007 Edition through 2008 Addenda, American Society of Mechanical Engineers, New York.
2. PWROG Letter 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, July 12, 2010, (ADAMS Accession Number ML11153A033).
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, November 2002, (ADAMS Accession Number ML023240437).
4. Westinghouse Report, WCAP-16168-NP-A, Revision 3, Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval, October 2011, (ADAMS Accession Number ML11306A084).
5. NUREG-1874, Recommended Screening Limits for Pressurized Thermal Shock (PTS), U.S.

Nuclear Regulatory Commission, March 2010, (ADAMS Accession Number ML15222A848).

6. NRC Letter Report, Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants, U.S. Nuclear Regulatory Commission, December 14, 2004 (ADAMS Accession Number ML042880482).
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.
8. NRC Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S. Nuclear Regulatory Commission, May 1988, (ADAMS Accession Number ML003740284).
9. Westinghouse Report, WCAP-17571-NP, Revision 0, Oconee Units 1, 2, and 3 Reactor Pressure Vessel Integrity Program Plans, June 2012.
10. Framatome Document, 86-9315280-000, Equivalent Fatigue Crack Growth for Oconee 1, 2 and 3 Units Beltline Shell Location, September 16, 2020.

10 CFR 50.55a Relief Request RA-20-0328 Attachment 1 Page 1 of 11 : Framatome Document 86-9315280-000 Equivalent Fatigue Crack Growth for Oconee Units 1, 2 and 3 Beltline Shell Location

10 CFR 50.55a Relief Request RA-20-0328 Attachment 1 Page 2 of 11 Controlled Document 0402-01-F01 (Rev. 021 , 03/12/2018) framatome CALCULATION

SUMMARY

SHEET (CSS)

Document No. 86 - 9315280 - 000 Safety Related: ~ Yes

  • No Title Equivalent Fatigue Crack Growth for Oconee 1, 2 and 3 Units Beltline Shell Location PURPOSE AND

SUMMARY

OF RESULTS:

Purpose:

Per the NRC's Safety Evaluation Report (SER), the results of WCAP-16168-NP-A, Revision 3 (Reference [11) can only be used on B&W-designed Oconee units after the assumption that an equivalent of 12 heat-up I 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 ofWCAP-16168-NP-A, Reference [1], for Oconee Units 1, 2, and 3, that 12 equ ivalent heat-up and cool-down cycles bound the fatigue crack growth from all the other test, normal / upset and emergency I faulted service level transients is valid.

Maximum RT MAX-FO values for each unit:

ONS-1 Forging= 81 .?°F (541 .3°R), ONS-2 Forging= 86. ?°F (546.4°R), ONS-3 Forging= 145.0°F (604.?°R).

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

ASSUMPTIONS THAT SHALL BE THE FOLLOW ING COMPUTER CODES HAVE BEEN USED IN THIS DOCUMENT: VERI FIED PRIOR TO USE N/A CODENERSION/REV CODENERSION/REV

  • Yes

~ No Page 1 of 10

10 CFR 50.55a Relief Request RA-20-0328 Attachment 1 Page 3 of 11 Controlled Document framatome 0402-01-F0 1 (Rev. 021 , 03/12/2018)

Document No. 86-9315280-000 Equivalent Fatigue Crack Growth for Oconee 1, 2 and 3 Units Beltline Shell Location Review Method: ~ Design Review (Detailed Check)

D Alternate Calculation Does this document establish design or technical requirements? D YES ~ NO Does this document contain Customer Required Fonnat? D YES ~ NO Signature Block P/R/A/M Name and Title Pages/Sections Signature and Date (printed or typed) Prepared/Reviewed/Approved LP/LR Maiti.n Kolai*, p MKOLAR All Pages / All Sections Principal Engineer 9/16/2020 Luziana Matte, LRMAITE 9/16/2020 R All Pages / All Sections Technical Consultant David R. Coffli.n, DRCOFFLIN A All Pages / All Sections Supervisory Engineer 9/16/2020 l'iotes : P/RJA 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 pennission, calculations may be revised without using the latest CSS fo1m. This deviation is pemritted when expediency and/or cost are a factor. Approver shall add a comment in the tight-most column that acknowledges and justifies this deviation.

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

Name Title Sig natu re Date Comments (printed or typed) (printed or typed)

NIA Page2

10 CFR 50.55a Relief Request RA-20-0328 Attachment 1 Page 4 of 11 Controlled Document framatome 0402-01-F01 (Rev. 021 , 03/12/2018)

Document No. 86-9315280-000 Equivalent Fatigue Crack Growth for Oconee 1, 2 and 3 Units Beltline Shell Location Record of Revision Revision Pages/Sections/Paragraphs No. Changed Brief Description / Change Authorization 000 All Pages / All Sections fuitial Issue of the Document Page 3

10 CFR 50.55a Relief Request RA-20-0328 Attachment 1 Page 5 of 11 Controlled Document framatome Document No. 86-9315280-000 Equivalent Fatigue Crack Growth for Oconee 1, 2 and 3 Units Beltline Shell Location 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 RT MAX Temperatures ............... ... .... ... .... .................. ................................... .... ................................... 9

6.0 REFERENCES

...... ... ... ......... ..... ..... ............... ....... ..... ..... ..... .......... .. ..... ..... .......... ..... ....... ..... ... . 10 Page4

10 CFR 50.55a Relief Request RA-20-0328 Attachment 1 Page 6 of 11 Controlled Document framatome Document No. 86-9315280-000 Equivalent Fatigue Crack Growth for Oconee 1, 2 and 3 Units Beltline Shell Location List of Tables Page Table 3-1 : List of Transient Events Considered for the ONS Units ..... ........... .......... ..... ....... .......... ......... 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 Page 5

10 CFR 50.55a Relief Request RA-20-0328 Attachment 1 Page 7 of 11 Controlled Document framatome Document No. 86-9315280-000 Equivalent Fatigue Crack Growth for Oconee 1, 2 and 3 Units Beltline Shell Location 1.0 PURPOSE This docmnent is patt of the Risk-Infonned Extension of Reactor Vessel In-Se1vice Inspection Inte1val PWR Owners Group project PA-MSC-0943, B&W Site Specific Fatigue Crack Growth Evaluation.

Westinghouse Electric Company (WEC) perfonned a Risk-Infonued Extension of the Reactor Vessel In-Se1vice Inspection Inte1val analysis documented in WCAP-16168-NP-A, Revision 3 and US NRC approved this work in 2011 (Reference [l]).

Per the NRC's Safety Evaluation Report (SER), the results of WCAP-16168-NP-A, Revision 3 (Reference [l])

can only be used on B&W designed Oconee units if the following requirements outlined in Appendix B (items 3 and 4) of Reference [l ] are addressed:

Item 3:

  • Licensees must ve1ify 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 U11derclad 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 inte1val for ASME Code,Section XI, Category B-A and B-D RPV welds.

This document is a summa1y of an analysis validating item 3 and item 4 for a 20 year service inte1val, where 20 years of operation cotTesponds to IO-years for the regular in-se1vice inspection (ISi) period plus an additional 10 years for ISi extension inte1val.

This cutTent summa1y document is a 11011-prop1ieta1y docmnent in suppo1t of a relief request being prepared by Westinghouse.

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

2.2 Justified Assumptions

1. For the pmpose of this document, based on the che1nical composition, the cladding weld material is taken as 18Cr-8Ni (Type 304 stainless steel) for all three Oconee tmit reactor vessels under consideration.
2. Due to the similarities in the base metal material chemical compositions for all Oconee plant units, the reactor vessel shell material is assumed as low-alloy steel SA-508 Class 2 for the entit-e region of interest and for all three Oconee units tmder consideration.

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10 CFR 50.55a Relief Request RA-20-0328 Attachment 1 Page 8 of 11 Controlled Document framatome Document No. 86-9315280-000 Equivalent Fatigue Crack Growth for Oconee 1, 2 and 3 Units Beltline Shell Location 3.0 INPUTS 3.1 Temperature/ Pressure Trans ients The inside surface of the reactor vessel is subjected to transient loads in the fo1m of prima1y coolant cold leg temperatures and pressures as defined by the reactor coolant system functional specification. The number of applicable transient cycles conesponds to 20 years of operation. In order to fotm complete stress cycles, these individual transients are combined into transient groups. These combined transients are used to calculate the cyclic va1iations 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 the ONS Units Transient Transient Name / Description Category Number111 lA Heat -up from 0% to 8% Power normal 1B Cool-down from 8% to 0% Power normal 2A Power Change from 0% to 15% normal 2B Power Change from 15% to 0% normal 3 Power Loading from 8% to 100% normal 4 Power Un loading from 100% to 8% normal 5 10% Step Load Increase normal 6 10% Step Load Decrease normal 7 Step Load Reduction from 100% to 8% Power upset 8A Reactor Trip - Type A (Loss of RC Flow) upset 8B Reactor Trip - Type B (Control System Malfunction) upset 8C Reactor Trip - Type C (Loss of MFW Flow) upset 9 Rap id Depressurization upset 10 Change of Flow upset 111') Rod Withdrawa l Accident upset 12 Hydro-test test 14 Control Rod Drop upset 1512) Loss of Station Power upset 16 Steam Line Failure fau lted 11 17A ' Loss of Feedwater to One Steam Generator upset 1

17B ' 1 Stuck Open Turbine Bypass Va lve emergency 1 Loss of Coolant fau lted 21

22A High Pressure Inj ection Test normal 22B Core Flood Check Valve Test normal Note (1): Transients that do not contribute to the growth of the postulated flaw are not considered .

Note (2): part of Reactor Trip group The calculation of fina l crack size follows methodology of Section A-5200, Reference (2)

Crack growth rates are based on Section A-4300 of Reference [2]

MFW = Main Feed Water RC= Reactor Coolant Page 7

10 CFR 50.55a Relief Request RA-20-0328 Attachment 1 Page 9 of 11 Controlled Document framatome Document No. 86-9315280-000 Equivalent Fatigue Crack Growth for Oconee 1, 2 and 3 Units Beltline Shell Location 4.0 RESULTS 4.1 Transient Fatigue Crack Growth Summary The initial (5% 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 of 0.4313 inch bom1ds all as-found flaws from the latest in-se1vice inspections (ISi) for Oconee 1, 2, and 3 Units.

The analysis considers a flaw length over depth aspect ratio Il a = 6 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 0.4313 20 0.4378 4.2 Equivalent Crack Growth The result of the calculation shown in this Section is an acceptance ciite1ion 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 a cycles

[year] [inch]

0 0 0.4313 20 240 0.4452 4.3 Transient Crack Growth Contribution Ratio This section presents results from the detailed transient fatigue crack growth calculation smmnarized in Table 4-1 for 20 years of operation. The lines in Table 4-3 are so1ted by their relative cont1ibution 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 cool-down transient, which together cont1ibute to nearly 90% of the total crack growth.

Remaining transients are considered negligible with a total contribution of less than 10% of the total crack growth.

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10 CFR 50.55a Relief Request RA-20-0328 Attachment 1 Page 10 of 11 Controlled Document framatome Document No. 86-9315280-000 Equivalent Fatigue Crack Growth for Oconee 1, 2 and 3 Units Beltline Shell Location Table 4-3: Crack Growth Relative Contribution participation transient name

% ratio power change power loading ~ 90%

cool -down rod withdrawal flow change power unloading

~10%

trip step load increase/ decrease feedwater loss

5.0 CONCLUSION

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

Fmthennore, power change, power loading and cool-down transients were identified as major contributors to the fatigue crack grov.rth.

This fulfills the intent of Item 3, Appendix B ofWCAP-16168-NP-A, Reference [1] for Oconee Units 1, 2, and 3.

5.2 RT MAX Temperatures The Maximum RTMAX-ro values for each unit are as follows:

ONS-1 Forging= 81.7°F (541.3°R)

ONS-2 Forging= 86.7°F (546.4°R)

ONS-3 Forging= 145.0°F (604.7°R).

This fulfills the intent of Item 4, Appendix B ofWCAP-16168-NP-A, Reference [1] for Oconee Uni.ts 1, 2, and 3.

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10 CFR 50.55a Relief Request RA-20-0328 Attachment 1 Page 11 of 11 Controlled Document framatome Document No. 86-9315280-000 Equivalent Fatigue Crack Growth for Oconee 1, 2 and 3 Units Beltline Shell Location

6.0 REFERENCES

l. Risk-Infonned Extension of the Reactor Vessel In-Se1v ice Inspection Inte1val, WCAP-16168-NP-A, Revision 3, October 2011 , Westinghouse Electric Company LLC
2. ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inse1vice Inspection ofNuclear Power Plant Components, 2013 Edition Page 10
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