RA-22-0256, Proposed Alternative for Steam Generator Welds in Accordance with 10 CFR 50.55a(z)(1)

Proposed Alternative for Steam Generator Welds in Accordance with 10 CFR 50.55a(z)(1)

ML23023A093
Person / Time
Site:	Oconee, Mcguire, Catawba, Harris, Robinson, McGuire
Issue date:	01/23/2023
From:	Ellis K Duke Energy Carolinas, Duke Energy Progress
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
RA-22-0256
Download: ML23023A093 (1)

Text

{{#Wiki_filter:10 CFR 50.55a Serial: RA-22-0256 January 23, 2023 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Catawba Nuclear Station, Unit Nos. 1 and 2 Docket Nos. 50-413, 50-414 / Renewed License Nos. NPF-35 and NPF-52 Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400 / Renewed License No. NPF-63 McGuire Nuclear Station, Unit Nos. 1 and 2 Docket Nos. 50-369, 50-370 / Renewed License Nos. NPF-9 and NPF-17 Oconee Nuclear Station, Unit Nos. 1, 2, and 3 Docket Nos. 50-269, 50-270, and 50-287 / Renewed License Nos. DPR-38, DPR-47, and DPR-55 H. B. Robinson Steam Electric Plant, Unit No. 2 Docket No. 50-261 / Renewed License No. DPR-23

SUBJECT:

Proposed Alternative for Steam Generator Welds in Accordance with 10 CFR 50.55a(z)(1) Ladies and Gentlemen: In accordance with 10 CFR 50.55a(z)(1), Duke Energy Carolinas, LLC and Duke Energy

Progress, LLC (collectively referred to as Duke Energy) requests U.S. Nuclear Regulatory Commission (NRC) approval of a proposed alternative to certain requirements of the American Society of Mechanical Engineers (ASME) Code, Section XI for certain Steam Generator welds at Catawba Nuclear Station Units 1 and 2 (CNS), McGuire Nuclear Station Units 1 and 2 (MNS), Oconee Nuclear Station Units 1, 2, and 3 (ONS), Shearon Harris Nuclear Power Plant, Unit 1 (HNP), and H. B. Robinson Steam Electric Plant, Unit 2 (RNP). contains details for the proposed alternative. Attachments 1 through 6 contain plant specific analyses to support the alternative for each site. Attachment 7 contains results achieved from an industry survey of past Steam Generator weld inspections that provide supporting information.

Duke Energy requests NRC approval of the proposed alternative within one year of acceptance for review. Should you have any question concerning this letter and its enclosure, please contact Ryan Treadway, Director - Nuclear Fleet Licensing at (980) 373-5873. Kevin M. Ellis General Manager Nuclear Regulatory Affairs, Policy & Emergency Preparedness Duke Energy 13225 Hagers Ferry Rd., MG011E Huntersville, NC 28078 843-951-1329 Kevin.Ellis@duke-energy.com ( ~ DUKE ENERGY@

U.S. Nuclear Regulatory Commission RA-22-0256 Page 2 No new regulatory commitments have been made in this submittal. Kevin llis General Manager, Nuclear Regulatory Affairs, Policy & Emergency Preparedness

Enclosure:

1. Request for Alternative Inspection Interval Extension for Steam Generator Pressure Retaining Welds and Full-Penetration Welded Nozzles Attachments:

1. Plant-Specific Applicability CNS1 2.

Plant-Specific Applicability CNS2 3. Plant-Specific Applicability MNS1/2

4. Plant-Specific Applicability ONS1/2/3

5. Plant-Specific Applicability HNP

6. Plant-Specific Applicability RNP 7.

Results of lndusry Survey

U.S. Nuclear Regulatory Commission RA-22-0256 Page 3 cc: L. Dudes, USNRC, Region II Regional Administrator N. Jordan, USNRC NRR Project Manager for Duke Fleet M. Mahoney, USNRC NRR Project Manager for HNP J. Klos, USNRC NRR Project Manager for MNS S. Williams, USNRC NRR Project Manager for ONS & CNS L. Haeg, USNRC NRR Project Manager for RNP D. Rivard, USNRC Senior Resident Inspector for CNS J. Zeiler, USNRC Senior Resident Inspector for HNP A. Hutto, USNRC Senior Resident Inspector for MNS J. Nadel, USNRC Senior Resident Inspector for ONS M. Fannon, USNRC Senior Resident Inspector for RNP

ENCLOSURE 1 10 CFR 50.55a(z)(1) Request for Alternative Inspection Interval Extension for Steam Generator Pressure-Retaining Welds and Full-Penetration Welded Nozzles RA-22-0256 Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 1 of 29 1.0 ASME CODE COMPONENTS AFFECTED: Code Class: Class 1 and Class 2

== Description:== Steam generator (SG) pressure-retaining welds and full penetration welded nozzles (nozzle-to-shell welds and inside radius sections) Examination Category: Class 1, Category B-B, pressure-retaining welds in vessels other than reactor vessels Class 2, Category C-A, pressure-retaining welds in pressure vessels Class 2, Category C-B (Pressure Retaining Nozzle Welds in Pressure Vessels, Section XI, Division 1) Item Numbers: B2.40 - Steam generators (primary side), tubesheet-to-head weld C1.10 - Shell circumferential welds C1.20 - Head circumferential welds C1.30 - Tubesheet-to-shell weld C2.21 - Nozzle-to-shell (nozzle-to-head or nozzle-to-nozzle) welds C2.22 - Nozzle inside radius sections Catawba Unit 1 (CNS1) ASME Category ASME Item No. Component ID Component Description B-B B2.40 1SGA-W22 Tubesheet-to-Head B-B B2.40 1SGB-W22 Tubesheet-to-Head B-B B2.40 1SGC-W22 Tubesheet-to-Head B-B B2.40 1SGD-W22 Tubesheet-to-Head C-A C1.20 1SGA-W144 Shell-to-Head C-A C1.20 1SGB-W144 Shell-to-Head C-A C1.20 1SGC-W144 Shell-to-Head C-A C1.20 1SGD-W144 Shell-to-Head C-A C1.30 1SGA-W65 Tubesheet-to-Shell C-A C1.30 1SGB-W65 Tubesheet-to-Shell C-A C1.30 1SGC-W65 Tubesheet-to-Shell C-A C1.30 1SGD-W65 Tubesheet-to-Shell C-B C2.21 1SGA-W258 Feedwater Nozzle-to-Shell C-B C2.21 1SGB-W258 Feedwater Nozzle-to-Shell Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 2 of 29 Catawba Unit 1 (CNS1) ASME Category ASME Item No. Component ID Component Description C-B C2.21 1SGC-W258 Feedwater Nozzle-to-Shell C-B C2.21 1SGD-W258 Feedwater Nozzle-to-Shell C-B C2.22 1SGA-W258 Feedwater Nozzle Inside Radius Section C-B C2.22 1SGB-W258 Feedwater Nozzle Inside Radius Section C-B C2.22 1SGC-W258 Feedwater Nozzle Inside Radius Section C-B C2.22 1SGD-W258 Feedwater Nozzle Inside Radius Section Catawba Unit 2 (CNS2) ASME Category ASME Item No. Component ID Component Description B-B B2.40 2SGA-01-02 Tubesheet-to-Head B-B B2.40 2SGB-01-02 Tubesheet-to-Head B-B B2.40 2SGC-01-02 Tubesheet-to-Head B-B B2.40 2SGD-01-02 Tubesheet-to-Head C-A C1.10 2SGA-03-04A Stub Barrel-to-Lower Shell C-A C1.10 2SGB-03-04A Stub Barrel-to-Lower Shell C-A C1.10 2SGC-03-04A Stub Barrel-to-Lower Shell C-A C1.10 2SGD-03-04A Stub Barrel-to-Lower Shell C-A C1.10 2SGA-04B-05 Lower Shell-to-Transition Cone C-A C1.10 2SGB-04B-05 Lower Shell-to-Transition Cone C-A C1.10 2SGC-04B-05 Lower Shell-to-Transition Cone C-A C1.10 2SGD-04B-05 Lower Shell-to-Transition Cone C-A C1.10 2SGA-05-06A Transition Cone-to-Upper Shell C-A C1.10 2SGB-05-06A Transition Cone-to-Upper Shell C-A C1.10 2SGC-05-06A Transition Cone-to-Upper Shell C-A C1.10 2SGD-05-06A Transition Cone-to-Upper Shell C-A C1.20 2SGA-06B-07 Upper Shell-to-Upper Head C-A C1.20 2SGB-06B-07 Upper Shell-to-Upper Head C-A C1.20 2SGC-06B-07 Upper Shell-to-Upper Head C-A C1.20 2SGD-06B-07 Upper Shell-to-Upper Head C-A C1.30 2SGA-02-03 Tubesheet-to-Stub Barrel C-A C1.30 2SGB-02-03 Tubesheet-to-Stub Barrel C-A C1.30 2SGC-02-03 Tubesheet-to-Stub Barrel Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 3 of 29 Catawba Unit 2 (CNS2) ASME Category ASME Item No. Component ID Component Description C-A C1.30 2SGD-02-03 Tubesheet-to-Stub Barrel C-B C2.21 2SGA-SB-11 Feedwater Nozzle-to-Stub Barrel C-B C2.21 2SGB-SB-11 Feedwater Nozzle-to-Stub Barrel C-B C2.21 2SGC-SB-11 Feedwater Nozzle-to-Stub Barrel C-B C2.21 2SGD-SB-11 Feedwater Nozzle-to-Stub Barrel C-B C2.21 2SGA-UH-15 Main Steam Nozzle-to-Shell C-B C2.21 2SGB-UH-15 Main Steam Nozzle-to-Shell C-B C2.21 2SGC-UH-15 Main Steam Nozzle-to-Shell C-B C2.21 2SGD-UH-15 Main Steam Nozzle-to-Shell C-B C2.22 2SGA-SB-11 Feedwater Nozzle-to-Stub Barrel Inside Radius Section C-B C2.22 2SGB-SB-11 Feedwater Nozzle-to-Stub Barrel Inside Radius Section C-B C2.22 2SGC-SB-11 Feedwater Nozzle-to-Stub Barrel Inside Radius Section C-B C2.22 2SGD-SB-11 Feedwater Nozzle-to-Stub Barrel Inside Radius Section McGuire Unit 1 (MNS1) ASME Category ASME Item No. Component ID Component Description B-B B2.40 1SGA-W22 Tubesheet-to-Head B-B B2.40 1SGB-W22 Tubesheet-to-Head B-B B2.40 1SGC-W22 Tubesheet-to-Head B-B B2.40 1SGD-W22 Tubesheet-to-Head C-A C1.20 1SGA-W144 Shell-to-Head C-A C1.20 1SGB-W144 Shell-to-Head C-A C1.20 1SGC-W144 Shell-to-Head C-A C1.20 1SGD-W144 Shell-to-Head C-A C1.30 1SGA-W65 Tubesheet-to-Shell C-A C1.30 1SGB-W65 Tubesheet-to-Shell C-A C1.30 1SGC-W65 Tubesheet-to-Shell C-A C1.30 1SGD-W65 Tubesheet-to-Shell C-B C2.21 1SGA-W258 Feedwater Nozzle-to-Shell C-B C2.21 1SGB-W258 Feedwater Nozzle-to-Shell C-B C2.21 1SGC-W258 Feedwater Nozzle-to-Shell C-B C2.21 1SGD-W258 Feedwater Nozzle-to-Shell C-B C2.22 1SGA-W258 Feedwater Nozzle-to-Shell Inside Radius Section C-B C2.22 1SGB-W258 Feedwater Nozzle-to-Shell Inside Radius Section Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 4 of 29 McGuire Unit 1 (MNS1) ASME Category ASME Item No. Component ID Component Description C-B C2.22 1SGC-W258 Feedwater Nozzle-to-Shell Inside Radius Section C-B C2.22 1SGD-W258 Feedwater Nozzle-to-Shell Inside Radius Section McGuire Unit 2 (MNS2) ASME Category ASME Item No. Component ID Component Description B-B B2.40 2SGA-W22 Tubesheet-to-Head B-B B2.40 2SGB-W22 Tubesheet-to-Head B-B B2.40 2SGC-W22 Tubesheet-to-Head B-B B2.40 2SGD-W22 Tubesheet-to-Head C-A C1.20 2SGA-W144 Shell-to-Head C-A C1.20 2SGB-W144 Shell-to-Head C-A C1.20 2SGC-W144 Shell-to-Head C-A C1.20 2SGD-W144 Shell-to-Head C-A C1.30 2SGA-W65 Tubesheet-to-Shell C-A C1.30 2SGB-W65 Tubesheet-to-Shell C-A C1.30 2SGC-W65 Tubesheet-to-Shell C-A C1.30 2SGD-W65 Tubesheet-to-Shell C-B C2.21 2SGA-W258 Feedwater Nozzle-to-Shell C-B C2.21 2SGB-W258 Feedwater Nozzle-to-Shell C-B C2.21 2SGC-W258 Feedwater Nozzle-to-Shell C-B C2.21 2SGD-W258 Feedwater Nozzle-to-Shell C-B C2.22 2SGA-W258 Feedwater Nozzle-to-Shell Inside Radius Section C-B C2.22 2SGB-W258 Feedwater Nozzle-to-Shell Inside Radius Section C-B C2.22 2SGC-W258 Feedwater Nozzle-to-Shell Inside Radius Section C-B C2.22 2SGD-W258 Feedwater Nozzle-to-Shell Inside Radius Section Oconee Unit 1 (ONS1) ASME Category ASME Item No. Component ID Component Description B-B B2.40 1-SGA-W22 Tubesheet-to-Head B-B B2.40 1-SGB-W22 Tubesheet-to-Head B-B B2.40 1-SGA-W23 Tubesheet-to-Head Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 5 of 29 Oconee Unit 1 (ONS1) ASME Category ASME Item No. Component ID Component Description B-B B2.40 1-SGB-W23 Tubesheet-to-Head C-A C1.30 1-SGA-W65 Tubesheet-to-Shell C-A C1.30 1-SGB-W65 Tubesheet-to-Shell C-A C1.30 1-SGA-W69 Tubesheet-to-Shell C-A C1.30 1-SGB-W69 Tubesheet-to-Shell C-B C2.21 1-SGA-W127 Main Steam Nozzle-to-Shell C-B C2.21 1-SGB-W127 Main Steam Nozzle-to-Shell C-B C2.21 1-SGA-W128 Main Steam Nozzle-to-Shell C-B C2.21 1-SGB-W128 Main Steam Nozzle-to-Shell Oconee Unit 2 (ONS2) ASME Category ASME Item No. Component ID Component Description B-B B2.40 2-SGA-W22 Tubesheet-to-Head B-B B2.40 2-SGB-W22 Tubesheet-to-Head B-B B2.40 2-SGA-W23 Tubesheet-to-Head B-B B2.40 2-SGB-W23 Tubesheet-to-Head C-A C1.30 2-SGA-W65 Tubesheet-to-Shell C-A C1.30 2-SGB-W65 Tubesheet-to-Shell C-A C1.30 2-SGA-W69 Tubesheet-to-Shell C-A C1.30 2-SGB-W69 Tubesheet-to-Shell C-B C2.21 2-SGA-W127 Main Steam Nozzle-to-Shell C-B C2.21 2-SGB-W127 Main Steam Nozzle-to-Shell C-B C2.21 2-SGA-W128 Main Steam Nozzle-to-Shell C-B C2.21 2-SGB-W128 Main Steam Nozzle-to-Shell Oconee Unit 3 (ONS3) ASME Category ASME Item No. Component ID Component Description B-B B2.40 3-SGA-W22 Tubesheet-to-Head B-B B2.40 3-SGB-W22 Tubesheet-to-Head B-B B2.40 3-SGA-W23 Tubesheet-to-Head B-B B2.40 3-SGB-W23 Tubesheet-to-Head C-A C1.30 3-SGA-W65 Tubesheet-to-Shell C-A C1.30 3-SGB-W65 Tubesheet-to-Shell Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 6 of 29 Oconee Unit 3 (ONS3) ASME Category ASME Item No. Component ID Component Description C-A C1.30 3-SGA-W69 Tubesheet-to-Shell C-A C1.30 3-SGB-W69 Tubesheet-to-Shell C-B C2.21 3-SGA-W127 Main Steam Nozzle-to-Shell C-B C2.21 3-SGB-W127 Main Steam Nozzle-to-Shell C-B C2.21 3-SGA-W128 Main Steam Nozzle-to-Shell C-B C2.21 3-SGB-W128 Main Steam Nozzle-to-Shell Shearon Harris Unit 1 (HNP) ASME Category ASME Item No. Component ID Component Description B-B B2.40 II-SG-001SGA-TSTHW-06-1 Tubesheet-to-Head B-B B2.40 II-SG-001SGB-TSTHW-06-1 Tubesheet-to-Head B-B B2.40 II-SG-001SGC-TSTHW-06-1 Tubesheet-to-Head C-A C1.20 II-SG-001SGA-STHW-02-1 Shell-to-Head C-A C1.20 II-SG-001SGB-STHW-02-1 Shell-to-Head C-A C1.20 II-SG-001SGC-STHW-02-1 Shell-to-Head C-A C1.30 II-SG-001SGA-TSTSW-09-1 Tubesheet-to-Shell C-A C1.30 II-SG-001SGB-TSTSW-09-1 Tubesheet-to-Shell C-A C1.30 II-SG-001SGC-TSTSW-09-1 Tubesheet-to-Shell C-B C2.21 II-SG-001SGA-FWNTSW-05-1 Feedwater Nozzle-to-Shell C-B C2.21 II-SG-001SGB-FWNTSW-05-1 Feedwater Nozzle-to-Shell C-B C2.21 II-SG-001SGC-FWNTSW-05-1 Feedwater Nozzle-to-Shell C-B C2.22 II-SG-001SGA-FWNIR-05-1 Feedwater Nozzle-to-Shell Inside Radius Section C-B C2.22 II-SG-001SGB-FWNIR-05-1 Feedwater Nozzle-to-Shell Inside Radius Section C-B C2.22 II-SG-001SGC-FWNIR-05-1 Feedwater Nozzle-to-Shell Inside Radius Section Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 7 of 29 H.B. Robinson Unit 2 (RNP) ASME Category ASME Item No. Component ID Component Description B-B B2.40 105/01 SG A Tubesheet-to-Head B-B B2.40 105A/01 SG B Tubesheet-to-Head B-B B2.40 105B/01 SG C Tubesheet-to-Head C-A C1.10 205/03 SG A Stub Barrel-to-Lower Shell C-A C1.10 205A/03 SG B Stub Barrel-to-Lower Shell C-A C1.10 205B/03 SG C Stub Barrel-to-Lower Shell C-A C1.10 205/04 SG A Upper Shell-to-Lower Shell C-A C1.10 205A/04 SG B Upper Shell-to-Lower Shell C-A C1.10 205B/04 SG C Upper Shell-to-Lower Shell C-A C1.10 205/05 SG A Upper Transition-to-Shell C-A C1.10 205A/05 SG B Upper Transition-to-Shell C-A C1.10 205B/05 SG C Upper Transition-to-Shell C-A C1.20 205/06 SG A Upper Shell-to-Head C-A C1.20 205A/06 SG B Upper Shell-to-Head C-A C1.20 205B/06 SG C Upper Shell-to-Head C-A C1.30 205/02 SG A Tubesheet-to-Stub Barrel C-A C1.30 205A/02 SG B Tubesheet-to-Stub Barrel C-A C1.30 205B/02 SG C Tubesheet-to-Stub Barrel C-B C2.21 205/07 SG A Upper Head-to-Main Steam Nozzle C-B C2.21 205A/07 SG B Upper Head-to-Main Steam Nozzle C-B C2.21 205B/07 SG C Upper Head-to-Main Steam Nozzle C-B C2.21 205/08 SG A Upper Head-to-Feedwater Nozzle C-B C2.21 205A/08 SG B Upper Head-to-Feedwater Nozzle C-B C2.21 205B/08 SG C Upper Head-to-Feedwater Nozzle C-B C2.22 205/07IR SG A Upper Head-to-Steam Nozzle Inner Radius C-B C2.22 205A/07IR SG B Upper Head-to-Steam Nozzle Inner Radius C-B C2.22 205B/07IR SG C Upper Head-to-Steam Nozzle Inner Radius C-B C2.22 205/08IR SG A Upper Head-to-Feedwater Nozzle Inner Radius C-B C2.22 205A/08IR SG B Upper Head-to-Feedwater Nozzle Inner Radius C-B C2.22 205B/08IR SG C Upper Head-to-Feedwater Nozzle Inner Radius Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 8 of 29 2.0 APPLICABLE CODE EDITION AND ADDENDA: Plants Included in This Request for Alternative and Their Current ISI Intervals and Applicable ASME Code Section XI Editions/Addenda Plant/Unit(s) ISI Interval ASME Section XI Code Edition/Addenda Current Interval Start Date Current Scheduled Interval End Date1 Catawba Nuclear Station Units 1 and 2 Fourth 2007 Edition, Through 2008 Addenda 08/19/2015 12/06/2024 (Unit 1) 02/24/2026 (Unit 2) H.B. Robinson Steam Electric Plant Unit 2 Fifth 2007 Edition, Through 2008 Addenda 07/21/2012 02/19/2023 McGuire Nuclear Station Unit 1 Fifth2 2007 Edition, Through 2008 Addenda 12/01/2021 11/30/2031 McGuire Nuclear Station Unit 2 Fourth 2007 Edition, Through 2008 Addenda 07/15/2014 02/29/2024 Oconee Nuclear Station Units 1, 2, and 3 Fifth 2007 Edition, Through 2008 Addenda 07/15/2014 07/15/2024 Shearon Harris Nuclear Power Plant Unit 1 Fourth 2007 Edition, Through 2008 Addenda 09/09/2017 09/08/2027 Notes:

1. The Interval End Date is subject to change in accordance with IWA-2430(c)(1).

2. Reference Relief Request RA-20-0031 (ADAMS Accession No. ML20230A205) that allowed Duke Energy to implement the requirements of ASME Section XI, 2007 Edition with the 2008 Addenda for Period 1 of the 5th Interval.

3.0 APPLICABLE CODE REQUIREMENT: ASME Section XI IWB-2500(a), Table IWB-2500-1, examination Category B-B and IWC-2500(a), Table IWC-2500-1, Examination Categories C-A and C-B require examination of the following Item Nos.: Item No. B2.40 - Volumetric examination of essentially 100% of the weld length of weld during the first Section XI inspection interval. For successive inspection intervals the examination may be limited to one vessel among the group of vessels performing a similar function. The examination volume is shown in Figure IWB-2500-6. Item No. C1.10 - Volumetric examination of essentially 100% of the weld length of the cylindrical-shell-to-conical shell-junction welds and shell (or head)-to-flange welds during each Section XI inspection interval. In the case of multiple vessels of similar design, size, and service (such as steam generators, heat exchangers), the required examinations may be limited to one vessel or distributed among the vessels. The examination volume is shown in Figure IWC-2500-1. Item No. C1.20 - Volumetric examination of essentially 100% of the weld length of the head-to-shell weld during each Section XI inspection interval. In the case of multiple vessels of similar design, size, and service (such as steam generators, heat exchangers), the required examinations may be limited to one vessel or distributed among the vessels. The examination volume is shown in Figure IWC-2500-1. Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 9 of 29 Item No. C1.30 - Volumetric examination of essentially 100% of the weld length of the tubesheet-to-shell welds during each Section XI inspection interval. In the case of multiple vessels of similar design, size, and service (such as steam generators, heat exchangers), the required examinations may be limited to one vessel or distributed among the vessels. The examination volume is shown in Figure IWC-2500-2. Item No. C2.21 - Volumetric and surface examination of all nozzle welds at terminal ends of piping runs during each Section XI inspection interval. In the case of multiple vessels of similar design, size, and service (such as steam generators, heat exchangers), the required examinations may be limited to one vessel or distributed among the vessels. The examination area and volume are shown in Figures IWC-2500-4(a), (b), or (d). Item No. C2.22 - Volumetric examination of all nozzles inside radius sections at terminal ends of piping runs during each Section XI inspection interval. In the case of multiple vessels of similar design, size, and service (such as steam generators, heat exchangers), the required examinations may be limited to one vessel or distributed among the vessels. The examination volume is shown in Figures IWC-2500-4(a), (b), or (d). 4.0 REASON FOR REQUEST: The Electric Power Research Institute (EPRI) performed assessments in References [9.1] and [9.2] of the bases for the ASME Code, Section XI examination requirements specified for the above listed ASME Code, Section XI, Division 1 examination categories for steam generator welds and components. The assessments include a survey of inspection results from 74 domestic and international nuclear units and flaw tolerance evaluations using probabilistic fracture mechanics (PFM) and deterministic fracture mechanics (DFM). The Reference [9.1] and [9.2] reports concluded that the current ASME Code, Section XI ISI examinations can be deferred for some time with no impact to plant safety. Based on the conclusions of the two EPRI reports supplemented by plant-specific evaluations contained herein, Duke Energy is requesting an ISI examination deferral for the subject welds. 5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE: CNS1 For CNS1, Duke Energy is requesting an inspection alternative to the examination requirements of ASME Code, Section XI, Tables IWB-2500-1 and IWC-2500-1, for the following examination categories and item numbers: Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 10 of 29 ASME Category Item No. Description B-B B2.40 Steam generators (primary side), tubesheet-to-head weld C-A C1.20 Shell circumferential welds C-A C1.30 Tubesheet-to-shell weld C-B C2.21 Nozzle-to-shell (nozzle to head or nozzle to nozzle) welds C-B C2.22 Nozzle inside radius sections In 1996 (1st period of the 2nd inspection interval) the CNS1 SGs were replaced. The new SG welds and components received the required PSI examinations prior to service followed by ISI examinations through the 2nd period of the current 4th inspection interval. The following item numbers have been examined during the current 4th interval with no relevant indications: B2.40, C1.30, C2.22. The proposed alternative is to defer the ISI examinations for these Item Nos. for the replacement steam generators at CNS1 from the current ASME Code, Section XI, Division 1,10-year requirement through the end of the 5th inspection interval, which is currently scheduled to end on June 28, 2035. This equates to a maximum extension of 19 years, 10 months, 10 days from the end of the 3rd inservice inspection interval (8/18/2015) at which time all ASME Code, Section XI, Division 1 requirements were satisfied. CNS2 For CNS2, Duke Energy is requesting an inspection alternative to the examination requirements of ASME Code, Section XI, Tables IWB-2500-1 and IWC-2500-1, for the following examination categories and item numbers: ASME Category Item No. Description B-B B2.40 Steam generators (primary side), tubesheet-to-head weld C-A C1.10 Shell circumferential welds C-A C1.20 Head circumferential welds C-A C1.30 Tubesheet-to-shell weld C-B C2.21 Nozzle-to-shell (nozzle to head or nozzle to nozzle) welds C-B C2.22 Nozzle inside radius sections CNS2 is still using its original SGs. The SG welds and components received the required PSI examinations prior to service followed by ISI examinations through the 2nd period of the current 4th inspection interval. The following item numbers have been examined during the current 4th interval with no relevant indications: B2.40, C1.10, and C2.21. The proposed alternative is to defer the ISI examinations for these Item Nos. for the steam generators at CNS2 from the current ASME Code, Section XI, Division 1 10-year requirement through the end of the 5th inspection interval, which is currently scheduled to Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 11 of 29 end on August 18, 2035. This equates to an extension of 20 years from the end of the 3rd inservice inspection interval (8/18/2015) at which time all ASME Code, Section XI, Division 1 requirements were satisfied. MNS1/2 For MNS1/2, Duke Energy is requesting an inspection alternative to the examination requirements of ASME Code, Section XI, Tables IWB-2500-1 and IWC-2500-1, for the following examination categories and item numbers: ASME Category Item No. Description B-B B2.40 Steam generators (primary side), tubesheet-to-head weld C-A C1.20 Head circumferential welds C-A C1.30 Tubesheet-to-shell weld C-B C2.21 Nozzle-to-shell (nozzle to head or nozzle to nozzle) welds C-B C2.22 Nozzle inside radius sections In 1997 (2nd period of the 2nd inspection interval) the MNS1/2 SGs were replaced. The new SG welds and components received the required PSI examinations prior to service followed by all ISI examinations through the 4th inspection intervals for MNS1/2. The proposed alternative is to defer the ISI examinations for these Item Nos. for the replacement steam generators at MNS1/2 from the 5th inspection interval requirements through the end of the 6th inspection interval, which is currently scheduled to end on November 30, 2041 for MNS1 and February 29, 2044 for MNS2. This equates to an extension of 20 years from the end of the 4th inservice inspection interval (11/30/2021) at MNS1 and an extension of 20 years from the end of the 4th inservice inspection interval (2/29/2024) at MNS2, at which time all ASME Code, Section XI, Division 1 requirements were satisfied. ONS1/2/3 For ONS1/2/3, Duke Energy is requesting an inspection alternative to the examination requirements of ASME Code, Section XI, Tables IWB-2500-1 and IWC-2500-1, for the following examination categories and item numbers: ASME Category Item No. Description B-B B2.40 Steam generators (primary side), tubesheet-to-head weld C-A C1.30 Tubesheet-to-shell weld C-B C2.21 Nozzle-to-shell (nozzle to head or nozzle to nozzle) welds In 2003 and 2004 (3rd period of the 3rd inspection interval) the ONS1/2/3 SGs were replaced. The new SG welds and components received the required PSI examinations prior to service followed by ISI examinations through the 5th inspection intervals. Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 12 of 29 The proposed alternative is to defer the ISI examinations for these Item Nos. for the replacement steam generators at ONS1/2/3 from only 6th interval requirements. The end of the 5th inservice inspection interval (at which time all ASME Code, Section XI, Division 1 requirements will be satisfied) is scheduled for 7/15/2024 for all three Oconee units. HNP For HNP, Duke Energy is requesting an inspection alternative to the examination requirements of ASME Code, Section XI, Tables IWB-2500-1 and IWC-2500-1, for the following examination categories and item numbers: ASME Category Item No. Description B-B B2.40 Steam generators (primary side), tubesheet-to-head weld C-A C1.20 Head circumferential welds C-A C1.30 Tubesheet-to-shell weld C-B C2.21 Nozzle-to-shell (nozzle to head or nozzle to nozzle) welds C-B C2.22 Nozzle inside radius sections In 2001 (2nd period of the 2nd inspection interval) the HNP SGs were replaced. The new SG welds and components received the required PSI examinations prior to service followed by ISI examinations through the 2nd period of the current 4th inspection interval. The following item number have been examined during the current 4th interval with no relevant indications: C2.21. The proposed alternative is to defer the ISI examinations for these Item Nos. for the replacement steam generators at HNP from the current ASME Code, Section XI, Division 1 10-year requirement through the end of the 5th inspection interval, which is currently scheduled to end on May 1, 2037. This equates to an extension of 19 years, 7 months, 23 days from the end of the 3rd inservice inspection interval (9/8/2017) at which time all ASME Code, Section XI, Division 1 requirements were satisfied. RNP For RNP, Duke Energy is requesting an inspection alternative to the examination requirements of ASME Code, Section XI, Tables IWB-2500-1 and IWC-2500-1, for the following examination categories and item numbers: Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 13 of 29 ASME Category Item No. Description B-B B2.40 Steam generators (primary side), tubesheet-to-head weld C-A C1.10 Shell circumferential welds C-A C1.20 Head circumferential welds C-A C1.30 Tubesheet-to-shell weld C-B C2.21 Nozzle-to-shell (nozzle to head or nozzle to nozzle) welds C-B C2.22 Nozzle inside radius sections In 1984 (1st period of the 2nd inspection interval) the RNP SGs were replaced. The new SG welds and components received the required PSI examinations prior to service followed by ISI examinations through the 5th inspection interval. All required 5th interval ISI examinations for replacement steam generator Item Nos. were satisfied with no relevant indications. The proposed alternative is to defer these ISI examinations at RNP from the 6th interval requirements. Technical Basis A summary of the key aspects of the technical basis for this request is summarized below. The applicability of the technical basis to CNS1/2, MNS1/2, ONS 1/2/3, HNP and RNP is shown in Attachments 1 through 6. Applicability of the Degradation Mechanism Evaluation in References [9.1] and [9.2] to the Duke Energy PWR Units An evaluation of degradation mechanisms that could potentially impact the reliability of the SG welds and components was performed in References [9.1] and [9.2]. The degradation mechanisms that were evaluated included stress corrosion cracking (SCC), environmental assisted fatigue (EAF), microbiologically influenced corrosion (MIC), pitting, crevice corrosion, erosion-cavitation, erosion, flow accelerated corrosion (FAC), general corrosion, galvanic corrosion, and mechanical/thermal fatigue. Other than the potential for EAF and mechanical/thermal fatigue, there were no known active degradation mechanisms identified that significantly affect the long-term structural integrity of the SG welds and components covered in this request. This observation was acknowledged by the NRC in Section 3.8, page 6, second paragraph of the Reference [9.14] Safety Evaluation (SE) for Vogtle Units 1 & 2 and Section 2.0, page 3, second paragraph of the Reference [9.16] SE for Millstone Unit 2. As shown in Attachments 1 through 6, the materials and operating conditions for the plants considered in this Request for Alternative are similar to those in the References [9.1] and [9.2] and therefore, the conclusions of these Reports apply to the plants in this Request for Alternative. The fatigue-related mechanisms were considered in the PFM and DFM evaluations in References [9.1] and [9.2]. As part of the technical basis in References [9.1 and 9.2], a comprehensive industry survey involving 74 PWR units was conducted to determine the degradation history of these components. The survey reviewed examination results from the start of plant operation. Most of these plants have operated for over 30 years and in some cases over 40 years. The results showed that no examinations identified any unknown degradation Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 14 of 29 mechanisms (i.e., mechanisms other than those listed above). Based on this exhaustive industry survey, it is concluded that although the emergence of an unknown degradation mechanism cannot be completely ruled out, the possibility of the occurrence of such an unknown degradation mechanism is highly unlikely. Applicability of the Stress Analysis in References [9.1] and [9.2] to the Duke Energy PWR Units Finite element analyses (FEA) were performed in References [9.1] and [9.2] to determine the stresses in the SG welds and components covered in this request. The finite element models used in References [9.1] and [9.2] are consistent with the configurations of CNS1/2, MNS1/2, ONS 1/2/3, HNP and RNP and therefore no new FEA model is required for the stress analysis of these plants. The analysis in References [9.1] and [9.2] was performed using representative pressurized water reactor (PWR) geometries, bounding transients, and typical material properties. The results of the stress analyses were used in a flaw tolerance evaluation. The applicability of the FEA analysis to CNS1/2, MNS1/2, ONS 1/2/3, HNP and RNP is demonstrated in Attachments 1 through 6 and confirms that all plant-specific requirements are met. In particular, the key geometric parameters used in the Reference [9.1] and [9.2] stress analyses are compared to those of the Duke Energy PWR units in Tables 1 and 2: Table 1. SG Vessel Dimensions Plant Primary Lower Head ID (in) Primary Lower Head Thk (in) Primary Lower Head Ri/t Secondary Upper Shell ID (in) Secondary Upper Shell Thk (in) Secondary Upper Shell Ri/t EPRI Report (Table 4-2 of [9.2]) 155.33 6.94 11.2 230.87 4.91 23.5 CNS1 123.63 6.625 9.3 168.5 4.125 20.4 CNS2 125.62 7.44 8.4 168.5 3.88 21.7 MNS1 123.63 6.625 9.3 168.5 4.125 20.4 MNS2 123.63 6.625 9.3 168.5 4.125 20.4 ONS1 119.06 6.06 9.8 137.88 3.13 22 ONS2 119.06 6.06 9.8 137.88 3.13 22 ONS3 119.06 6.06 9.8 137.88 3.13 22 HNP 125.18 5.26 11.9 168.5 3.88 21.7 RNP 118.88 5.22 11.4 158.76 3.62 21.9 Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 15 of 29 Table 2. SG Nozzle Dimensions Plant FW Nzl ID (in) FW Nzl Thk (in) FW Nzl Ri/t MS Nzl ID (in) MS Nzl Thk (in) MS Nzl Ri/t EPRI Report (Figures 4-8, 4-9 and 4-10 of [9.1]) 16.5 6 1.38 28 (West.) 22.25 (B&W) 6.5 (West.) 4.53 (B&W) 2.15 (West.) 2.46 (B&W) CNS1 14.812 4.50 1.65 N/A(1) N/A(1) N/A(1) CNS2 14.32 5.84 1.23 29.375 5.56 2.64 MNS1 14.812 4.50 1.65 N/A(1) N/A(1) N/A(1) MNS2 14.812 4.50 1.65 N/A(1) N/A(1) N/A(1) ONS1 N/A(1) N/A(1) N/A(1) 22.265 6.49 1.72 ONS2 N/A(1) N/A(1) N/A(1) 22.265 6.49 1.72 ONS3 N/A(1) N/A(1) N/A(1) 22.265 6.49 1.72 HNP 14.32 5.84 1.23 N/A(1) N/A(1) N/A(1) RNP 16.625 3.5 1.39 28.25 3.5 4.04 Notes:

1.

Nozzle has no C2.21 or C2.22 components. As discussed in Sections 4.3.3 and 4.6 of Reference [9.1] and noted by the NRC in Section 3.8.3.1, page 9, third paragraph of the SER for Vogtle [9.14], the dominant stress is the pressure stress. Therefore, the variation in the Ri/t ratio determined in Tables 1 and 2 can be used to scale up the stresses of the Reference [9.1] and [9.2] reports to obtain the plant-specific stresses for each unit and component. In the fracture mechanic evaluations performed in Reference [9.1], the transients that contribute the most to crack growth are the heatup and cooldown events. A total of 300 such events were evaluated during a 60-year plant life. Reference [9.1] also used a conservative rate of 200°F per hour which would exceed the limit allowed by the Catawba, McGuire, Oconee, Harris, and Robinson technical specifications. Due to small variations from the values in Reference [9.1], the conservative definition of heatup and cooldown transient, and the low number of transient cycles predicted in Attachments 1 through 6 over the life of the plant versus the number of cycles evaluated in Reference [9.1], it is determined that the stress analysis performed in Reference [9.1] is applicable to Catawba, Units 1 and 2, McGuire, Units 1 and 2, Oconee, Units 1, 2, and 3, Harris, Unit 1, and Robinson, Unit 2. In the selection of the transients in Section 5 of References [9.1] and [9.2] and the subsequent stress analyses in Section 7, test conditions beyond a system leakage test were not considered since system leakage tests at CNS1/2, MNS1/2, ONS 1/2/3, HNP and RNP are performed at normal operating conditions. No hydrostatic testing had been performed at CNS1/2, MNS1/2, ONS 1/2/3, HNP or RNP since the plants went into operation. In Reference [9.2], clad residual stress was not considered for the primary side welds. In a previous NRC RAI (Reference [9.17], RAI 3c), the NRC raised this issue. In response to the RAI (Reference [9.18], RAI Response 2.c), an evaluation was performed which showed that the clad residual stress has no significant impact on the conclusions Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 16 of 29 of Reference [9.2] and this was found acceptable by the NRC in Section 5.3 of Reference [9.16]. Applicability of the Flaw Tolerance Evaluation in References [9.1] and [9.2] to the Duke Energy PWR Units Flaw tolerance evaluations were performed in References [9.1] and [9.2] consisting of probabilistic fracture mechanics (PFM) evaluations and confirmatory deterministic fracture mechanics (DFM) evaluations. The Reference [9.1] and [9.2] reports were developed consistent with the recommendations provided in EPRIs White Paper on suggested content for PFM submittals [9.12], NRC Regulatory Guide 1.245 for PRM submittals [9.19] and the associated technical basis [9.20]. The results of the PFM analyses indicate that, after a preservice inspection (PSI) followed by subsequent in-service inspections (ISI), the U.S. Nuclear Regulatory Commissions (NRCs) safety goal of 10-6 failures per year is met. The PFM analysis in Reference [9.1] was performed using the PRobabilistic OptiMization of InSpEction (PROMISE) Version 1.0 software, developed by Structural Integrity Associates. As part of the NRCs review of Southern Nuclears alternative request, the NRC performed an audit of the PROMISE Version 1.0 software as discussed in the NRCs audit plan dated May 14, 2020 (ADAMS Accession No. ML20128J311). The PFM analysis in Reference [9.2] was performed using the PROMISE Version 2.0 software which has not been audited by the NRC. The only technical difference between the two versions is that in PROMISE Version 1.0, the user-specified examination coverage is applied to all inspections, whereas in PROMISE Version 2.0, the examination coverage can be specified by the user uniquely for each inspection. In both Versions 1.0 and 2.0, the software assumes 100% coverage for the PSI examination. In Section 8.2.2.2 of Reference [9.1] and Section 8.3.2.2 of Reference [9.2], a nozzle flaw density of 0.001 flaws per nozzle was assumed for the nozzle inside radius sections. In Section 3.8.5 of the SE for Vogtle in Reference [9.14], the NRC indicated that a nozzle flaw density of 0.1 flaws per nozzle should have been used. Sensitivity studies performed in Section 8.2.4.3.4 in Reference [9.2] indicated that by changing the number of flaws in the nozzle inside radius sections from 0.001 to 0.1, the probabilities of leak and rupture increased by two orders of magnitude but were still significantly below the acceptance criterion of 1x10-6 per year. A comparison of the PSI/ISI scenarios used in the sensitivity studies performed in References [9.1] and [9.2] to those at the Duke Energy PWR units is provided below. Note that the assumption below of a 30-year ISI deferral is conservative compared to the end of currently licensed operating life for each plant. CNS1 For the CNS Unit 1 replacement SGs installed in 10/4/96, PSI examinations have been performed followed by ISI examinations in the two 10-year intervals following SG replacement (the unit is currently in the 4th ISI interval). The PSI/ISI scenario considered is therefore PSI plus two 10-year ISI examination to be followed by a 30-year ISI deferral (PSI+10+20+50). Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 17 of 29 CNS2 For the CNS Unit 2 original SGs, PSI examinations have been performed followed by ISI examinations over three 10-year intervals (the unit is currently in its 4th ISI interval). The PSI/ISI scenario considered is therefore PSI plus three 10-year ISI examination to be followed by a 30-year ISI deferral (PSI+10+20+30+60). MNS1 For the MNS Unit 1 replacement SGs installed in 5/19/97, PSI examinations have been performed followed by ISI examinations in the two 10-year intervals following SG replacement. The PSI/ISI scenario considered is therefore PSI plus two 10-year ISI examinations to be followed by a 30-year ISI deferral (PSI+10+20+50). MNS2 For the MNS Unit 2 replacement SGs installed in 12/18/97, PSI examinations have been performed followed by ISI examinations in the two 10-year intervals following SG replacement. The PSI/ISI scenario considered is therefore PSI plus two 10-year ISI examinations to be followed by a 30-year ISI deferral (PSI+10+20+50). ONS1/2/3 For the ONS Units 1, 2 and 3 replacement SGs installed in 9/20/03, 3/20/04 and 10/9/04, respectively, PSI examinations have been performed followed by ISI examinations in the two 10-year intervals following SG replacement. The PSI/ISI scenario considered is therefore PSI plus two 10-year ISI examination to be followed by a 30-year ISI deferral (PSI+10+20+50). HNP For the HNP replacement SGs installed in 9/22/01, PSI examinations have been performed followed by ISI examinations in the two 10-year intervals following SG replacement. The PSI/ISI scenario considered is therefore PSI plus two 10-year ISI examination to be followed by a 30-year ISI deferral (PSI+10+20+50). RNP For the RNP replacement SGs installed in 1/26/84, PSI examinations have been performed followed by ISI examinations in the four 10-year intervals following SG replacement. The PSI/ISI scenario considered is therefore PSI plus four 10-year ISI examination to be followed by a 30-year ISI deferral (PSI+10+20+30+40+70). Limiting PSI/ISI Scenarios The most limiting PSI/ISI scenario for the Duke Energy plants is (PSI+10+20+50). This scenario was not specifically considered in the Reference [9.1] and [9.2] PFM evaluations in combination with key variables, as evaluated by the NRC in Section 4.0 (page 6) of the Reference [9.14] Safety Evaluation. The Westinghouse plants have Item Nos. C2.21 and C2.22 for both the feedwater and main steam nozzles. From Reference [9.1], the limiting component for Item Nos. C2.21 and C2.22 for the Westinghouse plants Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 18 of 29 is the feedwater nozzle. The B&W plants do not have Item Nos. C2.21 and C2.22 for the feedwater nozzles, only for the main steam nozzles. Hence, five separate evaluations are performed to bound the Duke Energy PWR fleet using a PSI/ISI scenario of (PSI+10+20+50).

1. Westinghouse feedwater nozzle inside radius sections

2. Westinghouse feedwater nozzle to-shell welds

3. B&W main steam nozzle inside radius sections

4. B&W main steam nozzle-to-shell welds

5. The remainder of the SG welds Westinghouse Feedwater Nozzle Inside Radius Section From Reference [9.1], the critical location for the inside radius section is feedwater nozzle Case ID FEW-P1N. An evaluation similar to that shown in Table 8-28 of Reference [9.1] was performed for this location assuming a nozzle flaw density of 0.1, a fracture toughness of 200 ksiin and a standard deviation 5 ksiin as recommended by the NRC in Reference [9.14]. A stress multiplier of 1.75 was applied. This stress multiplier was chosen to result in probability of rupture or probability of leakage close to the acceptance criteria after 80 years. The results of the evaluation, using PROMISE Version 1.0, are summarized in Table 3 and show that after 80 years of plant operation the probabilities of rupture and leakage are below the acceptance criterion of 1.0x10-6.

Table 3. Sensitivity to Combined Effects of Fracture Toughness, Stress, and Nozzle Flaw Density for 80 Years for Westinghouse Feedwater Nozzle Inside Radius Section (Case ID FEW-P1N from Reference [9.1]) Time (yr) Probability per Year for Combined Case KIC = 200 ksiin. SD = 5 ksiin. Stress Multiplier = 1.75 Nozzle Flaw Density = 0.1 PSI+10+20+50 Rupture Leak 10 3.97E-07 1.58E-07 20 2.41E-07 9.80E-08 30 1.60E-07 6.57E-08 40 1.22E-07 4.98E-08 50 1.06E-07 4.26E-08 60 8.85E-08 3.55E-08 70 7.60E-08 3.04E-08 80 6.65E-08 2.66E-08 Westinghouse Feedwater Nozzle-to-Shell Weld For the feedwater nozzle-to-shell weld, Table 8-16 of Reference [9.1] indicates that the critical Case ID is FEW-P3A. For the evaluation, a flaw density of 1 flaw per weld was Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 19 of 29 assumed. A fracture roughness of 200 ksiin and standard deviation 5 ksiin were also used. A stress multiplier of 1.5 was applied such that probability of rupture or probability of leakage are close to the acceptance criteria after 80 years. The results of the evaluation, using PROMISE Version 1.0, are summarized in Table 4 and show that after 80 years of plant operation the probabilities of rupture and leakage are below the acceptance criterion of 1.0x10-6. Table 4. Sensitivity to Combined Effects of Fracture Toughness, Stress, and Nozzle Flaw Density for 80 Years for Westinghouse Feedwater Nozzle-to-Shell Weld (Case ID FEW-P3A from Reference [9.1]) Time (year) Probability per Year for Combined Case KIC = 200 ksiin. SD = 5 ksiin. Stress Multiplier = 1.45 Flaw Density = 1 PSI+10+20+50 Rupture Leak 10 1.00E-08 1.00E-08 20 5.00E-09 1.00E-08 30 3.33E-09 6.67E-09 40 2.50E-09 3.75E-08 50 2.00E-09 7.48E-07 60 1.67E-09 6.30E-07 70 1.43E-09 5.63E-07 80 1.25E-09 5.61E-07 B&W Main Steam Nozzle Inside Radius Section From Reference [9.1], the critical Case ID for the main steam nozzle inside radius section is SGB-P1N. An evaluation similar to that shown in Table 8-28 of Reference [9.1] was performed for this location assuming a nozzle flaw density of 0.1, a fracture toughness of 200 ksiin and a standard deviation 5 ksiin as described by the NRC in Reference [9.13]. A relatively high stress multiplier of 2.0 was applied. The results of the evaluation, using PROMISE Version 1.0, are summarized in Table 5 and show that after 80 years of plant operation the probabilities of rupture and leakage are well below the acceptance criterion of 1.0x10-6 by at least three orders of magnitude. The results indicate that a much higher stress multiplier than 2.0 could have been used and the acceptance criteria would still be met. Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 20 of 29 Table 5. Sensitivity to Combined Effects of Fracture Toughness, Stress, and Nozzle Flaw Density for 80 Years for B&W Main Steam Nozzle Inside Radius Section (Case ID SGB-P1N from Reference [9.1]) Time (year) Probability per Year for Combined Case KIC = 200 ksiin. SD = 5 ksiin. Stress Multiplier = 2.45 Nozzle Flaw Density = 0.1 PSI+10+20+50 Rupture Leak 10 3.00E-09 1.00E-09 20 3.00E-09 5.00E-10 30 2.67E-09 3.33E-10 40 3.63E-08 2.50E-10 50 7.38E-07 1.60E-09 60 6.18E-07 1.33E-09 70 5.38E-07 1.43E-09 80 4.82E-07 2.50E-09 B&W Main Steam Nozzle-to-Shell Welds For the main steam nozzle-to-shell weld, Table 8-15 of Reference [9.1] indicates that the critical Case ID is SGB-P3A. For the evaluation, a flaw density of 1.0 flaw per weld was assumed, consistent with the evaluations in Reference [9.1]. A fracture toughness of 200 ksiin and standard deviation of 5 ksiin were also used, A relatively high stress multiplier of 2.0 was applied. The results of the evaluation, using PROMISE Version 1.0, are summarized in Table 6 and show that after 80 years of plant operation the probabilities of rupture and leakage are well below the acceptance criterion of 1.0x10-6. The results indicate that a much higher stress multiplier than 2.0 could have been used and the acceptance criteria would still be met. Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 21 of 29 Table 6. Sensitivity to Combined Effects of Fracture Toughness, Stress, and Nozzle Flaw Density for 80 Years for the B&W Main Steam Nozzle-to-Shell Weld Case ID SGB-P3A from Reference [9.1]) Time (yr) Probability per Year for Combined Case KIC = 200 ksiin. SD = 5 ksiin. Stress Multiplier = 2.2 Flaw Density = 1 PSI+10+20+50 Rupture Leak 10 2.00E-08 1.00E-08 20 1.00E-08 5.00E-09 30 6.67E-09 3.33E-09 40 2.00E-08 2.50E-09 50 8.50E-07 2.00E-09 60 7.15E-07 1.67E-09 70 6.40E-07 1.43E-09 80 6.33E-07 1.25E-09 Remainder of the SG Welds For the remaining SG welds, Table 8-32 of Reference [9.2] indicates that the critical Case ID is SGPTH-P4A. This case was evaluated for the inspection scenario of PSI+10+20+50, a flaw density of 1.0 flaw per weld, a fracture toughness of 200 ksiin and a standard deviation 5 ksiin. A relatively high with a stress multiplier of 2.0 was applied. The results of the evaluation, using PROMISE Version 2.0, are summarized in Table 7 and show that after 80 years of plant operation the probabilities of rupture and leakage are well below the acceptance criterion of 1.0x10-6. The results indicate that a much higher stress multiplier than 2.0 could have been used and the acceptance criteria would still be met. Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 22 of 29 Table 7. Sensitivity to Combined Effects of Fracture Toughness, Stress, and Nozzle Flaw Density for 80 Years for the SG Welds (Westinghouse or B&W) (Case ID SGPTH-P4A from Reference [9.2]) Time (year) Probability per Year for Combined Case KIC = 200 ksiin. SD = 5 ksiin. Stress Multiplier = 2.1 Nozzle Flaw Density = 1 PSI+10+20+50 Rupture Leak 10 2.20E-07 1.00E-08 20 2.20E-07 5.00E-09 30 1.53E-07 3.33E-09 40 1.43E-07 2.50E-09 50 3.24E-07 2.00E-09 60 2.70E-07 1.67E-09 70 2.40E-07 1.43E-09 80 2.26E-07 1.25E-09 The plant-specific PFM evaluations presented in Tables 3 to 7 above for the Duke Energy plants indicate that with conservative inputs of the critical parameters, the probabilities of rupture and leakage are well below the acceptance criterion of 1.0x10-6 failures per year. The stress multipliers applied in Tables 3 through 7 are greater than the ratio of R/t of the Duke Energy plants shown in Tables 1 and 2 relative to that of the model in the EPRI reports and therefore the analysis in Tables 3 through 7 are conservative. It should also be noted that the analyses involve most limiting case with regard to the PSI/ISI scenarios. Furthermore, the evaluation was performed for 80 years, which his longer than the deferral being sought by Duke Energy in this Request for Alternative. The PFM evaluations documented in References [9.1] and [9.2] and the plant-specific evaluations above used a Section XI, Appendix VIII-based probability of detection (POD) curve in the PFM evaluation because most ISI examinations of major plant Class 1 and Class 2 components are performed using Appendix VIII procedures. However, for Class 2 components, the use of Appendix VIII procedures is plant-specific. In most cases of the Duke Energy plants in this Request for Alternative, ASME Code, Section V procedures were used for the Class 2 components unless noted in the plant specific Attachments 1 - 6. Recently, examinations of Class 2 components were performed using Appendix VIII procedures. Based on the observations made by the NRC in Section 3.8.8.2, page 21 of the Vogtle SE [9.14], the use of the ASME Code, Section XI, Appendix VIII based POD curve for inspections based on ASME Code, Section V procedures would have minimal impact of the PFM results since the POD curve is not one of the parameters that significantly affect the PFM results. The DFM evaluations in Table 8-31 of Reference [9.1] and Table 8-3 of Reference [9.2] provide verification of the above PFM results for the Duke Energy PWR units by Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 23 of 29 demonstrating that it takes approximately 80 years for a postulated flaw with an initial depth equal to ASME Code, Section XI acceptance standards to grow to a depth where the maximum stress intensity factor (K) exceeds the ASME Code, Section XI allowable fracture toughness. Inspection History As described in Section 8.2.4.1.1 of Reference [9.1] and Section 8.3.4.1 of Reference [9.2], preservice examination (PSI) refers to the collective examinations required by ASME Code, Section III during fabrication and any ASME Code, Section XI examinations performed prior to service. The Section III fabrication examinations required for these components were robust and any Section XI preservice examinations further contributed to thorough initial examinations. CNS1 Inspection history for CNS1 (including examinations performed to date, examination findings, inspection coverage, and relief requests) is presented in Attachment 1. As shown in the attachment, all welds/components have examinations coverage greater than 90% (essentially 100%). As shown in Attachment 1, no flaws that exceeded the ASME Code, Section XI acceptance standards were identified during any examinations. CNS2 Inspection history for CNS2 (including examinations performed to date, examination findings, inspection coverage, and relief requests) is presented in Attachment 2. As shown in the attachment, some of the welds/components have limited exam coverage, with the minimum coverage being 46.872%. Section 8.3.5 of Reference [9.1] and Section 8.2.5 of Reference [9.2] discuss limited coverage and determine that the conclusions of the report are applicable to components with limited coverage. In addition, it is important to note that all other inspection activities, including the system leakage test (Examination Categories B-P and C-H) will continue to be performed in accordance with the ASME Section XI requirements, providing further assurance of safety. As shown in Attachment 2, no flaws that exceeded the ASME Code, Section XI acceptance standards were identified during any examinations. MNS1/2 Inspection history for MNS1/2 (including examinations performed to date, examination findings, inspection coverage, and relief requests) is presented in Attachment 3. As shown in the attachment, all welds/components have examinations coverage greater than 90% (essentially 100%). As shown in Attachment 3, no flaws that exceeded the ASME Code, Section XI acceptance standards were identified during any examinations. ONS1/2/3 Inspection history for ONS1/2/3 (including examinations performed to date, examination findings, inspection coverage, and relief requests) is presented in Attachment 4. As shown in the attachment, some of the welds/components have limited exam coverage, with the minimum coverage being 75.10%. Examination coverage greater than 50% is Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 24 of 29 acceptable per Section 3.8.7 of the Vogtle SE [9.14]. As shown in Attachment 4, no flaws that exceeded the ASME Code, Section XI acceptance standards were identified during any examinations. HNP Inspection history for HNP (including examinations performed to date, examination findings, inspection coverage, and relief requests) is presented in Attachment 5. As shown in the attachment, some of the welds/components have limited exam coverage, with the minimum coverage being 70.10%. Examination coverage greater than 50% is acceptable per Section 3.8.7 of the Vogtle SE [9.14]. As shown in Attachment 5, no flaws that exceeded the ASME Code, Section XI acceptance standards were identified during any examinations. RNP Inspection history for RNP (including examinations performed to date, examination findings, inspection coverage, and relief requests) is presented in Attachment 6. As shown in the attachment, some of the welds/components have limited exam coverage, with the minimum coverage being 80.42%. Examination coverage greater than 50% is acceptable per Section 3.8.7 of the Vogtle SE [9.14]. As shown in Attachment 6, no flaws that exceeded the ASME Code, Section XI acceptance standards were identified during any examinations. Industry Survey The inspection history for these components as obtained from an industry survey is presented in Attachment 7. The results of the survey indicate that these components are very flaw tolerant. Conclusion It is concluded that the SG pressure-retaining welds and full penetration welded nozzles are very flaw tolerant. PFM and DFM evaluations performed as part of the technical basis reports [9.1] and [9.2], supplemented by plant-specific evaluations performed as part of this Request for Alternative, demonstrate that using conservative PSI/ISI inspection scenarios for all plants, the NRC safety goal of 10-6 failures per reactor year is met with considerable margins. Plant-specific applicability of the technical basis to CNS1/2, MNS1/2, ONS 1/2/3, HNP and RNP is demonstrated in Attachments 1 through

6. The requested ISI deferrals provide an acceptable level of quality and safety in lieu of the current ASME Code, Section XI 10-year inspection frequency.

Operating and examination experience demonstrates that these components have performed with very high reliability, mainly due to their robust design. Attachments 1 through 6 show the examination history for the SG welds examined in the two most recent 10-year inspection intervals. In addition to the required PSI examinations for these SG welds and components, all Duke Energy units have performed multiple ISI examinations through the current 10-year inspection interval at each plant. Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 25 of 29 No flaws that exceeded the ASME Code, Section XI acceptance standards were identified during any examinations, as shown in Attachments 1 through 6. Some examinations listed in Attachments 4, 5 and 6 for ONS1/2/3, HNP and RNP have limited examination coverage (less than 90%). However, all coverage was greater than 50%, which was determined to be acceptable per Section 3.8.7 of the Vogtle SE [9.14]. This is consistent with Section 8.3.5 of Reference [9.1] and Section 8.2.5 of Reference [9.2], which discuss limited coverage and determined that the conclusions of the reports are applicable to components with limited coverage. In addition, it is important to note all other inspection activities, including the system leakage test (Examination Categories B-P and C-H) will continue to be performed in accordance with the ASME Code, Section XI requirements, providing further assurance of safety. Finally, as discussed in Reference [9.3], for situations where no active degradation mechanism is present, it was concluded that subsequent ISI examinations do not provide additional value after PSI has been performed and the inspection volumes have been confirmed to be free of defects. Therefore, Duke Energy requests the NRC grant this proposed alternative in accordance with 10 CFR 50.55a(z)(1). 6.0 DURATION OF PROPOSED ALTERNATIVE: Catawba Nuclear Station Units 1 and 2 The proposed alternative is requested for the remainder of the 4th inspection interval and through the end of the 5th inspection interval, which is currently scheduled to end on June 28, 2035 for CNS1 and August 18, 2035 for CNS2. McGuire Nuclear Station Units 1 and 2 The proposed alternative is requested for the 5th inspection interval and through the end of the 6th inspection interval, which is currently scheduled to end on November 30, 2041 for MNS1 and on February 29, 2044 for MNS2. Oconee Nuclear Station Units 1, 2, and 3 The proposed alternative is requested for the duration of the 6th inspection interval, which is currently scheduled to end on July 14, 2034 for ONS1/2/3. Harris Nuclear Plant The proposed alternative is requested for the remainder of the 4th inspection interval and through the end of the 5th inspection interval, which is currently scheduled to end on May 1, 2037. Robinson Nuclear Plant The proposed alternative is requested for the duration of the 6th inspection interval, which is currently scheduled to end on February 19, 2033. Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 26 of 29

7.0 PRECEDENTS

The following previous submittal has been made by Southern Nuclear to provide relief from the ASME Code, Section XI Examination Category C-B (Item Nos. C2.21 and C2.22) surface and volumetric examinations based on the Reference [9.1] technical basis report: Letter from C. A. Gayheart (Southern Nuclear) to the U.S. NRC, Vogtle Electric Generating Plant, Units 1 & 2 Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 Version 2.0, dated September 9, 2020, ADAMS Accession No. ML20253A311 [9.13]. The USNRC issued a safety evaluation of the Southern Nuclear request for alternative on January 11, 2021. Letter from Michael T. Markley (USNRC) to Cheryl A. Gayheart (Southern Nuclear), Vogtle Electric Generating Plant, Units 1 & 2 - Relief Request for Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 to the Requirements of ASME Code (EPID L-2020-LLR-0109), dated January 11, 2021, ADAMS Accession No. ML20352A155 [9.14]. The following previous submittal has been made by Dominion Energy to provide relief from the ASME Section XI Examination Category B-B (Item No. B2.40) and Category C-A (Item Nos. C1.10, C1.20 and C1.30) surface and volumetric examinations based on the Reference [9.2] technical basis report: Letter from Mark D. Sartain (Dominion Energy) to the U.S. NRC, Dominion Energy Nuclear Connecticut, Inc. Millstone Power Station Unit 2 Alternative Request RR-05 Inspection Interval Extension for Steam Generator Pressure-Retaining Welds and Full-Penetration Welded Nozzles, dated July 15, 2020, ADAMS Accession No. ML20198M682 [9.15]. The USNRC issued a safety evaluation of the Dominion Energy request for alternative on July 16, 2021. Letter from James G. Danna (USNRC) to Daniel G. Stoddard (Dominion Energy), Millstone Power Station Unit 2 - Authorization and Safety Evaluation for Alternative Request No. RR-05-06 (EPID L-2020-LLR-0097), dated July 16, 2021, ADAMS Accession No. ML21167A355 [9.16]. In addition, the following is a list of approved actions (including relief requests and topical reports) related to inspections of SG welds and components: Letter from J. W. Clifford (NRC) to S. E. Scace (Northeast Nuclear Energy Company), Safety Evaluation of the Relief Request Associated with the First and Second 10-Year Interval of the Inservice Inspection (ISI) Plan, Millstone Nuclear Power Station, Unit 3 (TAC No. MA 5446), dated July 24, 2000, ADAMS Accession No. ML003730922. Letter from R. L. Emch (NRC) to J. B. Beasley, Jr. (SNOC), Second 10-Year Interval Inservice Inspection Program Plan Requests for Relief 13, 14, 15, 21 and Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 27 of 29 33 for Vogtle Electric Generating Plant, Units 1 and 2 (TAC No. MB0603 and MB0604), dated June 20, 2001, ADAMS Accession No. ML011640178. Letter from T. H. Boyce (NRC) to C. L. Burton (CP&L), Shearon Harris Nuclear Power Plant Unit 1 - Request for Relief 2R1-019, 2R1-020, 2R1-021, 2R1-022, 2R2-009, 2R2-010, 2R2-011 for the Second Ten-Year Interval Inservice Inspection Program Plan (TAC Nos. ME0609, ME0610, ME0611, ME0612, ME0613, ME0614 and ME0615), dated January 7, 2010, ADAMS Accession No. ML093561419. Letter from M, Khanna (NRC) to D. A. Heacock (Dominion Nuclear Connecticut Inc.), Millstone Power Plant Unit No. 2 - Issuance of Relief Requests RR-89-69 Through RR-89-78 Regarding Third 10-Year Interval Inservice Inspection plan (TAC Nos. ME5998 Through ME6006), dated March 12, 2012, ADAMS Accession No. ML120541062. Letter from R. J. Pascarelli (NRC) to E. D. Halpin (PG&E), Diablo Canyon Plant, Units 1 and 2 - Relief Request; NDE SG-MS-IR, Main Steam Nozzle Inner Radius Examination Impracticality, Third 10-Year Interval, American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI, Inservice Inspection Program (CAC Nos. MF6646 and MF6647), dated December 8, 2015, ADAMS Accession No. ML15337A021. In addition, there are precedents related to similar topical reports that justify relief for Class 1 nozzles: Based on studies presented in Reference [9.4], the NRC approved extending PWR reactor vessel nozzle-to-shell welds from 10 to 20 years in Reference [9.5]. Based on work performed in BWRVIP-108 [9.6] and BWRVIP-241 [9.8], the NRC approved the reduction of BWR vessel feedwater nozzle-to-shell weld examinations (Item No. B3.90 for BWRs from 100% to a 25% sample of each nozzle type every 10 years) in References [9.7] and [9.9]. The work performed in BWRVIP-108 and BWRVIP-241 provided the technical basis for ASME Code Case N-702 [9.10], which was conditionally approved by the NRC in Revision 19 of Regulatory Guide 1.147 [9.11]. Note that in Revision 20 of Regulatory Guide 1.147 ASME Code Case N-702-1 has been approved for use without conditions as shown in Table 1.

8.0 ACRONYMS

ASME American Society of Mechanical Engineers B&W Babcock and Wilcox BWR Boiling Water Reactor BWRVIP Boiling Water Reactor Vessel and Internals Program CE Combustion Engineering CFR Code of Federal Regulations DFM Deterministic fracture mechanics EAF Environmentally assisted fatigue EPRI Electric Power Research Institute FAC Flow accelerated corrosion FEA Finite element analysis FW Feedwater ISI Inservice Inspection Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 28 of 29 MIC Microbiologically influenced corrosion MS Main Steam NPS Nominal pipe size NRC Nuclear Regulatory Commission NSSS Nuclear steam supply system O.D. Outside diameter POD Probability of detection PFM Probabilistic fracture mechanics PSI Preservice inspection PWR Pressurized Water Reactor SCC Stress corrosion cracking SG Steam Generator WEC Westinghouse Electric Company

9.0 REFERENCES

9.1 Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Inside Radius Sections. EPRI, Palo Alto, CA: 2019. 3002014590 9.2 Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906. 9.3 American Society of Mechanical Engineers, Risk-Based Inspection: Development of Guidelines, Volume 2-Part 1 and Volume 2-Part 2, Light Water Reactor (LWR) Nuclear Power Plant Components. CRTD-Vols. 20-2 and 20-4, ASME Research Task Force on Risk-Based Inspection Guidelines, Washington, D.C., 1992 and 1998. 9.4 B. A. Bishop, C. Boggess, N. Palm, Risk-Informed extension of the Reactor Vessel In-Service Inspection Interval, WCAP-16168-NP-A, Rev. 3, October 2011. 9.5 US NRC, Revised Safety Evaluation by the Office of Nuclear Reactor Regulation; Topical Report WCAP-16168-NP-A, Revision 2, Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval, Pressurized Water Reactor Owners Group, Project No. 694, July 26, 2011, ADAMS Accession No. ML111600303.

9.6 BWRVIP-108

BWR Vessels and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA 2002. 1003557. 9.7 US NRC, Safety Evaluation of Proprietary EPRI Report, BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108), December 19, 2007, ADAMS Accession No. ML073600374.

9.8 BWRVIP-241

BWR Vessels and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA 2010. 1021005. Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1) Page 29 of 29 9.9 US NRC, Safety Evaluation of Proprietary EPRI Report, BWR Vessel and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii (BWRVIP-241), April 19, 2013, ADAMS Accession Nos. ML13071A240 and ML13071A233. 9.10 Code Case N-702, Alternate Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds, ASME Code Section XI, Division 1, Approval Date: February 20, 2004. 9.11 U. S. NRC Regulatory Guide 1.147, Revision 19, Inservice Inspection Code Case Acceptability, ASME Code Section XI, Division 1, dated October 2019. 9.12 N. Palm (EPRI), BWR Vessel & Internals Project (BWRVIP) Memo No. 2019-016, White Paper on Suggested Content for PFM Submittals to the NRC, February 27, 2019, ADAMS Accession No. ML19241A545. 9.13 Letter from C. A. Gayheart (Southern Nuclear) to the U.S. NRC, Vogtle Electric Generating Plant, Units 1 & 2 Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 Version 2.0, dated September 9, 2020, ADAMS Accession No. ML20253A311. 9.14 Letter from Michael T. Markley (USNRC) to Cheryl A. Gayheart (Southern Nuclear), Vogtle Electric Generating Plant, Units 1 & 2 - Relief Request for Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 to the Requirements of ASME Code (EPID L-2020-LLR-0109), dated January 11, 2021, ADAMS Accession No. ML20352A155. 9.15 Letter from Mark D. Sartain (Dominion Energy) to the U.S. NRC, Dominion Energy Nuclear Connecticut, Inc. Millstone Power Station Unit 2 Alternative Request RR-05 Inspection Interval Extension for Steam Generator Pressure-Retaining Welds and Full-Penetration Welded Nozzles, dated July 15, 2020, ADAMS Accession No. ML20198M682. 9.16 Letter from James G. Danna (USNRC) to Daniel G. Stoddard (Dominion Energy), Millstone Power Station Unit 2 - Authorization and Safety Evaluation for Alternative Request No. RR-05-06 (EPID L-2020-LLR-0097), dated July 16, 2021, ADAMS Accession No. ML21167A355. 9.17 Email Letter from R. Guzman (USNRC) to S. Sinha (Dominion Energy Nuclear Connecticut, Inc.), Millstone Unit 2 - Request for Additional Information - Alternative Request RR-05-06 Inspection Interval Extension for SG Pressure Retaining Welds and Full-Penetration Welded Nozzles (EPID: L-2020-LLR-0097), dated February 3, 2021, ADAMS Accession No. ML21034A576. 9.18 Letter from G. T. Bischof (Dominion Energy Nuclear Connecticut, Inc.), Dominion Energy Nuclear Connecticut, Inc., Millstone Power Station Unit 2 - Response to Request for Additional Information for Alternative Request RR-05 Inspection Interval Extension for Steam Generator Pressure Retaining Welds and Full-Penetration Welded Nozzles, dated March 19, 2021, ADAMS Accession No. ML21081A136. 9.19 USNRC Regulatory Guide 1.245, Revision 0, Preparing Probabilistic Fracture Mechanics Submittals, January 2022. 9.20 USNRC Report NUREG/CR-7278, Technical Basis for the use of Probabilistic Fracture Mechanics in Regulatory Applications, January 2022.

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Page 1 of 14 ATTACHMENT 1 PLANT-SPECIFIC APPLICABILITY CNS1

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Page 2 of 14 Section 9 of References [1-1] and [1-2] provide requirements that must be demonstrated in order to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for CNS1 is provided in Table 1-1 and indicates that all plant-specific requirements are met. Therefore, the results and conclusions of the EPRI reports are applicable to CNS1. Table 1-1 Applicability of References [1-1] and [1-2] Representative Analyses to CNS1 Items No. B2.40 (SG Primary Side Shell Welds) Category Requirement from Reference [1-1] Applicability to CNS1 General Requirements The Loss of Power transient (involving auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of portion of the vessel) is not considered in this evaluation due to its rarity. In the event that such a significant thermal event occurs at a plant, its impact on the KIC (material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this reports guidance. For the replacement SGs that were installed in 1996 and are currently in service, CNS1 has not experienced a loss of power transient resulting in unheated auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of any portion of the vessel. The materials of the SG vessel heads and tubesheet must be low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. The CNS1 SG vessel heads and tubesheet are fabricated of SA-508, Class 3 material (References [1-3] and [1-10]). The RTNDT values for the CNS1 SG vessel head and tubesheet materials are 0°F or less (Reference [1-5]) (so the RTNDT of 60°F used in the EPRI report is bounding). This material is a low alloy ferritic steel which conforms to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. Specific Requirements The weld configurations must conform to those shown in Figures 1-1 and Figure 1-2 of Reference [1-1]. The CNS1 tubesheet-to-head weld configuration is shown in Figure 1-2 and shows conformance with Figure 1-2 of Reference [1-1].

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Page 3 of 14 Category Requirement from Reference [1-1] Applicability to CNS1 The SG vessel dimensions must be within 10% of the upper and lower bounds of the values provided in the table in Section 9.4.3 of Reference [1-1]. The CNS1 SG vessel dimensions are as follows (Reference [1-3 and 1-4]): SG Lower Head diameter = 136.88" (OD) SG Upper Shell diameter OD = 176.26 (OD) The dimensions are within 10% of those specified in Table 9-2 in Section 9.4.3 of Reference [1-1] for Westinghouse plants. The component must experience transients and cycles bounded by those shown in Table 5-7 of Reference [1-1] over a 60-year operating life. As shown in Table 1-2, there are slight variations on some temperature and pressure values between CNS1 and Table 5-7 of Reference [1-1]. However, the CNS1 number of cycles projected to occur over a 60-year operating life are significantly lower than those shown in Table 5-7 of Reference [1-1]. See the Stress Analysis discussion in Section 5 for a basis to the applicability of Reference [1-1] for transients having a maximum pressure/temperature higher than what is listed in Reference [1-1] or a minimum pressure/temperature lower than what is listed in Reference [1-1]. Items No. C1.20 and C1.30 (SG Secondary Side Shell Welds) Category Requirement from Reference [1-1] Applicability to CNS1 General Requirements The Loss of Power transient (involving auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of portion of the vessel) is not considered in this evaluation due to its rarity. In the event that such a significant thermal event occurs at a plant, its impact on the KIC (material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this reports guidance. For the replacement SGs that were installed in 1996 and are currently in service, CNS1 has not experienced a loss of power transient resulting in unheated auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of any portion of the vessel.

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Page 4 of 14 Category Requirement from Reference [1-1] Applicability to CNS1 The materials of the SG vessel shell and tubesheet must be low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. The CNS1 SG vessel shell is fabricated from SA-533 Type B Class 1 material and the tubesheet is fabricated of SA-508, Class 3 material (References [1-3] and [1-10]). The RTNDT values for the CNS1 SG vessel shell and tubesheet material are 0°F or less (Reference [1-5]) (so the RTNDT of 60°F used in the EPRI report is bounding). These materials are low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. Specific Requirements The weld configurations must conform to those shown in Figure 1-7 and Figure 1-8 of Reference [1-1]. The CNS1 weld configurations are shown in Figures 1-3a and 1-3b, and conform to Figure 1-7 and Figure 1-8 of Reference [1-1]. The SG vessel dimensions must be within 10% of the upper and lower bounds of the values provided in the table in Section 9.4.4 of Reference [1-1]. The CNS1 SG vessel dimensions are as follows (Reference [1-3 and 1-4]): SG Lower Head diameter = 136.88" (OD) SG Upper Shell diameter = 176.26 (OD) The dimensions are within 10% of those specified in Table 9-3 in Section 9.4.4 of Reference [1-1] for Westinghouse plants. The component must experience transients and cycles bounded by those shown in Table 5-9 of Reference [1-1] over a 60-year operating life. As shown in Table 1-3, there are slight variations on some temperature and pressure values between CNS1 and Table 5-9 of Reference [1-1]. However, the CNS1 number of cycles projected to occur over a 60-year operating life are significantly lower than those shown in Table 5-9 of Reference [1-1]. See the Stress Analysis discussion in Section 5 for a basis to the applicability of Reference [1-1] for transients having a maximum pressure/temperature higher than what is listed in Reference

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Page 5 of 14 Category Requirement from Reference [1-1] Applicability to CNS1 [1-1] or a minimum pressure/temperature lower than what is listed in Reference [1-1]. Items Nos. C2.21 and C2.22 (MS and FW Nozzle to Shell Welds and Inside Radius Sections) Category Requirement from Reference [1-2] Applicability to CNS1 General Requirements The nozzle-to-shell weld shall be one of the configurations shown in Figure 1-1 or Figure 1-2 of Reference [1-2]. The CNS1 FW nozzle-to-shell weld in shown in Figure 1-4 and is representative of the configuration shown in Figure 1-2 of Reference [1-2]. The CNS1 Main Steam nozzle is forged and does not have a C2.21 or C2.22 component. The materials of the SG shell, FW nozzles, and MS nozzles must be low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. The CNS1 SG shell is fabricated from SA-533 Type B Class 1 material and the FW nozzles are fabricated of SA-508, Class 3 material (Reference [1-3]). The RTNDT values for the CNS1 SG shell and FW nozzle materials are 0°F or less (Reference [1-5]) (so the RTNDT of 60°F used in the EPRI report is bounding). This material is a low alloy ferritic steel which conforms to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. Per above, the CNS1 Main Steam nozzle is forged and does not have a C2.21 or C2.22 component. The SG must not experience more than the number of all transients shown in Table 5-5 of Reference [1-2] over a 60-year operating life. As shown in Table 1-4, the CNS1 SGs are not projected to experience more than the number of transients shown in Table 5-5 of Reference [1-2] over a 60-year operating life.

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Page 6 of 14 Category Requirement from Reference [1-2] Applicability to CNS1 SG Feedwater Nozzle The piping attached to the FW nozzle must be 14-inch to 18-inch NPS. The piping attached to the CNS1 FW nozzle is 16-inch NPS Reference [1-12]. The FW nozzle design must have an integrally attached thermal sleeve. The CNS1 FW nozzle configuration is shown in Figure 1-4 and has an integrally attached thermal sleeve shown in Reference [1-11]. Auxiliary feedwater nozzles connected directly to the SG are not covered in this evaluation. N/A for CNS1. SG Main Steam Nozzle For Westinghouse and CE SGs, the piping attached to the SG main steam nozzle must be 28-inch to 36-inch NPS. N/A for CNS1 (the CNS1 Main Steam nozzle is forged and does not have a C2.21 or C2.22 component). For B&W SGs, the piping attached to the main steam nozzle must be 22-inch to 26-inch NPS. This requirement is not applicable for CNS1 because it is a Westinghouse 4-loop PWR. The SG must have one main steam nozzle that exits the top dome of the SG. For B&W plants, there may be more than one main steam nozzle; it will exit the side of the SG. N/A for CNS1 (the CNS1 Main Steam nozzle is forged and does not have a C2.21 or C2.22 component). The main steam nozzle shall not significantly protrude into the SG (e.g., see Figure 4-7 of Reference 1-2 or have a unique nozzle weld configuration (e.g., see Figure 4-6 of Reference [1-2]). N/A for CNS1 (the CNS1 Main Steam nozzle is forged and does not have a C2.21 or C2.22 component).

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Page 7 of 14 Table 1-2 CNS1 Data for Thermal Transients for Stress Analysis of the PWR SG Primary-Side Head Welds (Comparison to Table 5-7 of Reference [1-1]) Transient Max Thot °F Min Thot °F Max Tcold °F Min Tcold °F Max Press PSIG Min Press PSIG 60-Year Projected Cycles Heatup/Cooldown EPRI Report 3002015906 545 70 545 70 2235 0 300 Heatup/Cooldown CNS1(1)(5) 559 93 558 80 2238 0 167 Plant Loading / Unloading EPRI Report 3002015906 610 550 550 545 2300 2300 5000 Plant Loading / Unloading CNS1(2)(5) Note 4 Note 4 Note 4 Note 4 Note 4 Note 4 52 Reactor Trip EPRI Report 3002015906 615 530 565 530 2435 1700 360 Reactor Trip CNS1(3)(5) 616 556 562 553 2272 2031 58 Notes: 1. Heatup/Cooldown = RCS Heatup and RCS Cooldown from Table 45 of Reference [1-7]. 2. Plant Loading/Unloading = Loading 0% - 15% (S/G A etc.) from Table 45 of Reference [1-7]. 3. Reactor Trip = Reactor Trip (large & small deltaP) from Table 45 of Reference [1-7]. 4. Transient not monitored at CNS1. 5. Temperature and pressure values obtained from supporting databases associated with the subject references.

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Page 8 of 14 Table 1-3 CNS1 Data for Thermal Transients for Stress Analysis of the PWR SG Secondary-Side Vessel Welds (Comparison to Table 5-9 of Reference [1-1]) Transient Max Tss °F Min Tss °F Max Press PSIG Min Press PSIG 60-Year Projected Cycles Heatup/Cooldown EPRI Report 3002015906 545 70 1000 0 300 Heatup/Cooldown CNS1(1)(5) 557 89 1089 0 167 Plant Loading / Unloading EPRI Report 3002015906 545 540 1000 1000 5000 Plant Loading / Unloading CNS1(2)(5) Note 4 Note 4 Note 4 Note 4 52 Reactor Trip EPRI Report 3002015906 555 530 1130 1000 360 Reactor Trip CNS1(3)(5) 560 543 1121 975 58 Notes:

1. Heatup/Cooldown = RCS Heatup and RCS Cooldown from Table 45 of Reference [1-7].

2. Plant Loading/Unloading = Loading 0% - 15% (S/G A etc.) from Table 45 of Reference

[1-7].

3. Reactor Trip = Reactor Trip (large & small deltaP) from Table 45 of Reference [1-7].

4. Transient not monitored at CNS1.

5. Temperature and pressure values obtained from supporting databases associated with the subject references.

- CNS1 Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 9 of 14 Table 1-4 CNS1 Data for Thermal Transients Applicable to PWR SG Feedwater and Main Steam Nozzles (Comparison to Table 5-5 of Reference [1-2]) Transient 60-Year Allowable Cycles from Table 5-5 of EPRI Report 3002014590 [1-2] 60-Year Projected Cycles CNS1 Heatup/Cooldown(1) 300 167 Plant Loading(2) 5000 52 Plant Unloading(2) 5000 52 Loss of Load(3) 360 58 Loss of Power(4) 60 12 Notes:

1. Heatup/Cooldown = RCS Heatup and RCS Cooldown from Table 45 of Reference [1-7].

2. Plant Loading/Unloading = Loading 0% - 15% (S/G A etc.) from Table 45 of Reference

[1-7].

3. Loss of Load = Reactor Trip (large & small deltaP) from Table 45 of Reference [1-7].

4. Loss of Power = Loss/Pwr/Blackout+NatCirc from Table 45 of Reference [1-7].

Table 1-5. CNS1 Inspection History SG Primary Side Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam B2.40 1SGB-W22 5/29/2008 3rd/1st/C1R17 RSG Acceptable 99.80% N/A No B2.40 1SGB-W22 5/9/2017 4th/1st/C1R23 RSG Acceptable 99.96% N/A No

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Page 10 of 14 SG Secondary Side Shell Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam C1.20 1SGD-W144 5/23/2005 2nd/3rd/C1R15 RSG Acceptable 100% N/A No C1.20 1SGD-W144 5/6/2011 3rd/2nd/C1R19 RSG Acceptable 100% N/A No C1.30 1SGA-W65 5/29/2008 3rd/1st/C1R17 RSG Acceptable 96.70% N/A No C1.30 1SGA-W65 5/9/2017 4th/1st/C1R23 RSG Acceptable 100% N/A No SG Secondary Side Nozzle Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam C2.21 1SGD-W258 12/20/1997 2nd/1st/C1R10 RSG Acceptable 100% N/A No C2.21 1SGD-W258 5/12/2011 3rd/2nd/C1R19 RSG Acceptable 100% N/A No C2.22 1SGD-W258 5/12/2011 3rd/2nd/C1R19 RSG Acceptable 100% N/A No C2.22 1SGC-W258 11/25/2018 4th/2nd/C1R24 RSG Acceptable 100% N/A No

- CNS1 Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 11 of 14 Figure 1-1. CNS1 Steam Generator Layout [1-8] W7l \\~~ \\~~ ~ 11i '~~ \\ trn ~1 >170

• W23 l &. '.,;32

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Page 12 of 14 Figure 1-2. CNS1 Item No. B2.40 Weld Configuration [1-8] Figure 1-3a. CNS1 Item No. C1.20 Weld Configuration [1-8]

- CNS1 Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 13 of 14 Figure 1-3b. CNS1 Item No. C1.30 Weld Configuration [1-8] Figure 1-4. CNS1 Feedwater Nozzle Configuration [1-6] \\ \\ . \\\\\\ ' \\ -..--.. IV w-UlCD 0

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Page 14 of 14 References 1-1. Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906. 1-2. Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Inside Radius Sections. EPRI, Palo Alto, CA: 2019. 3002014590. 1-3. Drawing MCM 1201.01-0684.001, MNS1 Steam Generator Arrangement, Revision D5 (Note: CNS1 SG identical to MNS1/MNS2 SGs). 1-4. Drawing CNM 1201.01-0609 001, Primary Head and TS DL, Revision D9. 1-5. Chapter 5 CNS1 UFSAR Section 5.2.3.3.1. 1-6. Drawing CNM 1201.01-0567 001, Main Feedwater Nozzle, Revision 04. 1-7. CNC 1206.02-45-0031, SI Calculation FP-CNS-308 - Catawba SI:FatiguePro 4 Baseline Analysis Startup through 5/2/2020 (U1) and 3/30/2021 (U2), Revision 0. 1-8. Drawing CNM 1201.01-0617 001, Layout of Vessel Ref. Points for Welds, Revision 08. 1-9. Not used. 1-10. Drawing MCM 2201.01-0126.001, MNS Steam Generator Arrangement, Revision D8 (also applicable to CNS1). 1-11. Drawing CNM-1201.01-0581 001, Main Feedwater Nozzle Assy, Revision 04. 1-12. Drawing CN-1591-1.1, Flow Diagram of Feedwater System (CF), Revision 040.

- CNS2 Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 1 of 17 ATTACHMENT 2 PLANT-SPECIFIC APPLICABILITY CNS2

- CNS2 Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 2 of 17 Section 9 of References [2-1] and [2-2] provide requirements that must be demonstrated in order to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for CNS2 is provided in Table 2-1 and indicates that all plant-specific requirements are met. Therefore, the results and conclusions of the EPRI reports are applicable to CNS2. Table 2-1 Applicability of References [2-1] and [2-2] Representative Analyses to CNS2 Items No. B2.40 (SG Primary Side Shell Welds) Category Requirement from Reference [2-1] Applicability to CNS2 General Requirements The Loss of Power transient (involving auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of portion of the vessel) is not considered in this evaluation due to its rarity. In the event that such a significant thermal event occurs at a plant, its impact on the KIC (material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this reports guidance. For the original SGs that are currently in service, CNS2 has not experienced a loss of power transient resulting in unheated auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of any portion of the vessel. The materials of the SG vessel heads and tubesheet must be low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. The CNS2 SG vessel heads are fabricated of SA-533 Grade A, Class 2 material and the tubesheet is fabricated of SA-508, Class 2a material (Reference [2-3]). The RTNDT values for the CNS2 SG vessel heads and tubesheet materials are 60°F or less (Reference [2-3]) (so the RTNDT of 60°F used in the EPRI report is bounding). This material is a low alloy ferritic steel which conforms to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. Specific Requirements The weld configurations must conform to those shown in Figures 1-1 and Figure 1-2 of Reference [2-1]. The CNS2 tubesheet-to-shell weld configuration is shown in Figure 2-2 and conforms to Figure 1-2 of Reference [2-1].

- CNS2 Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 3 of 17 Category Requirement from Reference [2-1] Applicability to CNS2 The SG vessel dimensions must be within 10% of the upper and lower bounds of the values provided in the table in Section 9.4.3 of Reference [2-1]. The CNS2 SG vessel dimensions are as follows (Reference [2-4]): SG Lower Head diameter = 140.5" (OD) SG Upper Shell diameter = 176.26 (OD) The dimensions are within 10% of those specified in Table 9-2 in Section 9.4.3 of Reference [2-1] for Westinghouse plants. The component must experience transients and cycles bounded by those shown in Table 5-7 of Reference [2-1] over a 60-year operating life. As shown in Table 2-2, there are slight variations on some temperature and pressure values between CNS2 and Table 5-7 of Reference [2-1]. However, the CNS2 number of cycles projected to occur over a 60-year operating life are significantly lower than those shown in Table 5-7 of Reference [2-1]. See the Stress Analysis discussion in Section 5 for a basis to the applicability of Reference [2-1] for transients having a maximum pressure/temperature higher than what is listed in Reference [2-1] or a minimum pressure/temperature lower than what is listed in Reference [2-1]. Items No. C1.10, C1.20, and C1.30 (SG Secondary Side Shell Welds) Category Requirement from Reference [2-1] Applicability to CNS2 General Requirements The Loss of Power transient (involving auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of portion of the vessel) is not considered in this evaluation due to its rarity. In the event that such a significant thermal event occurs at a plant, its impact on the KIC (material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this reports guidance. For the original SGs are currently in service, CNS2 has not experienced a loss of power transient resulting in unheated auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of any portion of the vessel.

- CNS2 Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 4 of 17 Category Requirement from Reference [2-1] Applicability to CNS2 The materials of the SG vessel shell and tubesheet must be low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. The CNS2 SG vessel shell is fabricated of SA-533, Grade A, Class 2 material and the tubesheet is fabricated of SA-508, Class 2a material (Reference [2-5]). The RTNDT values for the CNS2 SG vessel shell and tubesheet materials is 60°F or less (Reference [2-3]) (so the RTNDT of 60°F used in the EPRI report is bounding). These materials are low alloy ferritic steel which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. Specific Requirements The weld configurations must conform to those shown in Figure 1-7 and Figure 1-8 of Reference [2-1]. The CNS2 weld configurations are shown in Figures 2-3a and 2-3b, and conforms to Figure 1-7 and Figure 1-8 of Reference [2-1]. The SG vessel dimensions must be within 10% of the upper and lower bounds of the values provided in the table in Section 9.4.4 of Reference [2-1]. The CNS2 SG vessel dimensions are as follows (Reference [2-3]): SG Lower Head diameter = 140.5" (OD) SG Upper Shell diameter = 176.26 (OD) The dimensions are within 10% of those specified in Table 9-3 in Section 9.4.4 of Reference [2-1] for Westinghouse plants. The component must experience transients and cycles bounded by those shown in Table 5-9 of Reference [2-1] over a 60-year operating life. As shown in Table 2-3, there are slight variations on some temperature and pressure values between CNS2 and Table 5-9 of Reference [2-1]. However, the CNS2 number of cycles projected to occur over a 60-year operating life are significantly lower than those shown in Table 5-9 of Reference [2-1]. See the Stress Analysis discussion in Section 5 for a basis to the applicability of Reference [2-1] for transients having a maximum pressure/temperature higher than what is listed in Reference [2-1] or a minimum

- CNS2 Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 5 of 17 Category Requirement from Reference [2-1] Applicability to CNS2 pressure/temperature lower than what is listed in Reference [2-1]. Items Nos. C2.21 and C2.22 (MS and FW Nozzle to Shell Welds and Inside Radius Sections) Category Requirement from Reference [2-2] Applicability to CNS2 General Requirements The nozzle-to-shell weld shall be one of the configurations shown in Figure 1-1 or Figure 1-2 of Reference [2-2]. The CNS2 FW nozzle-to-shell weld in shown in Figure 2-4 and is representative of the configuration shown in Figure 1-2 of Reference [2-2]. The CNS2 MS nozzle-to-shell weld in shown in Figure 2-5 and is representative of the configuration shown in Figure 1-2 of Reference [2-2]. The materials of the SG shell, FW nozzles, and MS nozzles must be low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. The CNS2 SG shell and heads are fabricated of SA-533, Gr. A Class 2 material (Reference [2-3]). The CNS2 FW and MS nozzles are fabricated of SA-508 Class 2a. The RTNDT value for the material of CNS2 SG shell and nozzle materials is 60°F or less (Reference [2-3]) (so the RTNDT of 60°F used in the EPRI report is bounding). This material is a low alloy ferritic steel which conforms to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. The SG must not experience more than the number of all transients shown in Table 5-5 of Reference [2-2] over a 60-year operating life. As shown in Table 2-4, the CNS2 SGs are not projected to experience more than the number of transients shown in Table 5-5 of Reference [2-2] over a 60-year operating life. SG Feedwater Nozzle The piping attached to the FW nozzle must be 14-inch to 18-inch NPS. The piping attached to the CNS2 FW nozzle is 16-inch NPS (Reference [2-6]).

- CNS2 Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 6 of 17 Category Requirement from Reference [2-2] Applicability to CNS2 The FW nozzle design must have an integrally attached thermal sleeve. The CNS2 FW nozzle configuration is shown in Figure 2-4 and has an integrally attached thermal sleeve (Reference [2-6]). Auxiliary feedwater nozzles connected directly to the SG are not covered in this evaluation. N/A for CNS2. SG Main Steam Nozzle For Westinghouse and CE SGs, the piping attached to the SG main steam nozzle must be 28-inch to 36-inch NPS. CNS2 is a Westinghouse 4-Loop PWR. The piping attached to the CNS2 SG main steam nozzle is 32-inch NPS (Reference [2-9]). For B&W SGs, the piping attached to the main steam nozzle must be 22-inch to 26-inch NPS. This requirement is not applicable for CNS2 because it is a Westinghouse 4-loop PWR. The SG must have one main steam nozzle that exits the top dome of the SG. For B&W plants, there may be more than one main steam nozzle; it will exit the side of the SG. As shown in Figure 2-5, CNS2 has one MS nozzle per SG that exits the top dome of the SG. The main steam nozzle shall not significantly protrude into the SG (e.g., see Figure 4-7 of Reference [2-2] or have a unique nozzle weld configuration (e.g., see Figure 4-6 of Reference [2-2]). The CNS2 MS nozzle configuration (shown in Figure 2-5) does not protrude significantly into the SG as shown in Figure 4-7 of Reference [2-2] and does not have a unique weld configuration as shown in Figure 4-6 of Reference [2-2].

- CNS2 Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 7 of 17 Table 2-2 CNS2 Data for Thermal Transients for Stress Analysis of the PWR SG Primary-Side Head Welds (Comparison to Table 5-7 of Reference [2-1]) Transient Max Thot °F Min Thot °F Max Tcold °F Min Tcold °F Max Press PSIG Min Press PSIG 60-Year Projected Cycles Heatup/Cooldown EPRI Report 3002015906 545 70 545 70 2235 0 300 Heatup/Cooldown CNS2(1)(5) 561 94 558 90 2257 30 75/79 Plant Loading / Unloading EPRI Report 3002015906 610 550 550 545 2300 2300 5000 Plant Loading / Unloading CNS2(2)(5) Note 4 Note 4 Note 4 Note 4 Note 4 Note 4 60 Reactor Trip EPRI Report 3002015906 615 530 565 530 2435 1700 360 Reactor Trip CNS2(3)(5) 617 540 562 502 2234 1940 67 Notes:

1. Heatup/Cooldown = RCS Heatup and RCS Cooldown from Table 49 of Reference [2-8].

2. Plant Loading/Unloading = Loading 0% - 15% (S/G A etc.) from Table 49 of Reference

[2-8].

3. Reactor Trip = Reactor Trip (large & small deltaP) from Table 49 of Reference [2-8].

4. Transient not monitored at CNS2.

5. Temperature and pressure values obtained from supporting databases associated with the subject references.

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Page 8 of 17 Table 2-3 CNS2 Data for Thermal Transients for Stress Analysis of the PWR SG Secondary-Side Vessel Welds [Ref(s)] (Comparison to Table 5-9 of Reference [2-1]) Transient Max Tss °F Min Tss °F Max Press PSIG Min Press PSIG 60-Year Projected Cycles Heatup/Cooldown EPRI Report 3002015906 545 70 1000 0 300 Heatup/Cooldown CNS2(1)(5) 556 95 1085 0 75/79 Plant Loading / Unloading EPRI Report 3002015906 545 540 1000 1000 5000 Plant Loading / Unloading CNS2(2)(5) Note 4 Note 4 Note 4 Note 4 60 Reactor Trip EPRI Report 3002015906 555 530 1130 1000 360 Reactor Trip CNS2(3)(5) 562 503 1139 686 67 Notes:

1. Heatup/Cooldown = RCS Heatup and RCS Cooldown from Table 49 of Reference [2-8].

2. Plant Loading/Unloading = Loading 0% - 15% (S/G A etc.) from Table 49 of Reference

[2-8].

3. Reactor Trip = Reactor Trip (large & small deltaP) from Table 49 of Reference [2-8].

4. Transient not monitored at CNS2.

5. Temperature and pressure values obtained from supporting databases associated with the subject references.

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Page 9 of 17 Table 2-4 CNS2 Data for Thermal Transients Applicable to PWR SG Feedwater and Main Steam Nozzles (Comparison to Table 5-5 of Reference [2-2]) Transient 60-Year Allowable Cycles from Table 5-5 of EPRI Report 3002014590 [2-2] 60-Year Projected Cycles CNS2 Heatup/Cooldown(1) 300 75/79 Plant Loading(2) 5000 60 Plant Unloading(2) 5000 60 Loss of Load(3) 360 67 Loss of Power(4) 60 4 Notes:

1. Heatup/Cooldown = RCS Heatup and RCS Cooldown from Table 49 of Reference [2-8].

2. Plant Loading/Unloading = Loading 0% - 15% (S/G A etc.) from Table 49 of Reference

[2-8].

3. Loss of Load = Reactor Trip (large & small deltaP) from Table 49 of Reference [2-8].

4. Loss of Power = Loss/Pwr/Blackout+NatCirc from Table 49 of Reference [2-8].

Table 2-5. CNS2 Inspection History SG Primary Side Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam B2.40 2SGA-01-02 9/22/2013 3rd/3rd/C2R19 OSG Acceptable 98.33% N/A No B2.40 2SGA-01-02 4/19/2021 4th/2nd/C2R24 OSG Acceptable 100% N/A Yes

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Page 10 of 17 SG Secondary Side Shell Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam C1.10 2SGB-03-04A 3/27/2009 3rd/2nd/C2R16 OSG Acceptable 93.12% N/A No C1.10 2SGB-03-04A 3/25/2018 4th/1st/C2R22 OSG Acceptable 100% N/A No C1.10 2SGC-04B-05 9/26/2004 2nd/3rd/C2R13 OSG Acceptable 48.30% 05-CN-003* No C1.10 2SGC-04B-05 3/27/2009 3rd/2nd/C2R16 OSG Acceptable 46.872% 11-CN-001** No C1.10 2SGD-05-06A 9/27/2001 2nd/2nd/C2R11 OSG Acceptable 100% N/A No C1.10 2SGD-05-06A 3/25/2009 3rd/2nd/C2R16 OSG Acceptable 100% N/A No C1.20 2SGD-06B-07 3/25/2009 3rd/2nd/C2R16 OSG Acceptable 100% N/A No C1.20 2SGD-06B-07 9/25/2022 4th/2nd/C2R25 OSG Acceptable 100% N/A Yes C1.30 2SGA-02-03 4/3/2009 3rd/2nd/C2R16 OSG Acceptable 100% N/A No C1.30 2SGA-02-03 3/23/2018 4th/1st/C2R22 OSG Acceptable 91.30% N/A No

• NRC SER via ADAMS Accession Number ML053550370
  • NRC SER via ADAMS Accession Number ML12228A723 SG Secondary Side Nozzle Welds Item No.

Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam C2.21 2SGB-SB-11 3/17/2015 3rd/3rd/C2R20 OSG Acceptable 100% N/A No C2.21 2SGB-SB-11 3/25/2018 4th/1st/C2R22 OSG Acceptable 100% N/A No C2.21 2SGD-UH-15 3/21/2006 2nd/3rd/C2R14a OSG Acceptable 100% N/A No C2.21 2SGD-UH-15 3/10/2015 3rd/3rd/C2R20 OSG Acceptable 100% N/A No C2.22 2SGB-SB-11 9/23/2004 2nd/3rd/C2R13 OSG Acceptable 100% N/A No C2.22 2SGB-SB-11 3/17/2015 3rd/3rd/C2R20 OSG Acceptable 100% N/A No

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Page 11 of 17 J.68 \\ ./lct;,tl'-. I ,r \\ I ~ 6J.e4 I l.88 }i-"* "' '*168.!:0 ,. I

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,.<z I 06 ~ ';,.~-. 3.10 '1'03 31.00

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Page 12 of 17 Figure 2-1. CNS2 Steam Generator Layout [2-3] Figure 2-2. CNS2 Item No. B2.40 Weld Configuration [2-5] .!SOR. [12."f) r-- 7* I *. n.... 1 L---------.-f \\ .12(3.0] R. \\

- CNS2 Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 13 of 17 Transition Cone to Upper Shell (C1.10) ~ \\ .1 \\

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Page 14 of 17 Lower Shell to Transition Cone (C1.10) Stub Barrel to Lower Shell (C1.10) Shell to Head (C1.20) Figure 2-3a. CNS2 Item Nos. C1.10 & C1.20 Weld Configurations [2-5] --j /--- ;I'm>.> IJ I t] Jf ( . J J,.-.75 (19.1) t ,.o. I

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Page 15 of 17 Figure 2-3b. CNS2 Item No. C1.30 Weld Configuration [2-5] . -l L-.1s (19.IJ 1 DETAIL *o* SEE NOTE 4 \\ 2.00 ~-81 I.D. I \\OE.TAI!-*E 11 ~ \\.;,r~*~L t L

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Page 16 of 17 Figure 2-4. CNS2 Feedwater Nozzle Configuration [2-6] Figure 2-5. CNS2 Main Steam Nozzle Configuration [2-7] .oe ] ~1-so r L w r::=. 0 ,,,Jt§L 3.19 UIN. SHEL\\. TUICKN(SS 0£ lf,ll "!( 26.00 OtA. SEE

1. 81 I

4 WRAPPER BARREL )l--. -:_:_:_:_:_:_:_:_:_:_:_-71-,0-0-R!~--'= '---- --.'--'--'----11 .50 6~.69 TO (., Of' VESSEL '--LI-_-_-_-_-:_-:_-_---- -- -- U4.Z ~ ~, ___ _:::,,-.;;::=:==----=::---,.-i

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Page 17 of 17 References 2-1. Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906. 2-2. Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Inside Radius Sections. EPRI, Palo Alto, CA: 2019. 3002014590. 2-3. Drawing CNM 2201.01-0217.001, Model D5 Steam Generator Stress Report Unit 2, Revision D4. 2-4. Drawing CNM 2201.01-0059.001, Vertical Steam Generator Outline Model D-5, Revision D12. 2-5. Drawing CNM 2201.01-0102, Steam Generator Model D5-3 General Arrangement, Revision 0. 2-6. Drawing EDSK 3802178, Feedwater Nozzle (SCH 80), Revision 0. 2-7. Drawing EDSK 3802168, Stm. Outlet Noz. & Elliptical Head, Revision 0. 2-8. CNC 1206.02-45-0031, SI Calculation FP-CNS-308 - Catawba SI:FatiguePro 4 Baseline Analysis Startup through 5/2/2020 (U1) and 3/30/2021 (U2), Revision 0. 2-9. Drawing CN-2591-01.01, Flow Diagram Feedwater System (CF), Revision 49.

- MNS1/2 Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 1 of 17 ATTACHMENT 3 PLANT-SPECIFIC APPLICABILITY MNS1/2

- MNS1/2 Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 2 of 17 Section 9 of References [3-1] and [3-2] provide requirements that must be demonstrated in order to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for MNS1/2 is provided in Table 3-1 and indicates that all plant-specific requirements are met. Therefore, the results and conclusions of the EPRI reports are applicable to MNS1/2. Table 3-1 Applicability of References [3-1] and [3-2] Representative Analyses to MNS1/2 Items No. B2.40 (SG Primary Side Shell Welds) Category Requirement from Reference [3-1] Applicability to MNS1/2 General Requirements The Loss of Power transient (involving auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of portion of the vessel) is not considered in this evaluation due to its rarity. In the event that such a significant thermal event occurs at a plant, its impact on the KIC (material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this reports guidance. For the replacement SGs that were installed in 1997 and are currently in service, MNS1/2 have not experienced a loss of power transient resulting in unheated auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of any portion of the vessel. The materials of the SG vessel heads and tubesheet must be low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. The MNS1/2 SG vessel heads and tubesheet are fabricated of SA-508, Class 3 material (References [3-3] and [3-4]). The RTNDT values for the SG vessel head and tubesheet materials are 0°F or less (Reference [3-11]) (so the RTNDT of 60°F used in the EPRI report is bounding). This material is a low alloy ferritic steel which conforms to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. Specific Requirements The weld configurations must conform to those shown in Figures 1-1 and Figure 1-2 of Reference [3-1]. The MNS1/2 tubesheet-to-shell weld configuration is shown in Figure 3-2 and conforms to Figure 1-2 of Reference [3-1].

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Page 3 of 17 Category Requirement from Reference [3-1] Applicability to MNS1/2 The SG vessel dimensions must be within 10% of the upper and lower bounds of the values provided in the table in Section 9.4.3 of Reference [3-1]. The MNS1/2 SG vessel dimensions are as follows (References [3-3] and [3-4]): SG Lower Head diameter = 136.88" (OD) SG Upper Shell diameter = 176.26 (OD) The dimensions are within 10% of those specified in Table 9-2 in Section 9.4.3 of Reference [3-1] for Westinghouse plants. The component must experience transients and cycles bounded by those shown in Table 5-7 of Reference [3-1] over a 60-year operating life. As shown in Table 3-2, there are slight variations on some temperature and pressure values between MNS1/2 and Table 5-7 of Reference [3-1]. However, the MNS1/2 number of cycles projected to occur over a 60-year operating life are significantly lower than those shown in Table 5-7 of Reference [3-1]. See the Stress Analysis discussion in Section 5 for a basis to the applicability of Reference [3-1] for transients having a maximum pressure/temperature higher than what is listed in Reference [3-1] or a minimum pressure/temperature lower than what is listed in Reference [3-1]. Items No. C1.20 and C1.30 (SG Secondary Side Shell Welds) Category Requirement from Reference [3-1] Applicability to MNS1/2 General Requirements The Loss of Power transient (involving auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of portion of the vessel) is not considered in this evaluation due to its rarity. In the event that such a significant thermal event occurs at a plant, its impact on the KIC (material fracture toughness) value may require more For the replacement SGs that were installed in 1997 and are currently in service, MNS1/2 have not experienced a loss of power transient resulting in unheated auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of any portion of the vessel.

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Page 4 of 17 Category Requirement from Reference [3-1] Applicability to MNS1/2 frequent examinations and other plant actions outside the scope of this reports guidance. The materials of the SG vessel shell and tubesheet must be low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. The MNS1/2 SG vessel shell is fabricated from SA-533, Type B, Class 1 material and the tubesheet is fabricated of SA-508, Class 3 material (References [3-3] and [3-4]). The RTNDT values for the SG vessel shell and tubesheet materials are 0°F or less (Reference [3-11]) (so the RTNDT of 60°F used in the EPRI report is bounding). These materials are low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. Specific Requirements The weld configurations must conform to those shown in Figure 1-7 and Figure 1-8 of Reference [3-1]. The MNS1/2 weld configurations are shown in Figures 3-3a and 3-3b, and conforms to Figure 1-7and Figure 1-8 of Reference [3-1]. The SG vessel dimensions must be within 10% of the upper and lower bounds of the values provided in the table in Section 9.4.4 of Reference [3-1]. The MNS1/2 SG vessel dimensions are as follows (References [3-3] and [3-4]): SG Lower Head diameter = 136.88" (OD) SG Upper Shell diameter = 176.26 (OD) The dimensions are within 10% of those specified in Table 9-3 in Section 9.4.4 of Reference [3-1] for Westinghouse plants. The component must experience transients and cycles bounded by those shown in Table 5-9 of Reference [3-1] over a 60-year operating life. As shown in Table 3-3, there are slight variations on some temperature and pressure values between MNS1/2 and Table 5-9 of Reference [3-1]. However, the MNS1/2 number of cycles projected to occur over a 60-year operating life are significantly lower than those shown in Table 5-9 of Reference [3-1]. See the Stress

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Page 5 of 17 Category Requirement from Reference [3-1] Applicability to MNS1/2 Analysis discussion in Section 5 for a basis to the applicability of Reference [3-1] for transients having a maximum pressure/temperature higher than what is listed in Reference [3-1] or a minimum pressure/temperature lower than what is listed in Reference [3-1]. Items Nos. C2.21 and C2.22 (MS and FW Nozzle to Shell Welds and Inside Radius Sections) Category Requirement from Reference [3-2] Applicability to MNS1/2 General Requirements The nozzle-to-shell weld shall be one of the configurations shown in Figure 1-1 or Figure 1-2 of Reference [3-2]. The MNS1/2 FW nozzle-to-shell weld in shown in Figure 3-4 and is representative of the configuration shown in Figure 1-2 of Reference [3-2]. The MNS1/2 MS nozzle is forged and does not have a C2.21 or C2.22 component. The materials of the SG shell, FW nozzles, and MS nozzles must be low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. The MNS1/2 SG vessel shell is fabricated of SA-533 Type B Class 1 material and the FW nozzles are fabricated of SA-508, Class 3 material (References [3-3] and [3-4]). The RTNDT values for the SG shell and FW nozzles are 0°F or less (Reference [3-11]) (so the RTNDT of 60°F used in the EPRI report is bounding). These materials are low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. As noted above, the MNS1/2 MS nozzle is forged and does not have a C2.21 or C2.22 component. The SG must not experience more than the number of all transients shown in Table 5-5 of Reference [3-2] over a 60-year operating life. As shown in Table 3-4, the MNS1/2 SGs are not projected to experience more than the number of transients

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Page 6 of 17 Category Requirement from Reference [3-2] Applicability to MNS1/2 shown in Table 5-5 of Reference [3-2] over a 60-year operating life. SG Feedwater Nozzle The piping attached to the FW nozzle must be 14-inch to 18-inch NPS. The piping attached to the MNS1/2 FW nozzle is 16-inch NPS (Reference [3-3]). The FW nozzle design must have an integrally attached thermal sleeve. The MNS1/2 FW nozzle configuration is shown in Figure 2-4 and has an integrally attached thermal sleeve shown in References [3-12] and [3-13]. Auxiliary feedwater nozzles connected directly to the SG are not covered in this evaluation. N/A for MNS1/2. SG Main Steam Nozzle For Westinghouse and CE SGs, the piping attached to the SG main steam nozzle must be 28-inch to 36-inch NPS. N/A for MNS1/2 (the MS nozzle is forged and does not have a C2.21 or C2.22 component). For B&W SGs, the piping attached to the main steam nozzle must be 22-inch to 26-inch NPS. This requirement is not applicable for MNS1/2 because they are both Westinghouse 4-loop PWRs. The SG must have one main steam nozzle that exits the top dome of the SG. For B&W plants, there may be more than one main steam nozzle; it will exit the side of the SG. N/A for MNS1/2 (the MS nozzle is forged and does not have a C2.21 or C2.22 component). The main steam nozzle shall not significantly protrude into the SG (e.g., see Figure 4-7 of Reference [3-2] or have a unique nozzle weld configuration (e.g., see Figure 4-6 of Reference [3-2]). N/A for MNS1/2 (the MS nozzle is forged and does not have a C2.21 or C2.22 component). Table 3-2 MNS1/2 Data for Thermal Transients for Stress Analysis of the PWR SG Primary-Side Head Welds (Comparison to Table 5-7 of Reference [3-1])

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Page 7 of 17 Transient Max Thot °F Min Thot °F Max Tcold °F Min Tcold °F Max Press PSIG Min Press PSIG 60-Year Projected Cycles Heatup/Cooldown EPRI Report 3002015906 545 70 545 70 2235 0 300 Heatup/Cooldown MNS1/2(1)(5) 601/566 91/91 559/566 82/85 2245/2266 2/15 92/81 Plant Loading / Unloading EPRI Report 3002015906 610 550 550 545 2300 2300 5000 Plant Loading / Unloading MNS1/2(2)(5) Note 4 Note 4 Note 4 Note 4 Note 4 Note 4 61/50 Reactor Trip EPRI Report 3002015906 615 530 565 530 2435 1700 360 Reactor Trip MNS1/2(3)(5) 613/ Note 4 558/ Note 4 558/ Note 4 554/ Note 4 2234/ Note 4 2106/ Note 4 101/79 Notes:

1. Heatup/Cooldown = RCS Heatup and RCS Cooldown from Tables 5-1 and 5-12 of [3-5].

2. Plant Loading/Unloading = Loading 0% - 15% (S/G A etc.) from Tables 5-2 and 5-13 of

[3-5].

3. Reactor Trip = Reactor Trip (large & small deltaP) from Tables 5-1 and 5-12 of [3-5].

4. Transient not monitored at MNS1/2.

5. Temperature and pressure values obtained from supporting databases associated with the subject references.

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Page 8 of 17 Table 3-3 MNS1/2 Data for Thermal Transients for Stress Analysis of the PWR SG Secondary-Side Vessel Welds [Ref(s)] (Comparison to Table 5-9 of Reference [3-1]) Transient Max Tss °F Min Tss °F Max Press PSIG Min Press PSIG 60-Year Projected Cycles Heatup/Cooldown EPRI Report 3002015906 545 70 1000 0 300 Heatup/Cooldown MNS1/2(1)(5) 557/557 88/90 1095/1091 0/0 92/81 Plant Loading / Unloading EPRI Report 3002015906 545 540 1000 1000 5000 Plant Loading / Unloading MNS1/2(2)(5) Note 4/Note 4 Note 4/Note 4 Note 4/Note 4 Note 4/Note 4 61/50 Reactor Trip EPRI Report 3002015906 555 530 1130 1000 360 Reactor Trip MNS1/2(3)(5) 558/555 545/530 1099/1130 991/1000 101/79 Notes:

1. Heatup/Cooldown = RCS Heatup and RCS Cooldown from Tables 5-1 and 5-12 of [3-5].

2. Plant Loading/Unloading = Loading 0% - 15% (S/G A etc.) from Tables 5-2 and 5-13 of

[3-5].

3. Reactor Trip = Reactor Trip (large & small deltaP) from Tables 5-1 and 5-12 of [3-5].

4. Transient not monitored at MNS1/2.

5. Temperature and pressure values obtained from supporting databases associated with the subject references.

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Page 9 of 17 Table 3-4 MNS1/2 Data for Thermal Transients Applicable to PWR SG Feedwater and Main Steam Nozzles [Ref(s)] (Comparison to Table 5-5 of Reference [3-2]) Transient 60-Year Allowable Cycles from Table 5-5 of EPRI Report 3002014590 [3-2] 60-Year Projected Cycles MNS1/2 Heatup/Cooldown(1) 300 92/81 Plant Loading(2) 5000 61/50 Plant Unloading(2) 5000 61/50 Loss of Load(3) 360 101/79 Loss of Power(4) 60 4/1 Notes:

1. Heatup/Cooldown = RCS Heatup and RCS Cooldown from Tables 5-1 and 5-12 of [3-5].

2. Plant Loading/Unloading = Loading 0% - 15% (S/G A etc.) from Tables 5-2 and 5-13 of

[3-5].

3. Loss of Load = Reactor Trip (large & small deltaP) from Tables 5-1 and 5-12 of [3-5].

4. Loss of Power = Loss/Pwr/Blackout+NatCirc from Tables 5-1 and 5-12 of [3-5].

Table 3-5. MNS1 Inspection History SG Primary Side Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam B2.40 1SGB-W22 3/23/2010 3rd/3rd/M1R20 RSG Acceptable 98% N/A No B2.40 1SGB-W22 3/31/2019 4th/3rd/M1R26 RSG Acceptable 100% N/A Yes

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Page 10 of 17 SG Secondary Side Shell Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam C1.20 1SGD-W144 9/23/2002 3rd/1st/M1R15 RSG Acceptable 100% N/A No C1.20 1SGD-W144 3/24/2013 4th/1st/M1R22 RSG Acceptable 100% N/A No C1.30 1SGA-W65 9/24/2002 3rd/1st/M1R15 RSG Acceptable 100% N/A No C1.30 1SGA-W65 3/20/2013 4th/1st/M1R22 RSG Acceptable 99.90% N/A No SG Secondary Side Nozzle Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam C2.21 1SGA-W258 9/25/2002 3rd/1st/M1R15 RSG Acceptable 92.90% N/A No C2.21 1SGA-W258 10/3/2014 4th/1st/M1R23 RSG Acceptable 100% N/A No C2.22 1SGA-W258 9/25/2002 3rd/1st/M1R15 RSG Acceptable 100% N/A No C2.22 1SGA-W258 10/3/2014 4th/1st/M1R23 RSG Acceptable 100% N/A No

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Page 11 of 17 Table 3-6. MNS 2 Inspection History SG Primary Side Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam B2.40 2SGC-W22 9/27/2006 3rd/1st/M2R17 RSG Acceptable 100% N/A No B2.40 2SGB-W22 3/29/2020 4th/2nd/M2R26 RSG Acceptable 99% N/A Yes SG Secondary Side Shell Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam C1.20 2SGC-W144 9/19/2003 2nd/3rd/M2R15 RSG Acceptable 100% N/A No C1.20 2SGC-W144 4/4/2014 3rd/3rd/M2R22 RSG Acceptable 100% N/A No C1.30 2SGA-W65 3/14/2005 3rd/1st/M2R16 RSG Acceptable 95.10% N/A No C1.30 2SGA-W65 9/23/2015 4th/1st/M2R23 RSG Acceptable 99.90% N/A No

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Page 12 of 17 SG Secondary Side Nozzle Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam C2.21 2SGA-W258 3/18/2005 3rd/1st/M2R16 RSG Acceptable 95.80% N/A No C2.21 2SGA-W258 9/24/2015 4th/1st/M2R23 RSG Acceptable 100% N/A No C2.22 2SGA-W258 3/18/2005 3rd/1st/M2R16 RSG Acceptable 100% N/A No C2.22 2SGA-W258 4/7/2017 4th/1st/M2R24 RSG Acceptable 100% N/A No

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Page 13 of 17 Figure 3-1. MNS1/2 Steam Generator Layout [3-6] I < ----1------ - i i W)*J ,/ ~~~+----+-~wre'-"-----le----ll--- ,1t*,o,

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Page 14 of 17 Figure 3-2. MNS1/2 Item No. B2.40 Weld Configuration ([3-7] and [3-8]) Figure 3-3a. MNS1/2 Item No. C1.20 Weld Configuration [3-9] ll.&:SHEET TO tEfoO -'ELD Ce1Al!. Wl44

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Page 15 of 17 Figure 3-3b. MNS1/2 Item No. C1.30 Weld Configuration [3-9] f\\) f\\) I V CD CD ( d l ll ~ I\\)

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Page 16 of 17 Figure 3-4. MNS1/2 Feedwater Nozzle Configuration [3-10] References 3-1. Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906. 3-2. Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Inside Radius Sections. EPRI, Palo Alto, CA: 2019. 3002014590. 3-3. Drawing MCM 1201.01-0684.001, MNS1 Steam Generator Arrangement, Revision D5. 3-4. Drawing MCM 2201.01-0126.001, MNS2 Steam Generator Arrangement, Revision D8. 3-5. MCC-1206.02-45-0040, SI Calculation FP-MNS-311 - McGuire SI:FatiguePro 4.0 Baseline Analysis Startup through 9/24/2017 (U1) and 9/16/2018 (U2), Revision 0.

• 0.000' 0.7*0" *O.QISO I,..

PA.RTI/.L SE'CTIONIL ELf'V/.TION VlfH MA(N FEEDWATER NJZZLE

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Page 17 of 17 3-6. Drawing MCM 2201.01-0207.001, Layout of Vessel Ref. Points for Welds, Revision 10. 3-7. Drawing MCM 1201.01-0791.001, Tubesheet & Primary Head Assy, Revision 5. 3-8. Drawing MCM 2201.01-0216.001, Tubesheet & Primary Head Assy, Revision 5. 3-9. Drawing MCM 1201.01-0782.001, Layout of Vessel Ref. Points for Welds, Revision 10. 3-10. Drawing MCM 2201.01-0159.001, Main Feedwater Nozzle, Revision 4. 3-11. MNS1/2 UFSAR Section 5.2.4.1. 3-12. Drawing MCM 1201.01-0748 001, Main Feedwater Nozzle Assy, Revision 4. 3-13. Drawing MCM 2201.01-0078 001, Thermal Sleeve, Revision 9.

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Page 1 of 18 ATTACHMENT 4 PLANT-SPECIFIC APPLICABILITY ONS1/2/3

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Page 2 of 18 Section 9 of References 4-1 and 4-2 provide requirements that must be demonstrated in order to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for ONS1/2/3 is provided in Table 4-1 and indicates that all plant-specific requirements are met. Therefore, the results and conclusions of the EPRI reports are applicable to ONS1/2/3. Table 4-1 Applicability of References [4-1] and [4-2] Representative Analyses to ONS1/2/3 Items No. B2.40 (SG Primary Side Shell Welds) Category Requirement from Reference [4-1] Applicability to ONS1/2/3 General Requirements The Loss of Power transient (involving auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of portion of the vessel) is not considered in this evaluation due to its rarity. In the event that such a significant thermal event occurs at a plant, its impact on the KIC (material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this reports guidance. For the replacement SGs that were installed in 2003 and 2004 and are currently in service, ONS has not experienced a loss of power transient resulting in unheated auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of any portion of the vessel. The materials of the SG vessel heads and tubesheet must be low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. The ONS1/2/3 SG vessel heads and tubesheet are fabricated of SA-508, Class 3a material (Reference [4-3]). The RTNDT values for the ONS1/2/3 SG vessel head and tubesheet materials are 0°F or less (Reference [4-5]) (so the RTNDT of 60°F used in the EPRI report is bounding). This material is a low alloy ferritic steel which conforms to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. Specific Requirements The weld configurations must conform to those shown in Figures 1-1 and Figure 1-2 of Reference [4-1]. The ONS1/2/3 tubesheet-to-shell weld configuration is shown in Figure 4-2 and conforms to Figure 1-2 of Reference 4-1.

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Page 3 of 18 Category Requirement from Reference [4-1] Applicability to ONS1/2/3 The SG vessel dimensions must be within 10% of the upper and lower bounds of the values provided in the table in Section 9.4.3 of Reference [4-1]. The ONS1/2/3 SG vessel dimensions are as follows (Reference [4-4]): SG Lower & Upper Head diameter = 131.2 (OD) The dimension of the lower head is inconsistent with the 149 OD given in Table 9-2 of Reference [4-1]. This diameter was assumed the same as the secondary shell but did not account for the reduction in diameter of the head. Upon comparison with Figure 4-3 of Reference [4-1], it can be seen that the head dimension is consistent with that of the B&W design evaluated and is therefore deemed to be within acceptable geometrical tolerances. The component must experience transients and cycles bounded by those shown in Table 5-7 of Reference [4-1] over a 60-year operating life. As shown in Table 4-2, there are slight variations on some temperature and pressure values between ONS1/2/3 and Table 5-7 of Reference [4-1]. However, the ONS1/2/3 number of cycles projected to occur over a 60-year operating life are significantly lower than those shown in Table 5-7 of Reference [4-1]. See the Stress Analysis discussion in Section 5 for a basis to the applicability of Reference [4-1] for transients having a maximum pressure/temperature higher than what is listed in Reference [4-1] or a minimum pressure/temperature lower than what is listed in Reference [4-1]. Items No. C1.30 (SG Secondary Side Shell Welds) Category Requirement from Reference [4-1] Applicability to ONS1/2/3 General Requirements The Loss of Power transient (involving auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of portion of the vessel) is not considered in this evaluation due to its rarity. In For the replacement SGs that were installed in 2003 and 2004 are currently in service, ONS1/2/3 have not experienced a loss of power transient resulting in unheated auxiliary

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Page 4 of 18 Category Requirement from Reference [4-1] Applicability to ONS1/2/3 the event that such a significant thermal event occurs at a plant, its impact on the KIC (material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this reports guidance. feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of any portion of the vessel. The materials of the SG vessel shell and tubesheet must be low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. The ONS1/2/3 SG vessel shell and tubesheet are fabricated of SA-508, Class 3a material (Reference [4-3]). The RTNDT values for the ONS1/2/3 SG vessel shell and tubesheet material is 0°F or less (Reference [4-5]) (so the RTNDT of 60°F used in the EPRI report is bounding). This material is a low alloy ferritic steel which conforms to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. Specific Requirements The weld configurations must conform to those shown in Figure 1-7 and Figure 1-8 of Reference [4-1]. The ONS1/2/3 weld configuration is shown in Figure 4-3 and conforms to Figure 1-8 of Reference [4-1]. The SG vessel dimensions must be within 10% of the upper and lower bounds of the values provided in the table in Section 9.4.4 of Reference [4-1]. The ONS1/2/3 SG vessel dimensions are as follows (Reference [4-4]): SG Lower & Upper Head diameter = 144.14" (OD) The dimensions are within 10% of that specified in Table 9-3 in Section 9.4.4 of Reference 4-1 for B&W plants. The component must experience transients and cycles bounded by those shown in Table 5-9 of Reference [4-1] over a 60-year operating life. As shown in Table 4-3, there are slight variations on some temperature and pressure values between ONS1/2/3 and Table 5-9 of Reference [4-1]. However, the ONS1/2/3 number of cycles projected to occur over a 60-year operating life are significantly lower than those shown in Table 5-9 of Reference [4-1]. See the Stress Analysis discussion in Section 5 for a basis to the applicability of Reference [4-1] for transients having a maximum pressure/temperature higher than what

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Page 5 of 18 Category Requirement from Reference [4-1] Applicability to ONS1/2/3 is listed in Reference [4-1] or a minimum pressure/temperature lower than what is listed in Reference [4-1]. Items Nos. C2.21 and C2.22 (MS and FW Nozzle to Shell Welds and Inside Radius Sections) Category Requirement from Reference [4-2] Applicability to ONS1/2/3 General Requirements The nozzle-to-shell weld shall be one of the configurations shown in Figure 1-1 or Figure 1-2 of Reference [4-2]. The ONS1/2/3 MS nozzle-to-shell weld in shown in Figure 4-4 and is representative of the configuration shown in Figure 1-2 of Reference [4-2]. Per Section 4.3.1.3, Item 3 of Reference [4-2], B&W plants (like ONS1/2/3) do not have FW nozzles welded into the SG shells (the nozzle is actually a bolted joint) and have multiple penetrations in the shell that riser pipes enter to provide feedwater flow to the feedwater ring inside the SG. There are therefore no C2.21 or C2.22 components for the FW nozzle. The materials of the SG shell, FW nozzles, and MS nozzles must be low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. The ONS1/2/3 SG vessel heads, side shell, and MS nozzles are fabricated of SA-508, Class 3a material (Reference [4-5]). The RTNDT value for the material of ONS1/2/3 SG nozzle-to-shell welds are 0°F or less (Reference [4-5]) (so the RTNDT of 60°F used in the EPRI report is bounding). This material is a low alloy ferritic steel which conforms to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. Per above, there are no C2.21 or C2.22 components for the FW nozzle.

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Page 6 of 18 Category Requirement from Reference [4-2] Applicability to ONS1/2/3 The SG must not experience more than the number of all transients shown in Table 5-5 of Reference [4-2] over a 60-year operating life. As shown in Table 4-4, the ONS1/2/3 SGs are not projected to experience more than the number of transients shown in Table 5-5 of Reference [4-2] over a 60-year operating life. SG Feedwater Nozzle The piping attached to the FW nozzle must be 14-inch to 18-inch NPS. Per above, there are no C2.21 or C2.22 components for the FW nozzle. The FW nozzle design must have an integrally attached thermal sleeve. Per above, there are no C2.21 or C2.22 components for the FW nozzle. Auxiliary feedwater nozzles connected directly to the SG are not covered in this evaluation. N/A for ONS1/2/3. SG Main Steam Nozzle For Westinghouse and CE SGs, the piping attached to the SG main steam nozzle must be 28-inch to 36-inch NPS. N/A for ONS 1/2/3 (B&W design). For B&W SGs, the piping attached to the main steam nozzle must be 22-inch to 26-inch NPS. The piping attached to the ONS 1/2/3 MS nozzle is 24-inch NPS per References [4-9], [4-10], and [4-11]. The SG must have one main steam nozzle that exits the top dome of the SG. For B&W plants, there may be more than one main steam nozzle; it will exit the side of the SG. ONS1/2/3 are B&W design, with the main steam nozzle exiting the side of the SG. The main steam nozzle shall not significantly protrude into the SG (e.g., see Figure 4-7 of Reference [4-2] or have a unique nozzle weld configuration (e.g., see Figure 4-6 of Reference [4-2]). The ONS1/2/3 MS nozzle configuration (shown in Figure 4-4) does not protrude significantly into the SG as shown in Figure 4-7 of Reference [4-2] and does not have a unique weld configuration as shown in Figure 4-6 of Reference [4-2].

- ONS1/2/3 Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 7 of 18 Table 4-2 ONS1/2/3 Data for Thermal Transients for Stress Analysis of the PWR SG Primary-Side Head Welds (Comparison to Table 5-7 of Reference [4-1]) Transient Max Thot °F Min Thot °F Max Tcold °F Min Tcold °F Max Press PSIG Min Press PSIG 60-Year Projected Cycles Heatup/Cool down EPRI Report 3002015906 545 70 545 70 2235 0 300 Heatup/Cool down ONS1/2/3(1)(5) 547/551/ 559 99/78/75 546/550/ 559 84/81/74 2252/2287/ 2245 28/6/11 132/134/ 104 Plant Loading / Unloading EPRI Report 3002015906 610 550 550 545 2300 2300 5000 Plant Loading / Unloading ONS1/2/3(2)(3) (5) 599/601/ 603 590/597/ 599 561/560/ 558 544/556/ 554 2262/2241/ 2210 2191/2170 /2210 41/5/440 // 12/9/1084 Reactor Trip EPRI Report 3002015906 615 530 565 530 2435 1700 360 Reactor Trip ONS1/2/3(4)(5) 600/600/ 601 554/554/ 558 566/570/ 571 552/550/ 558 2246/2235/ 2208 1856/1826 /1849 62/33/41 Notes:

1. Heatup/Cooldown = RCS Heatup and RCS Cooldown from Tables 13, 14 and 15 of [4-7]

scaled down from 80 to 60 years.

2. Plant Loading = Power Loading 70% - 100% from Tables 13, 14 and 15 of [4-7] scaled down from 80 to 60 years.

3. Plant Unloading = Power Unloading 100% - 70% from Tables 13, 14 and 15 of [4-7]

scaled down from 80 to 60 years.

4. Reactor Trip = Rx Trip No Loss of Flow from Tables 13, 14 and 15 of [4-7] scaled down from 80 to 60 years.

5. Temperature and pressure values obtained from supporting databases associated with the subject references.

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Page 8 of 18 Table 4-3 ONS1/2/3 Data for Thermal Transients for Stress Analysis of the PWR SG Secondary-Side Vessel Welds (Comparison to Table 5-9 of Reference [4-1]) Transient Max Tss °F Min Tss °F Max Press PSIG Min Press PSIG 60-Year Projected Cycles Heatup/Cooldown EPRI Report 3002015906 545 70 1000 0 300 Heatup/Cooldown ONS1/2/3(1)(5) 542/539/552 94/93/75 913/906/1036 5/6/17 132/134/104 Plant Loading / Unloading EPRI Report 3002015906 545 540 1000 1000 5000 Plant Loading / Unloading ONS1/2/3(2)(3)(5) 549/552/559 538/550/550 956/910/912 863/909/908 41/5/440 // 12/9/1084 Reactor Trip EPRI Report 3002015906 555 530 1130 1000 360 Reactor Trip ONS1/2/3(4)(5) 555/557/552 550/551/548 1101/1111/1095 909/908/998 62/33/41 Notes:

1. Heatup/Cooldown = RCS Heatup and RCS Cooldown from Tables 13, 14 and 15 of [4-7]

scaled down from 80 to 60 years.

2. Plant Loading = Power Loading 70% - 100% from Tables 13, 14 and 15 of [4-7] scaled down from 80 to 60 years.

3. Plant Unloading = Power Unloading 100% - 70% from Tables 13, 14 and 15 of [4-7]

scaled down from 80 to 60 years.

4. Reactor Trip = Rx Trip No Loss of Flow from Tables 13, 14 and 15 of [4-7] scaled down from 80 to 60 years.

5. Temperature and pressure values obtained from supporting databases associated with the subject references.

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Page 9 of 18 Table 4-4 ONS1/2/3 Data for Thermal Transients Applicable to PWR SG Feedwater and Main Steam Nozzles (Comparison to Table 5-5 of Reference [4-2]) Transient 60-Year Allowable Cycles from Table 5-5 of EPRI Report 3002014590 [4-2] 60-Year Projected Cycles ONS1/2/3 Heatup/Cooldown(1) 300 132/134/104 Plant Loading(2) 5000 41/5/440 Plant Unloading(3) 5000 12/9/1084 Loss of Load(4) 360 62/33/41 Loss of Power 60 6/6/2 Notes:

1. Heatup/Cooldown = RCS Heatup and RCS Cooldown from Tables 13, 14 and 15 of [4-7]

scaled down from 80 to 60 years.

2. Plant Loading = Power Loading 70% - 100% from Tables 13, 14 and 15 of [4-7] scaled down from 80 to 60 years.

3. Plant Unloading = Power Unloading 100% - 70% from Tables 13, 14 and 15 of [4-7]

scaled down from 80 to 60 years.

4. Loss of Load = Rx Trip No Loss of Flow from Tables 13, 14 and 15 of [4-7] scaled down from 80 to 60 years.

5. Loss of Power = Rx Trip Loss of RC Flow from Tables 13, 14 and 15 of [4-7] scaled down from 80 to 60 years.

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Page 10 of 18 Table 4-5. ONS1 Inspection History SG Primary Side Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam B2.40 1-SGA-W22 11/3/2009 4th/2nd/O1R25 RSG Acceptable 91.90% N/A No B2.40 1-SGA-W22 11/3/2018 4th/2nd/O1R30 RSG Acceptable 91.70% N/A Yes B2.40 1-SGA-W23 11/2/2006 4th/1st/O1R23 RSG Acceptable 92.30% N/A No B2.40 1-SGA-W23 11/21/2016 5th/1st/O1R29 RSG Acceptable 92.30% N/A No SG Secondary Side Shell Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam C1.30 1-SGB-W65 4/15/2011 4th/3rd/O1R26 RSG Acceptable 98.30% N/A No C1.30 1-SGB-W65 10/30/2020 5th/2nd/O1R31a RSG Acceptable 100% N/A Yes C1.30 1-SGB-W69 4/29/2008 4th/2nd/O1R24 RSG Acceptable 90.70% N/A No C1.30 1-SGB-W69 10/29/2020 5th/3rd/O1R31 RSG Acceptable 86% Yes

• End of 5th Interval Limited exam relief pending via NCR 02244461-01

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Page 11 of 18 SG Secondary Side Nozzle Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam C2.21 1-SGA-W128 11/14/2006 4th/1st/O1R23 RSG Acceptable 85.66% 07-ON-002* No C2.21 1-SGA-W128 11/13/2016 5th/1st/O1R29 RSG Acceptable 100% N/A No C2.21 1-SGA-W127 11/3/2009 4th/2nd/O1R25 RSG Acceptable 100% N/A No C2.21 1-SGA-W127 10/29/2018 5th/2nd/O1R30 RSG Acceptable 100% N/A No

• NRC SER via ADAMS Accession Number ML082480215 Table 4-6. ONS 2 Inspection History SG Primary Side Welds Item No.

Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam B2.40 2-SGA-W22 11/17/2008 4th/2nd/O2R23 RSG Acceptable 92.30% N/A No B2.40 2-SGA-W22 11/21/2019 5th/2nd/O2R29 RSG Acceptable 100% N/A Yes B2.40 2-SGA-W23 11/17/2008 4th/2nd/O2R23 RSG Acceptable 92.30% N/A No B2.40 2-SGA-W23 11/21/2019 5th/2nd/O2R29 RSG Acceptable 100% N/A Yes

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Page 12 of 18 SG Secondary Side Shell Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam C1.30 2-SGB-W65 10/21/2013 4th/3rd/O2R26 RSG Acceptable 97.90% N/A No C1.30 2-SGB-W65 11/16/2021 5th/2nd/O230 RSG Acceptable 100% N/A Yes C1.30 2-SGB-W69 10/23/2013 4th/3rd/O2R26 RSG Acceptable 75.10% 15-ON-002* No

• NRC SER via ADAMS Accession Number ML16197A011 SG Secondary Side Nozzle Welds Item No.

Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam C2.21 2-SGA-W127 11/5/2011 4th/3rd/O2R25 RSG Acceptable 100% N/A No C2.21 2-SGA-W127 11/19/2019 5th/2nd/O2R29 RSG Acceptable 100% N/A No C2.21 2-SGA-W128 11/19/2008 4th/2nd/O2R23 RSG Acceptable 100% N/A No C2.21 2-SGA-W128 11/19/2019 5th/2nd/O2R29 RSG Acceptable 100% N/A No Table 4-7. ONS 3 Inspection History SG Primary Side Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam B2.40 3-SGB-W23 11/10/2010 4th/2nd/O3R25 RSG Acceptable 92.30% N/A No B.40 3-SGB-W23 5/21/2022 5th/2nd/O3R31a RSG Acceptable 100% N/A Yes B2.40 3-SGB-W22 5/11/2009 4th/2nd/O3R24 RSG Acceptable 100% N/A No B2.40 3-SGB-W22 4/20/2020 5th/2nd/O3R30 RSG Acceptable 100% N/A Yes

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Page 13 of 18 SG Secondary Side Shell Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam C1.30 3-SGA-W69 5/7/2012 4th/3rd/O3R26 RSG Acceptable 75.60% 13-ON-001* No C1.30 3-SGA-W69 5/20/2022 5th/3rd/O3R31 RSG Acceptable 86% Yes C1.30 3-SGB-W65 5/4/2009 4th/2nd/O3R24 RSG Acceptable 97.60% N/A No C1.30 3-SGB-W65 4/20/2020 5th/2nd/O3R30 RSG Acceptable 100% N/A Yes

• NRC SER via ADAMS Accession Number ML14216A476
  • End of 5th Interval Limited exam relief pending via NCR 02244461-01 SG Secondary Side Nozzle Welds Item No.

Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam C2.21 3-SGA-W127 11/13/2007 4th/1st/O3R23 RSG Acceptable 100% N/A No C2.21 3-SGA-W127 4/29/2018 5th/1st/O3R29 RSG Acceptable 100% N/A No C2.21 3-SGA-W128 11/13/2007 4th/1st/O3R23 RSG Acceptable 100% N/A No C2.21 3-SGA-W128 4/30/2018 5th/1st/O3R29 RSG Acceptable 100% N/A No

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Page 14 of 18 Figure 4-1. ONS1/2/3 Steam Generator Layout (Figure 1.2.1 of Reference [4-3]) UPPER HEAD AUXILIARY FEEOWATER INLET LOWER HEAD REACTOR COOtANT ouner MMWAY U1'PER TUSE.SMEET MANWAY INSPECTION PORT STEAM OIJTI.ET '-MAIN FEEOWATER INLET 1\\JBlNG WHWAY LOWER TVBESHEET MANWAY BASE S\\JPPORT

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Page 15 of 18 Figure 4-2. ONS1/2/3 Item No. B2.40 Weld Configuration [4-8] e* 'rlANOHOIL_ / l'tl80 DUAIL 'D' VEN-CONM W162 DETAIL '(I PUNCH MARK: 16 \\ -,J AR~ TO BE ON \\ \\ BASE. MATERlA:L W2J DETAIL 'B' OR DETAIL 'Bl' J ! w2&W3 I DETA!L 'K' I !SUPPORT STOOL I wn DET A1L '8' OR DETAIL '81

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Page 16 of 18 Figure 4-3. ONS1/2/3 Item No. C1.30 Weld Configuration [4-8] rs Q. Q. I I I 1 3" INSP PORT W87 (Y-2) DETAIL 'E' I DE' W69 c.5,.,.,1 I 2* 2 W65 om,*A2t) LI l~ 2* 2"

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Page 17 of 18 Figure 4-4. ONS1/2/3 Main Steam Nozzle Configuration [4-4] References 4-1. Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906. 4-2. Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Inside Radius Sections. EPRI, Palo Alto, CA: 2019. 3002014590. 4-3. Manual OSC-8319, Replacement Steam Generators Transient Analysis Stress Report, Revision D5. 4-4. Drawing OM-201.S-0001, General Arrangement, Revision 5. 4-5. Manual OM-201.S-0005.001, Replacement Once Through Steam Generator O&M Manual, Revision 5. 4-6. Drawing OM-201.S0089.001, Steam Outlet Nozzle, Revision 1. 4-7. SI Calculation FP-ONS-304, Oconee SI:FatiguePro 4 Baseline Analysis, Startup through 11/3/2020 (U1), 11/21/2019 (U2) and 4/24/2020 (U3), Revision 0. 4-8. Drawing OM 201.S--0156.001, Drawing UT/RT Markings Layout, Revision 3. 4-9. Drawing OFD-122A-1.1, Flow Diagram of Main Steam System, Revision 27. 24.25" DIA 1

• 22,265" ID

" I (D STEAM OUTLET NOZZLf;_ SCALE: 1",.,1'-0°

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Page 18 of 18 4-10. Drawing OFD-122A-2.1, Flow Diagram of Main Steam System, Revision 28. 4-11. Drawing OFD-122A-3.1, Flow Diagram of Main Steam System, Revision 35.

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Page 1 of 16 ATTACHMENT 5 PLANT-SPECIFIC APPLICABILITY HNP

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Page 2 of 16 Section 9 of References [5-1] and [5-2] provide requirements that must be demonstrated in order to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for HNP is provided in Table 5-1 and indicates that all plant-specific requirements are met. Therefore, the results and conclusions of the EPRI reports are applicable to HNP. Table 5-1 Applicability of References [5-1] and [5-2] Representative Analyses to HNP Items No. B2.40 (SG Primary Side Shell Welds) Category Requirement from Reference [5-1] Applicability to HNP General Requirements The Loss of Power transient (involving auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of portion of the vessel) is not considered in this evaluation due to its rarity. In the event that such a significant thermal event occurs at a plant, its impact on the KIC (material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this reports guidance. For the replacement SGs that were installed in 2001 and are currently in service, HNP has not experienced a loss of power transient resulting in unheated auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of any portion of the vessel. The materials of the SG vessel heads and tubesheet must be low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. The HNP SG vessel heads and tubesheet are fabricated of SA-508, Class 3a material (Reference [5-3]). The RTNDT values for the HNP SG vessel head and tubesheet materials are 10°F or less (Reference [5-9]) (so the RTNDT of 60°F used in the EPRI report is bounding). This material is a low alloy ferritic steel which conforms to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. Specific Requirements The weld configurations must conform to those shown in Figures 1-1 and Figure 1-2 of Reference [5-1]. The HNP tubesheet-to-shell weld configuration is shown in Figure 5-2 and conforms to Figure 1-2 of Reference [5-1].

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Page 3 of 16 Category Requirement from Reference [5-1] Applicability to HNP The SG vessel dimensions must be within 10% of the upper and lower bounds of the values provided in the table in Section 9.4.3 of Reference [5-1]. The HNP SG vessel dimensions are as follows (Reference [5-4]): SG Lower Head diameter = 135.7" (OD) SG Upper Shell diameter = 176.26 (OD) The dimensions are within 10% of that specified in Table 9-2 in Section 9.4.3 of Reference [5-1]. The component must experience transients and cycles bounded by those shown in Table 5-7 of Reference [5-1] over a 60-year operating life. As shown in Table 5-2, there are slight variations in some temperature and pressure values between HNP and the Reference [5-1] values. However, the HNP number of cycles projected to occur over a 60-year life are significantly lower than those in shown in Table 5-7 of Reference [5-1]. See the Stress Analysis discussion in Section 5 for a basis to the applicability of Reference [5-1] for transients having a maximum pressure/temperature higher than what is listed in Reference [5-1] or a minimum pressure/temperature lower than what is listed in Reference [5-1]. Items No. C1.20 and C1.30 (SG Secondary Side Shell Welds) Category Requirement from Reference [5-1] Applicability to HNP General Requirements The Loss of Power transient (involving auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of portion of the vessel) is not considered in this evaluation due to its rarity. In the event that such a significant thermal event occurs at a plant, its impact on the KIC (material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this reports guidance. For the replacement SGs that were installed in 2001 and are currently in service, HNP has not experienced a loss of power transient resulting in unheated auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of any portion of the vessel.

- HNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 4 of 16 Category Requirement from Reference [5-1] Applicability to HNP The materials of the SG vessel shell and tubesheet must be low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. The HNP SG vessel shell and tubesheet are fabricated of SA-508, Class 3a material (Reference [5-3]). The RTNDT values for the HNP SG vessel shell and tubesheet materials are 10°F or less (Reference [5-9]) (so the RTNDT of 60°F used in the EPRI report is bounding). This material is a low alloy ferritic steel which conforms to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. Specific Requirements The weld configurations must conform to those shown in Figure 1-7 and Figure 1-8 of Reference [5-1]. The HNP weld configurations are shown in Figures 5-3a and 5-3b, and conform to Figure 1-7 and Figure 1-8 of Reference [5-1]. The SG vessel dimensions must be within 10% of the upper and lower bounds of the values provided in the table in Section 9.4.4 of Reference [5-1]. The HNP SG vessel dimensions are as follows (Reference [5-4]): SG Lower Head diameter = 135.7" (OD) SG Upper Shell diameter = 176.26 (OD) The dimension of the upper shell is within 10% of that specified in Table 9-3 in Section 9.4.4 of Reference [5-1]. The component must experience transients and cycles bounded by those shown in Table 5-9 of Reference [5-1] over a 60-year operating life. As shown in Table 5-3, there are slight variations on some temperature and pressure values between HNP and the Reference [5-1] values. However, the HNP number of cycles projected to occur over a 60-year life are significantly lower than those in shown in Table 5-9 of Reference [5-1]. See the Stress Analysis discussion in Section 5 for a basis to the applicability of Reference [5-1] for transients having a maximum pressure/temperature higher than what is listed in Reference [5-1] or a minimum

- HNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 5 of 16 Category Requirement from Reference [5-1] Applicability to HNP pressure/temperature lower than what is listed in Reference [5-1]. Items Nos. C2.21 and C2.22 (MS and FW Nozzle to Shell Welds and Inside Radius Sections) Category Requirement from Reference [5-2] Applicability to HNP General Requirements The nozzle-to-shell weld shall be one of the configurations shown in Figure 1-1 or Figure 1-2 of Reference [5-2]. The HNP FW nozzle-to-shell weld in shown in Figure 5-4 and is representative of the configuration shown in Figure 1-2 of Reference [5-2]. The HNP MS nozzle is forged and does not have a C2.21 or C2.22 component). The materials of the SG shell, FW nozzles, and MS nozzles must be low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. The HNP SG vessel heads, side shell, and FW nozzles are fabricated of SA-508, Class 3a material (Reference [5-3]). The RTNDT value for the material of the HNP SG shell and nozzles is 10°F or less (Reference [5-9]) (so the RTNDT of 60°F used in the EPRI report is bounding). This material is a low alloy ferritic steel which conforms to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. As noted above, the HNP MS nozzle is forged and does not have a C2.21 or C2.22 component. The SG must not experience more than the number of all transients shown in Table 5-5 of Reference [5-2] over a 60-year operating life. As shown in Table 5-4, the HNP SGs are not projected to experience more than the number of transients shown in Table 5-5 of Reference [5-2] over a 60-year operating life. SG Feedwater Nozzle The piping attached to the FW nozzle must be 14-inch to 18-inch NPS. The piping attached to the FW nozzle is 16-inch NPS per Reference [5-10].

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Page 6 of 16 Category Requirement from Reference [5-2] Applicability to HNP The FW nozzle design must have an integrally attached thermal sleeve. The HNP FW nozzle configuration is shown in Figure 5-4 and has an integrally attached thermal sleeve per Reference [5-6]. Auxiliary feedwater nozzles connected directly to the SG are not covered in this evaluation. N/A for HNP. SG Main Steam Nozzle For Westinghouse and CE SGs, the piping attached to the SG main steam nozzle must be 28-inch to 36-inch NPS. N/A for HNP (the MS nozzle is forged and does not have a C2.21 or C2.22 component). For B&W SGs, the piping attached to the main steam nozzle must be 22-inch to 26-inch NPS. This requirement is not applicable for HNP because it is a Westinghouse 3-loop PWR. The SG must have one main steam nozzle that exits the top dome of the SG. For B&W plants, there may be more than one main steam nozzle; it will exit the side of the SG. N/A for HNP (the MS nozzle is forged and does not have a C2.21 or C2.22 component). The main steam nozzle shall not significantly protrude into the SG (e.g., see Figure 4-7 of Reference [5-2] or have a unique nozzle weld configuration (e.g., see Figure 4-6 of Reference [5-2]). N/A for HNP (the MS nozzle is forged and does not have a C2.21 or C2.22 component).

- HNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 7 of 16 Table 5-2 HNP Data for Thermal Transients for Stress Analysis of the PWR SG Primary-Side Head Welds (Comparison to Table 5-7 of Reference [5-1]) Transient Max Thot °F Min Thot °F Max Tcold °F Min Tcold °F Max Press PSIG Min Press PSIG 60-Year Projected Cycles Heatup/Cooldown EPRI Report 3002015906 545 70 545 70 2235 0 300 Heatup/Cooldown HNP(1)(5) 566 121 561 111 2242 21 84 Plant Loading / Unloading EPRI Report 3002015906 610 550 550 545 2300 2300 5000 Plant Loading / Unloading HNP(2)(3)(5) 603 593 555 555 2241 2239 819/728 Reactor Trip EPRI Report 3002015906 615 530 565 530 2435 1700 360 Reactor Trip HNP(4)(5) 622 565 559 558 2238 1997 1 Notes:

1. Heatup/Cooldown = Plant Heatup and Plant Cooldown from Table 5-2 of [5-7].

2. Plant Loading = Unit Loading 15% - 100% from Table 5-2 of [5-7].

3. Plant Unloading = Unit Unloading 100% - 15% from Table 5-2 of [5-7]

4. Reactor Trip = Loss of Load from Table 5-2 of [5-7].

5. Temperature and pressure values obtained from supporting databases associated with the subject references.

- HNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 8 of 16 Table 5-3 HNP Data for Thermal Transients for Stress Analysis of the PWR SG Secondary-Side Vessel Welds (Comparison to Table 5-9 of Reference 5-1) Transient Max Tss °F Min Tss °F Max Press PSIG Min Press PSIG 60-Year Projected Cycles Heatup/Cooldown EPRI Report 3002015906 545 70 1000 0 300 Heatup/Cooldown HNP(1)(6) Note 5 Note 5 1079 62 84 Plant Loading / Unloading EPRI Report 3002015906 545 540 1000 1000 5000 Plant Loading / Unloading HNP(2)(3)(6) Note 5 Note 5 1028 1011 819/728 Reactor Trip EPRI Report 3002015906 555 530 1130 1000 360 Reactor Trip HNP(4)(6) Note 5 Note 5 1112 1014 1 Notes:

1. Heatup/Cooldown = Plant Heatup and Plant Cooldown from Table 5-2 of [5-7].

2. Plant Loading = Unit Loading 15% - 100% from Table 5-2 of [5-7].

3. Plant Unloading = Unit Unloading 100% - 15% from Table 5-2 of [5-7]

4. Reactor Trip = Loss of Load from Table 5-2 of [5-7].

5. Value not available.

6. Temperature and pressure values obtained from supporting databases associated with the subject references.

- HNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 9 of 16 Table 5-4 HNP Data for Thermal Transients Applicable to PWR SG Feedwater and Main Steam Nozzles [Ref(s)] (Comparison to Table 5-5 of Reference 5-2) Transient 60-Year Allowable Cycles from Table 5-5 of EPRI Report 3002014590 [5-2] 60-Year Projected Cycles HNP Heatup/Cooldown(1) 300 84 Plant Loading(2) 5000 819 Plant Unloading(3) 5000 728 Loss of Load(4) 360 1 Loss of Power(5) 60 2 Notes:

1. Heatup/Cooldown = Plant Heatup and Plant Cooldown from Table 5-2 of [5-7].

2. Plant Loading = Unit Loading 15% - 100% from Table 5-2 of [5-7].

3. Plant Unloading = Unit Unloading 100% - 15% from Table 5-2 of [5-7]

4. Loss of Load = Loss of Load from Table 5-2 of [5-7].

5. Loss of Power = Loss of Power from Table 5-2 of [5-7].

Table 5-5. HNP Inspection History SG Primary Side Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam B2.40 II-SG-001SGA-TSTHW-06-01 10/25/2004 2nd/3rd/H1R12 RSG Acceptable 74.20% 2R1-022* No B2.40 II-SG-001SGB-TSTHW-06-01 5/4/2012 3rd/2nd/H1R17 RSG Acceptable 70.10% I3R-19** No

• NRC SER ADAMS Accession No. ML093561419
  • NRC SER ADAMS Accession No. ML20080G950

- HNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 10 of 16 SG Secondary Side Shell Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam C1.20 II-SG-001SGA-STH-02-1 4/24/2006 2nd/3rd/H1R13 RSG Acceptable 100% N/A No C1.20 II-SG-001SGA-STH-02-1 4/23/2015 3rd/3rd/H1R19 RSG Acceptable 100% N/A No C1.30 II-SG-001SGA-TSTSW-09-1 4/26/2006 2nd/3rd/H1R13 RSG Acceptable 90.50% N/A No C1.30 II-SG-001SGA-TSTSW-09-1 4/18/2015 3rd/3rd/H1R19 RSG Acceptable 99.70% N/A No SG Secondary Side Nozzle Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam C2.21 II-SG-001SGB-FWNTSW-05-1 5/5/2012 3rd/2nd/H1R17 RSG Acceptable 90.07% N/A No C2.21 II-SG-001SGB-FWNTSW-05-1 5/3/2021 4th/2nd/H1R23 RSG Acceptable 94.60% N/A Yes C2.22 II-SG-001SGB-FWNIR-05-1 4/26/2006 2nd/3rd/H1R13 RSG Acceptable 100% N/A No C2.22 II-SG-001SGB-FWNIR-05-1 5/5/2012 3rd/2nd/H1R17 RSG Acceptable 100% N/A No

- HNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 11 of 16 Figure 5-1. HNP Steam Generator Layout (page 247 of Reference [5-3])

- HNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 12 of 16 Figure 5-2. HNP Item No. B2.40 Weld Configuration ([5-4] and [5-8]) .oo 'P. 00 SE£ NOTE J FOR UHIT AXES IOENTIFICA TION OMAW. SFA-5.14 CL. ERNiC,Fo OR OTAW. SFA-5.14 CL. ERNICr F.-7 OR SMAW, SFA-5.11 CL. EN1Cr Fe-7 Wf:LO OE PO St TEO CL A DOINO ON THE TUBC PLATE PRIMARY SURI' ACE SHOWN. B CLOSURE RING WITH C20l X.94 MIN. f"IJLL THREAC SEE OWG. 6148E72 f"OR OETAILS. 2.00 PARTITION F S8-168 ALLOY 6~ .22 THK. CLADDING R 62.59 SPHERICAi RADIUS (CLADDING: e~ *rYP. R.38 TYP

• 121121.75 REI',

l . 10 .28 62.81 SPHERICAL RADIUS BASE METAL I OET All 8 (SH.II '='::::: TUBE PLATE TO Cl!AHNEL l<EAO GlllTli WELD SEE NOTE 2

- HNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 13 of 16 Figure 5-3a. HNP Item No. C1.20 Weld Configuration ([5-4] and [5-8]) R,38 TYP. 03 THRU 09 I 'l OET AIL A tsH.11 SEC0NOARY SIDE 0IRTtl Wf:LDS SHELL BARRELS, TAMSITI0II C0'4E, UPPER tlEAO ANO TUBE PLATE SEE NOTE 2 0 IT'I.... w-C, 10

- HNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 14 of 16 Figure 5-3b. HNP Item No. C1.30 Weld Configuration ([5-4] and [5-8]) R.38 TYP. ~---- 03 THRU 09 DETAIL A (SH. 3) DETAIL G (SH. J} O~TAI~ B SH. J 2.00 PARTITION PLATE SS-168 ALLOY 690 6° TYP, 02 THRU 08 I .,o .,e 'l DETAIL A tsK.11 SECONOARY SIDE OUHH WELOS SMELi. BAAREI..S, TAAN51TIOH COHE, UPPER ~EA.DANO TUBE PLATE SEE NOTE 2

- HNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 15 of 16 Figure 5-4. HNP Feedwater Nozzle Configuration [5-6] References 5-1. Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906. 5-2. Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Inside Radius Sections. EPRI, Palo Alto, CA: 2019. 3002014590. 5-3. Manual VM-GEN, Vertical Steam Generator, Revision 4. -s,_s* TYP. I DETAIL B .. ~u..ea.v o, f l-lEIINM. ~teve TO f'£E0WA TER 1111.t!f NOlll..£ SEE NOT"£ 7 !&3,MJ..... ,_ ______ n.za 11u--------<

- HNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 16 of 16 5-4. Drawing 1364-098001-S01, Westinghouse Delta 75 Steam Generator Arrangement, Revision 4. 5-5. Drawing 1364-098000-S01, Westinghouse Delta 75 Steam Generator Outline, Revision 8. 5-6. Drawing 1364-098099-S02, Westinghouse Delta 75 Steam Generator Feedwater Nozzle and Thermal Sleeve Assembly, Revision 0. 5-7. FP-HNP-315, SI Calculation Harris SI:FatiguePro 4.0 Baseline Analysis Startup through October 12, 2019, Revision 0. 5-8. Drawing 1364-098001-S03, Westinghouse Delta 75 Steam Generator Outline, Revision 4. 5-9. Design Specification No. 412A86, Revision 5. 5-10. Drawing 5-G-0044, Flow Diagram Feedwater & Auxiliary Feedwater Systems Unit 1, Revision 055.

- RNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 1 of 18 ATTACHMENT 6 PLANT-SPECIFIC APPLICABILITY RNP

- RNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 2 of 18 Section 9 of References [6-1] and [6-2] provide requirements that must be demonstrated in order to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for RNP is provided in Table 6-1 and indicates that all plant-specific requirements are met. Therefore, the results and conclusions of the EPRI reports are applicable to RNP. Table 6-1 Applicability of References [6-1] and [6-1] Representative Analyses to RNP Items No. B2.40 (SG Primary Side Shell Welds) Category Requirement from Reference [6-1] Applicability to RNP General Requirements The Loss of Power transient (involving auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of portion of the vessel) is not considered in this evaluation due to its rarity. In the event that such a significant thermal event occurs at a plant, its impact on the KIC (material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this reports guidance. For the replacement SGs that were installed in 1984 and are currently in service, RNP has not experienced a loss of power transient resulting in unheated auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of any portion of the vessel. The materials of the SG vessel heads and tubesheet must be low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. The RNP SG vessel heads are fabricated of SA-302 Grade B and the tubesheet is fabricated of SA-508, Class 2a material (Reference [6-3]). The RTNDT values for the RNP SG vessel head and tubesheet materials are 60°F or less (Reference [6-11]) (so the RTNDT of 60°F used in the EPRI report is bounding). This material is a low alloy ferritic steel which conforms to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. Specific Requirements The weld configurations must conform to those shown in Figures 1-1 and Figure 1-2 of Reference [6-1]. The RNP tubesheet-to-shell weld configuration is shown in Figure 6-2 and conforms to Figure 1-2 of Reference [6-1].

- RNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 3 of 18 Category Requirement from Reference [6-1] Applicability to RNP The SG vessel dimensions must be within 10% of the upper and lower bounds of the values provided in the table in Section 9.4.3 of Reference [6-1]. The RNP SG vessel dimensions are as follows (Reference [6-5]): SG Lower Head diameter = 129.32" (OD) SG Upper Shell diameter = 166 (OD) The dimensions are within 10% of that specified in Table 9-2 in Section 9.4.3 of Reference [6-1]. The component must experience transients and cycles bounded by those shown in Table 5-7 of Reference [6-1] over a 60-year operating life. As shown in Table 6-2, there are slight variations in some temperature and pressure values between RNP and the Reference [6-1] values. However, the RNP number of cycles projected to occur over a 60-year life are significantly lower than those in shown in Table 5-7 of Reference [6-1]. See the Stress Analysis discussion in Section 5 for a basis to the applicability of Reference [6-1] for transients having a maximum pressure/temperature higher than what is listed in Reference [6-1] or a minimum pressure/temperature lower than what is listed in Reference [6-1]. Items No. C1.10, C1.20, and C1.30 (SG Secondary Side Shell Welds) Category Requirement from Reference [6-1] Applicability to RNP General Requirements The Loss of Power transient (involving auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of portion of the vessel) is not considered in this evaluation due to its rarity. In the event that such a significant thermal event occurs at a plant, its impact on the KIC (material fracture toughness) value may require more frequent examinations and other plant actions outside the scope of this reports guidance. For the replacement SGs that were installed in 1984 and are currently in service, RNP has not experienced a loss of power transient resulting in unheated auxiliary feedwater being introduced into a hot SG that has been boiled dry following blackout, resulting in thermal shock of any portion of the vessel.

- RNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 4 of 18 Category Requirement from Reference [6-1] Applicability to RNP The materials of the SG vessel shell and tubesheet must be low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. The RNP SG vessel shell is fabricated from SA-533 Grade A Class 2 material and the tubesheet is fabricated of SA-508, Class 2a material (Reference [6-3]). The RTNDT values for the RNP SG vessel shell and tubesheet materials are 60°F or less (Reference [6-11]) (so the RTNDT of 60°F used in the EPRI report is bounding). These materials are low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. Specific Requirements The weld configurations must conform to those shown in Figure 1-7 and Figure 1-8 of Reference [6-1]. The RNP weld configurations are shown in Figures 6-3a and 6-3b, and conform to Figure 1-7 and Figure 1-8 of Reference [6-1]. The SG vessel dimensions must be within 10% of the upper and lower bounds of the values provided in the table in Section 9.4.4 of Reference [6-1]. The RNP SG vessel dimensions are as follows (Reference [6-5 & 6-12]): SG Lower Head diameter = 129.32" (OD) SG Upper Shell diameter = 166 (OD) The dimensions are within 10% of that specified in Table 9-3 in Section 9.4.4 of Reference [6-1]. The component must experience transients and cycles bounded by those shown in Table 5-9 of Reference [6-1] over a 60-year operating life. As shown in Table 6-3, there are slight variations in some temperature and pressure values between RNP and the Reference [6-1] values. However, the RNP number of cycles projected to occur over a 60-year life are significantly lower than those in shown in Table 5-9 of Reference [6-1]. See the Stress Analysis discussion in Section 5 for a basis to the applicability of Reference [6-1] for transients having a maximum pressure/temperature higher than what is listed in Reference [6-1] or a minimum

- RNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 5 of 18 Category Requirement from Reference [6-1] Applicability to RNP pressure/temperature lower than what is listed in Reference [6-1]. Items Nos. C2.21 and C2.22 (MS and FW Nozzle to Shell Welds and Inside Radius Sections) Category Requirement from Reference [6-2] Applicability to RNP General Requirements The nozzle-to-shell weld shall be one of the configurations shown in Figure 1-1 or Figure 1-2 of Reference [6-2]. The RNP FW nozzle-to-shell weld in shown in Figure 6-4 and is representative of the configuration shown in Figure 1-2 of Reference [6-2]. The RNP MS nozzle-to-shell weld in shown in Figure 6-5 and is representative of the configuration shown in Figure 1-2 of Reference [6-2]. The materials of the SG shell, FW nozzles, and MS nozzles must be low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. The RNP SG vessel head is fabricated of SA-302 Grade B and side shells are fabricated from SA-533 Grade A Class 2 material (Reference [6-4]). The RNP FW nozzles are fabricated from SA-508 Class 2A (Reference [6-4]). The RNP MS nozzles are fabricated from SA-533 Grade A Class 2 material (Reference [6-4]). The RTNDT value for the material of RNP SG shell and nozzle materials is 60°F or less (Reference [6-11]) (so the RTNDT of 60°F used in the EPRI report is bounding). These materials are low alloy ferritic steels which conform to the requirements of ASME Code, Section XI, Appendix G, Paragraph G-2110. The SG must not experience more than the number of all transients shown in Table 5-5 of Reference [6-2] over a 60-year operating life. As shown in Table 6-4, the RNP SGs are not projected to experience more than the number of transients shown in Table 5-5 of Reference [6-2].

- RNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 6 of 18 Category Requirement from Reference [6-2] Applicability to RNP SG Feedwater Nozzle The piping attached to the FW nozzle must be 14-inch to 18-inch NPS. The piping attached to the RNP FW nozzle is 16-inch NPS per Reference [6-9]. The FW nozzle design must have an integrally attached thermal sleeve. The RNP FW nozzle configuration is shown in Figure 6-4 and has an integrally attached thermal sleeve. Auxiliary feedwater nozzles connected directly to the SG are not covered in this evaluation. N/A for RNP. SG Main Steam Nozzle For Westinghouse and CE SGs, the piping attached to the SG main steam nozzle must be 28-inch to 36-inch NPS. The RNP main steam nozzles have 28 to 26 reducers Reference [6-7 & 6-10]. The pipe size of the attached reducer to the nozzle end is 28-inch NPS which satisfies the intent of this requirement Reference [6-10]. For B&W SGs, the piping attached to the main steam nozzle must be 22-inch to 26-inch NPS. This requirement is not applicable for RNP because it is a Westinghouse 3-loop PWR. The SG must have one main steam nozzle that exits the top dome of the SG. For B&W plants, there may be more than one main steam nozzle; it will exit the side of the SG. As shown in Figure 6-1, RNP has one MS nozzle per SG that exits the top dome of the SG. The main steam nozzle shall not significantly protrude into the SG (e.g., see Figure 4-7 of Reference [6-2] or have a unique nozzle weld configuration (e.g., see Figure 4-6 of Reference [6-2]). The RNP MS nozzle configuration (shown in Figure 6-5) does not protrude significantly into the SG as shown in Figure 4-7 of Reference [6-2] and does not have a unique weld configuration as shown in Figure 4-6 of Reference [6-2].

- RNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 7 of 18 Table 6-2 RNP Data for Thermal Transients for Stress Analysis of the PWR SG Primary-Side Head Welds [Ref(s)] (Comparison to Table 5-7 of Reference [6-1]) Transient Max Thot °F Min Thot °F Max Tcold °F Min Tcold °F Max Press PSIG Min Press PSIG 60-Year Projected Cycles Heatup/Cooldown EPRI Report 3002015906 545 70 545 70 2235 0 300 Heatup/Cooldown RNP(1)(5) 553 84 548 85 2245 19 127/126 Plant Loading / Unloading EPRI Report 3002015906 610 550 550 545 2300 2300 5000 Plant Loading / Unloading RNP(2)(3)(5) Note 6 Note 6 Note 6 Note 6 Note 6 Note 6 359/339 Reactor Trip EPRI Report 3002015906 615 530 565 530 2435 1700 360 Reactor Trip RNP(4)(5) Note 6 Note 6 Note 6 Note 6 Note 6 Note 6 246 Notes:

1. Heatup/Cooldown = Plant Heatup/Plant Cooldown from [6-6] projected to 60 years.

2. Plant Loading = Unit Loading+Step-Load Increase from [6-6] projected to 60 years.

3. Plant Unloading = Unit Unloading+Step-Load Decrease+Large Step-Load Decrease from [6-6] projected to 60 years.

4. Reactor Trip = Loss of Flow+Loss of Load+Reactor Trip from [6-6] projected to 60 years.

5. Temperature and pressure values obtained from supporting databases associated with the subject references.

6. Value not available.

- RNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 8 of 18 Table 6-3 RNP Data for Thermal Transients for Stress Analysis of the PWR SG Secondary-Side Vessel Welds [Ref(s)] (Comparison to Table 5-9 of Reference [6-1]) Transient Max Tss °F Min Tss °F Max Press PSIG Min Press PSIG 60-Year Projected Cycles Heatup/Cooldown EPRI Report 3002015906 545 70 1000 0 300 Heatup/Cooldown RNP(1)(5) 546 70 1000 0 127/126 Plant Loading / Unloading EPRI Report 3002015906 545 540 1000 1000 5000 Plant Loading / Unloading RNP(2)(3)(5) Note 6 Note 6 Note 6 Note 6 359/339 Reactor Trip EPRI Report 3002015906 555 530 1130 1000 360 Reactor Trip RNP(4)(5) Note 6 Note 6 Note 6 Note 6 246 Notes:

1. Heatup/Cooldown = Plant Heatup/Plant Cooldown from [6-6] projected to 60 years.

2. Plant Loading = Unit Loading+Step-Load Increase from [6-6] projected to 60 years.

3. Plant Unloading = Unit Unloading+Step-Load Decrease+Large Step-Load Decrease from [6-6] projected to 60 years.

4. Reactor Trip = Loss of Flow+Loss of Load+Reactor Trip from [6-6] projected to 60 years.

5. Temperature and pressure values obtained from supporting databases associated with the subject references.

6. Value not available.

- RNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 9 of 18 Table 6-4 RNP Data for Thermal Transients Applicable to PWR SG Feedwater and Main Steam Nozzles [Ref(s)] (Comparison to Table 5-5 of Reference [6-2]) Transient 60-Year Allowable Cycles from Table 5-5 of EPRI Report 3002014590 [6-2] 60-Year Projected Cycles RNP Heatup/Cooldown(1) 300 127/126 Plant Loading(2) 5000 359 Plant Unloading(3) 5000 339 Loss of Load(4) 360 246 Loss of Power 60 Note 5 Notes:

1. Heatup/Cooldown = Plant Heatup/Plant Cooldown from [6-6] projected to 60 years.

2. Plant Loading = Unit Loading+Step-Load Increase from [6-6] projected to 60 years.

3. Plant Unloading = Unit Unloading+Step-Load Decrease+Large Step-Load Decrease from [6-6] projected to 60 years.

4. Loss of Load = Loss of Flow+Loss of Load+Reactor Trip from [6-6] projected to 60 years.

5. Value not available.

Table 6-5. RNP Inspection History SG Primary Side Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/ RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam B2.40 105A/01 10/4/2005 4th/1st/R2R23 RSG Acceptable 94.84% N/A No B2.40 105A/01 6/1/2015 5th/1st/R2R29 RSG Acceptable 97.90% N/A No

- RNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 10 of 18 SG Secondary Side Shell Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam C1.10 205A/03 9/28/2005 4th/2nd/R2R23 RSG Acceptable 100% N/A No C1.10 205A/03 10/9/2018 5th/2nd/R2R31 RSG Acceptable 97.10% N/A No C1.10 205A/04 9/27/2005 4th/2nd/R2R23 RSG Acceptable 98.50% N/A No C1.10 205A/04 10/11/2018 5th/2nd/R2R31 RSG Acceptable 99.90% N/A No C1.10 205A/05 5/1/2004 4th/1st/R2R22 RSG Acceptable 99.20% N/A No C1.10 205A/05 11/25/2020 5th/3rd/R2R32 RSG Acceptable 94.78% N/A Yes C1.20 205A/06 5/4/2004 4th/1st/R2R22 RSG Acceptable 100% N/A No C1.20 205A/06 11/24/2020 5th/3rd/R2R32 RSG Acceptable 100% N/A Yes C1.30 205A/02 9/29/2005 4th/2nd/R2R23 RSG Acceptable 92.37% N/A No C1.30 205A/02 6/2/2015 5th/1st/R2R29 RSG Acceptable 99.80% N/A No SG Secondary Side Nozzle Welds Item No. Component ID Exam Date Interval/Period/ Outage OSG/ RSG Exam Results Coverage Relief Request Appendix VIII Qualified Exam C2.21 205A/07 10/15/2005 4th/2nd/R2R23 RSG Acceptable 100% N/A No C2.21 205A/07 6/1/2015 5th/1st/R2R29 RSG Acceptable 100% N/A No C2.21 205A/08 5/3/2004 4th/1st/R2R22 RSG Acceptable 80.42% RR-20* No C2.21 205A/08 5/26/2015 5th/1st/R2R29 RSG Acceptable 99.60% N/A No C2.22 205A/07IR 5/6/2004 4th/1st/R2R22 RSG Acceptable 100% N/A No C2.22 205A/07IR 10/11/2018 5th/2nd/R2R31 RSG Acceptable 100% N/A Yes C2.22 205A/08IR 5/4/2004 4th/1st/R2R22 RSG Acceptable 100% N/A No C2.22 205A/08IR 11/17/2020 5th/3rd/R2R32 RSG Acceptable 100% N/A Yes

• NRC SER ADAMS Accession No. ML14162A094

- RNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 11 of 18 Figure 6-1. RNP Steam Generator Layout [6-7]

- RNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 12 of 18 Figure 6-2. RNP Item No. B2.40 Weld Configuration [6-7] Stub Barrel to Lower Shell (C1.10) .aefJ.1) -~.7]R._/ 129.31 --- - ~ Dia. , i TYP.

- RNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 13 of 18 Upper Shell to Lower Transition Cone (C1.10) Upper Transition Cone to Upper Shell (C1.10) i

I).

I DETAJL-F .SO~Z.~R.--- .25[6~ 11° 15,;, -'fVP, "rT ye. AIL-K

• SEE NOTE 6

- RNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 14 of 18 Shell to Head (C1.20) Figure 6-3a. RNP Item No. C1.10 & C1.20 Weld Configurations [6-8 & 6-13] ~

- RNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 15 of 18 Figure 6-3b. RNP Item No. C1.30 Weld Configuration [6-13] .38[9.7]R. 1.0.. I DEJAIL aw> P

- RNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 16 of 18 Figure 6-4. RNP Feedwater Nozzle Configuration [6-7 & 6-13] FEED~ATER NOZZLE ~'ELD PRD' Feedwater' Nozzle End Forging __________ ! Existing Upper Shell EfMAT0R S) so II: .12 ~2.7] j - [3D] JITAIL-L 2 18.06 18.00 Dia.7 16.625 16.615 Dia. ----~. .065 1rr=m-

16. 32 16.30 Dia.

9. 50 9,38

- RNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 17 of 18 Figure 6-5. RNP Main Steam Nozzle Configuration [6-8] References 6-1. Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906. 6-2. Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Inside Radius Sections. EPRI, Palo Alto, CA: 2019. 3002014590. 6-3. Manual 728-208-63, Vertical Steam Generator, Revision 22. 6-4. Design Specification 955479, Replacement Steam Generator Lower Assembly and Upper Internals, Revision 1. r i:t,

• F/sfl..-{NoTe *9

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- RNP Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 18 of 18 6-5. Drawing HBR2-10618, Sheet 61, Inservice Inspection Drawing, Steam Generator A, CPL-205, Revision 3. 6-6. EC 419898, Updated Low Cycle Fatigue Transients Summary, Revision 0. 6-7. SG-84-04-037, Model 44F Replacement Steam Generator Stress Report for Carolina Power and Light Company H. B. Robinson Unit No. 2, Revision 0. 6-8. Drawing HBR2-10726, Upper Shell Fabrication, Revision 11. 6-9. Drawing G-190197, Sheet 4 of 5, Feedwater Condensate and Air Evacuation System Flow Diagram, Revision 75. 6-10. Drawing G-190196, Sheet 1 of 4, Main & Extraction Steam System Flow Diagram, Revision 62. 6-11. NRC Adams ML020520347, HB Robinson Unit 2, License Amendment 91, Change TS to Increase Minimum Temperature Requirement for Pressurizing the Secondary Side of the Steam Generators Above 200 psig from 70F to 120F. 6-12. Drawing HBR2-10750, #44 Series Vert. Steam Generator Outline, Revision 1. 6-13. Drawing HBR2-10735, Steam Generator - Model 44F General Arrangement, Rev. 0.

- Industry Survey Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 1 of 5 ATTACHMENT 7 RESULTS OF INDUSTRY SURVEY

- Industry Survey Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 2 of 5 Overall Industry Inspection Summary for Code Items B2.31, B2.32, B2.40, B3.130, C1.10, C1.20, and C1.30 The results of an industry survey of past inspections of SG nozzle-to-shell welds, inside radius sections and shell welds are summarized in Reference [7-1]. Table 7-1 provides a summary of the combined survey results for Item Nos. B2.31, B2.32 (see Table Note 3), B.240, B3.130, C1.10, C1.20, and C1.30. The results of the industry survey identified numerous steam generator (SG) examinations being performed with no service-induced flaws being detected. Performing these examinations adversely impact outage activities including worker exposure, personnel safety, and radwaste. A total of 74 domestic and international boiling water reactor (BWR) and pressurized water reactor (PWR) units responded to the survey and provided information representing all PWR plant designs currently in operation in the United States. This included 2-loop, 3-loop, and 4-loop PWR designs from each of the PWR nuclear steam supply system (NSSS) vendors (i.e., Babcock and Wilcox (B&W), Combustion Engineering (CE), and Westinghouse). A total of 1324 examinations for the components of the affected Item Nos. were conducted, with 1098 of these specifically for PWR components. The majority of PWR examinations were performed on SG welds. A relatively small number of flaws were identified during these examinations which required flaw evaluation. None of these flaws were found to be service-induced. For Item No. B2.40, examinations at two units at a single plant site identified multiple flaws exceeding the acceptance criteria of ASME Code Section XI; however, these were determined to be subsurface-embedded fabrication flaws and non-service-induced (see Table Note 1). For Item No. C1.20, two PWR units reported flaws exceeding the acceptance criteria of ASME Code, Section XI. In the first unit, a single flaw was identified, and was evaluated as an inner diameter surface imperfection. Reference [7-3] indicates that this was a spot indication with no measurable through-wall depth. This indication is therefore not considered to be service-induced but rather fabrication-related. A flaw evaluation per IWC-3600 was performed for this flaw and it was found to be acceptable for continued operation. In the second unit, multiple flaws were identified (see Table Note 2). As discussed in References [7-4] and [7-5], these flaws were most likely subsurface weld defects typical of thick vessel welds and not service-induced. A flaw evaluation for IWC-3600 was performed for these flaws and they were found to be acceptable for continued operation.

- Industry Survey Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 3 of 5 Table 7-1. Summary of Survey Results for SG Nozzle-to-Shell, Inside Radius Section, and Shell Weld Components Item No. No. of Examinations No. of Reportable Indications BWR PWR Total BWR PWR Total B2.31 0 30 30 0 0 0 B2.32 (Note 3) 0 13 13 0 0 0 B2.40 0 183 183 0 Note 1 Note 1 B3.130 0 135 135 0 0 0 C1.10 140 305 445 0 0 0 C1.20 54 319 373 0 Note 2 Note 2 C1.30 32 113 145 0 0 0 Totals 226 1098 1324 0 Notes 1 and 2 Notes 1 and 2 Notes:

1. Two PWR W-2 Loop units at a single plant reported multiple subsurface embedded fabrication flaws.

2. A single PWR W-2 Loop unit reported multiple flaws [7-4, 7-5].

3. Item No. B2.32 was evaluated in the Reference [7-1] technical basis and included in the industry survey, but is not contained in the scope of this alternative request.

- Industry Survey Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 4 of 5 Overall Industry Inspection Summary for Code Items C2.21, C2.22, and C2.32 The results of an industry survey of past inspections of SG main steam (MS) and feedwater (FW) nozzles are summarized in Reference [7-2]. Table 7-2 provides a summary of the combined survey results for Item Nos. C2.22, C2.21, and C2.32 (see Table Note 1). The results identify that SG MS and FW Nozzle-to-Shell Welds and Nozzle Inside Radius Section examinations being performed with no service-induced flaws being detected. Performing these examinations adversely impact outage activities including worker exposure, personnel safety, and radwaste. A total of 74 domestic and international BWR and PWR units responded to the survey and provided information representing all PWR plant designs currently in operation in the U.S. This included 2-loop, 3-loop, and 4-loop PWR designs from each of the PWR NSSS vendors (i.e., B&W, CE, and Westinghouse). A total of 727 examinations for Item Nos. C2.21, C2.22, and C2.32 (see Table Note 1) components were conducted, with 563 of these specifically for PWR components. The majority of the PWR examinations were performed on SG MS and FW nozzles. Only one PWR examination identified two (2) flaws that exceeded ASME Code, Section XI acceptance criteria. The flaws were linear indications of 0.3 and 0.5 in length and were detected in a MS nozzle-to-shell weld using magnetic particle examination techniques. The indications were dispositioned by light grinding (ADAMS Accession No. ML13217A093). Table 7-2. Summary of Survey Results for SG Main Steam and Feedwater Nozzle Components Plant Type Number of Units Number of Examinations Number of Reportable Indications BWR 27 164 0 PWR 47 563 2 Totals 74 727 (Note 1) 2 Notes:

1. Item No. C2.32 was evaluated in the Reference [7-2] technical basis and included in the industry survey, but is not contained in the scope of this alternative request.

- Industry Survey Proposed Alternative RA-22-0256, in Accordance with 10 CFR 50.55a(z)(1)

Page 5 of 5 References 7-1. Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906. 7-2. Technical Bases for Inspection Requirements for PWR Steam Generator Feedwater and Main Steam Nozzle-to-Shell Welds and Inside Radius Sections. EPRI, Palo Alto, CA: 2019. 3002014590. 7-3. Letter from F. A. Kearney (Exelon) to U. S. NRC, Byron Station Unit 2 90-Day Inservice Inspection Report for Interval 3, Period 3, (B2R17), dated July 29, 2013, Docket No. 50-455, ADAMS Accession Number ML13217A093. 7-4. Letter from J. M. Sorensen (NMC) to U. S. NRC, Unit 1 Inservice Inspection Summary Report, Interval 3, Period 3 Refueling Outage Dates 1-19-2001 to 2-25-2001 Cycle 20 / 05-26-99 to 02-25-2001, dated May 29, 2001, Docket Nos. 50-282 and 50-306, ADAMS Accession Number ML011550346. 7-5. Letter from J. P. Solymossy (NMC) to U. S. NRC, Response to Opportunity For Comment On Task Interface Agreement (TIA) 2003-01, Application of ASME Code Section XI, IWB-2430 Requirements Associated With Scope of Volumetric Weld Expansion at the Prairie Island Nuclear Generating Plant (Tac Nos. MB7294 and MB7295), dated April 4, 2003, Docket Nos. 50-282 and 50-306, ADAMS Accession Number ML031040553.}}