RA-22-0257, Proposed Alternative for Pressurizer Welds in Accordance with 10 CFR 50.55a(z)(1)

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

ML23048A148
Person / Time
Site:	Oconee, Mcguire, Catawba, Harris, Robinson, McGuire
Issue date:	02/17/2023
From:	Ellis K Duke Energy Carolinas, Duke Energy Progress
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
RA-22-0257
Download: ML23048A148 (1)

Text

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 10 CFR 50.55a Serial: RA-22-0257 February 17, 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 Pressurizer 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 the American Society of Mechanical Engineers (ASME) Code,Section XI for 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). Specifically, Duke Energy is requesting an alternative to the requirements of Article IWB-2500(a), Table IWB-2500-1, examination category B-B and B-D for Pressurizer Pressure-Retaining Welds and Full-Penetration Welded Nozzles.

The enclosure to this letter contains details for the proposed alternative.

Duke Energy requests NRC approval of the proposed alternative within one year of acceptance for review. Should you have any questions concerning this letter and its enclosure, please contact Ryan Treadway, Director - Nuclear Fleet Licensing at (980) 373-5873.

U.S. Nuclear Regulatory Commission RA-22-0257 Page 2 No new regulatory commitments have been made in this submittal.

j!

Kevin Ellis General Manager, Nuclear Regulatory Affairs, Policy & Emergency Preparedness

Enclosure:

Proposed Alternative for Examination of Pressurizer Pressure-Retaining Welds and Full Penetration Welded Nozzles Attachments:

1. Plant-Specific Applicability CNS 112

2. Plant-Specific Applicability MNS 1/2

3. Plant-Specific Applicability HNP

4. Plant-Specific Applicability RNP

5. Plant-Specific Applicability ONS 1/2/3

6. Results of Industry Survey

7. SI CALCULATION 2100561.302, REV. 1 (40 PAGES) "FINITE ELEMENT MODEL DEVELOPMENT AND THERMAL/MECHANICAL STRESS ANALYSIS OF BABCOCK &

WILCOX PWR PRESSURIZER SURGE NOZZLE AND BOTTOM HEAD"

8. SI CALCULATION 2100561.303, REV. 2 (40 PAGES) "DETERMINISTIC AND PROBABILISTIC FRACTURE MECHANICS ANALYSES OF OCONEE UNITS 1, 2 AND 3 BABCOCK & WILCOX PWR PRESSURIZER SURGE NOZZLE AND BOTTOM HEAD"

U.S. Nuclear Regulatory Commission RA-22-0257 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 and CNS L. Haeg, USNRC NRR Project Manager for RNP A. Donley, USNRC Senior Resident Inspector for CNS P. Boguszewski, USNRC Senior Resident Inspector for HNP C. Safouri, USNRC Senior Resident Inspector for MNS J. Nadel, USNRC Senior Resident Inspector for ONS J. Zeiler, USNRC Senior Resident Inspector for RNP

ENCLOSURE Duke Energy Carolinas, LLC Duke Energy Progress, LLC Request for Alternative RA-22-0257 Proposed Alternative for Examination of Pressurizer Pressure-Retaining Welds and Full-Penetration Welded Nozzles

Enclosure Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 1 of 23 1.0 ASME CODE COMPONENTS AFFECTED:

Code Class:

Class 1

==

Description:==

Pressurizer vessel head, shell-to-head, and nozzle-to-vessel welds Examination Category: Class 1, Category B-B, pressure-retaining welds in vessels other than reactor vessels Class 1, Category B-D, full penetration welded nozzles in vessels Item Numbers:

B2.11 - Pressurizer, shell-to-head welds, circumferential B2.12 - Pressurizer, shell-to-head welds, longitudinal B3.110 - Pressurizer, nozzle-to-vessel welds Catawba Unit 1 (CNS1)

Component ID Component Description ASME No.

1PZR-W8A Circumferential Lower Shell-to-Head B2.11 1PZR-W8E Circumferential Upper Shell-to-Head B2.11 1PZR-W9A Longitudinal Lower Shell-to-Head B2.12 1PZR-W9D Longitudinal Upper Shell-to-Head B2.12 1PZR-W1 Surge Nozzle-to-Lower Head B3.110 1PZR-W2 Spray Nozzle-to-Upper Head B3.110 1PZR-W3 Relief Nozzle-to-Upper Head B3.110 1PZR-W4A Safety Nozzle-to-Upper Head B3.110 1PZR-W4B Safety Nozzle-to-Upper Head B3.110 1PZR-W4C Safety Nozzle-to-Upper Head B3.110 Catawba Unit 2 (CNS2)

Component ID Component Description ASME No.

2PZR-W8A Circumferential Lower Shell-to-Head B2.11 2PZR-W8E Circumferential Upper Shell-to-Head B2.11 2PZR-W9A Longitudinal Lower Shell-to-Head B2.12 2PZR-W9D Longitudinal Upper Shell-to-Head B2.12 2PZR-W1 Surge Nozzle-to-Lower Head B3.110 2PZR-W2 Spray Nozzle-to-Upper Head B3.110 2PZR-W3 Safety Nozzle-to-Upper Head B3.110 2PZR-W4A Safety Nozzle-to-Upper Head B3.110 2PZR-W4B Safety Nozzle-to-Upper Head B3.110 2PZR-W4C Relief Nozzle-to-Upper Head B3.110

Enclosure Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 2 of 23 McGuire Unit 1 (MNS1)

Component ID Component Description ASME No.

1PZR-1 Circumferential Lower Shell-to-Head B2.11 1PZR-5 Circumferential Upper Shell-to-Head B2.11 1PZR-6 Longitudinal Lower Shell-to-Head B2.12 1PZR-9 Longitudinal Upper Shell-to-Head B2.12 1PZR-10 Surge Nozzle-to-Lower Head B3.110 1PZR-12 Spray Nozzle-to-Upper Head B3.110 1PZR-13 Safety Nozzle-to-Upper Head B3.110 1PZR-14 Safety Nozzle-to-Upper Head B3.110 1PZR-15 Safety Nozzle-to-Upper Head B3.110 1PZR-16 Relief Nozzle-to-Upper Head B3.110 McGuire Unit 2 (MNS2)

Component ID Component Description ASME No.

2PZR-1 Circumferential Lower Shell-to-Head B2.11 2PZR-5 Circumferential Upper Shell-to-Head B2.11 2PZR-6 Longitudinal Lower Shell-to-Head B2.12 2PZR-9 Longitudinal Upper Shell-to-Head B2.12 2PZR-10 Surge Nozzle-to-Lower Head B3.110 2PZR-12 Spray Nozzle-to-Upper Head B3.110 2PZR-13 Safety Nozzle-to-Upper Head B3.110 2PZR-14 Safety Nozzle-to-Upper Head B3.110 2PZR-15 Safety Nozzle-to-Upper Head B3.110 2PZR-16 Relief Nozzle-to-Upper Head B3.110 Oconee Unit 1 (ONS1)

Component ID Component Description ASME No.

1-PZR-WP76 Circumferential Upper Shell-to-Head B2.11 1-PZR-WP28 Circumferential Lower Shell-to-Head B2.11 1-PZR-WP1-1 Longitudinal Upper Shell-to-Head B2.12 1-PZR-WP7-1 Longitudinal Lower Shell-to-Head (Y-Z Quadrant)

B2.12 1-PZR-WP7-2 Longitudinal Lower Shell-to-Head (W-X Quadrant)

B2.12 1-PZR-WP15 Surge Nozzle-to-Lower Head B3.110 1-PZR-WP34 Spray Nozzle-to-Upper Head B3.110 1-PZR-WP33-1 Relief Nozzle-to-Upper Head (W-X Quadrant)

B3.110 1-PZR-WP33-2 Relief Nozzle-to-Upper Head (X-Y Quadrant)

B3.110 1-PZR-WP33-3 Relief Nozzle-to-Upper Head (Z-W Quadrant)

B3.110 Oconee Unit 2 (ONS2)

Component ID Component Description ASME No.

2-PZR-WP76 Circumferential Upper Shell-to-Head B2.11 2-PZR-WP28 Circumferential Lower Shell-to-Head B2.11 2-PZR-WP1-1 Longitudinal Upper Shell-to-Head B2.12

Enclosure Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 3 of 23 Oconee Unit 2 (ONS2)

Component ID Component Description ASME No.

2-PZR-WP7-1 Longitudinal Lower Shell-to-Head (Y-Z Quadrant)

B2.12 2-PZR-WP7-2 Longitudinal Lower Shell-to-Head (W-X Quadrant)

B2.12 2-PZR-WP15 Surge Nozzle-to-Lower Head B3.110 2-PZR-WP34 Spray Nozzle-to-Upper Head B3.110 2-PZR-WP33-1 Relief Nozzle-to-Upper Head (W-X Quadrant)

B3.110 2-PZR-WP33-2 Relief Nozzle-to-Upper Head (X-Y Quadrant)

B3.110 2-PZR-WP33-3 Relief Nozzle-to-Upper Head (Z-W Quadrant)

B3.110 Oconee Unit 3 (ONS3)

Component ID Component Description ASME No.

3-PZR-WP76 Circumferential Upper Shell-to-Head B2.11 3-PZR-WP28 Circumferential Lower Shell-to-Head B2.11 3-PZR-WP1-1 Longitudinal Upper Shell-to-Head B2.12 3-PZR-WP7-1 Longitudinal Lower Shell-to-Head (Y-Z Quadrant)

B2.12 3-PZR-WP7-2 Longitudinal Lower Shell-to-Head (W-X Quadrant)

B2.12 3-PZR-WP15 Surge Nozzle-to-Lower Head B3.110 3-PZR-WP34 Spray Nozzle-to-Upper Head B3.110 3-PZR-WP33-1 Relief Nozzle-to-Upper Head (W-X Quadrant)

B3.110 3-PZR-WP33-2 Relief Nozzle-to-Upper Head (X-Y Quadrant)

B3.110 3-PZR-WP33-3 Relief Nozzle-to-Upper Head (Z-W Quadrant)

B3.110 Shearon Harris Unit 1 (HNP)

Component ID Component Description ASME No.

II-PZR-01STHW-01 Circumferential Lower Shell-to-Head B2.11 II-PZR-01STHW-04 Circumferential Upper Shell-to-Head B2.11 II-PZR-01LSW-05 Longitudinal Lower Shell-to-Head B2.12 II-PZR-01LSW-07 Longitudinal Upper Shell-to-Head B2.12 II-PZR-01NTHW-08 Surge Nozzle-to-Lower Head B3.110 II-PZR-01NTHW-09 Spray Nozzle-to-Upper Head B3.110 II-PZR-01NTHW-10 Safety Nozzle-to-Upper Head B3.110 II-PZR-01NTHW-11 Safety Nozzle-to-Upper Head B3.110 II-PZR-01NTHW-12 Safety Nozzle-to-Upper Head B3.110 II-PZR-01NTHW-13 Relief Nozzle-to-Upper Head B3.110 H.B. Robinson Unit 2 (RNP)

Component ID Component Description ASME No.

103/05 Circumferential Upper Shell-to-Head B2.11 103/09 Circumferential Lower Shell-to-Head B2.11 103/01 Longitudinal Upper Shell-to-Head B2.12 103/04 Longitudinal Lower Shell-to-Head B2.12

Enclosure Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 4 of 23 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 12/14/2024 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 Categories B-B and B-D require examination of the following Item Nos.:

Item No. B2.11 - Volumetric examination of essentially 100% of the weld length for both circumferential shell-to-head welds during each inspection interval. The examination volume is shown in Figure IWB-2500-1 Item No. B2.12 - Volumetric examination of one (1) foot of all longitudinal shell-to-head welds that intersect circumferential welds during the first interval and one foot of one longitudinal shell-to-head weld that intersects a circumferential weld during successive intervals. The examination volume is shown in Figure IWB-2500-2.

Item No. B3.110 - Volumetric examination of all full penetration nozzle-to-vessel welds during each inspection interval. The examination volume is shown in Figures IWB-2500-7(a), (b), and (c).

4.0 REASON FOR REQUEST:

The Electric Power Research Institute (EPRI) performed assessments in Reference [9.1]

of the basis for the ASME Section XI examination requirements specified for the above listed ASME Section XI, Division 1 examination categories for pressurizer welds. The assessments include a survey of inspection results from 74 domestic and international

Enclosure Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 5 of 23 nuclear units and flaw tolerance evaluations using probabilistic fracture mechanics (PFM) and deterministic fracture mechanics (DFM). The Reference [9.1] report concluded that the current ASME Code,Section XI ISI examinations can be deferred with no impact to plant safety. Based on the conclusions of the EPRI report 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:

CNS1/2 For CNS1/2, Duke Energy is requesting an inspection alternative to the examination requirements of ASME Section XI, Table IWB-2500-1 for the following examination categories and item numbers:

ASME Category Item No.

Description B-B B2.11 Pressurizer, shell-to-head welds, circumferential B-B B2.12 Pressurizer, shell-to-head welds, longitudinal B-D B3.110 Pressurizer, nozzle-to-vessel welds The pressurizer welds received the required preservice examinations (PSI) prior to service followed by ISI examinations through the 2nd period of the current 4th inspection interval.

Welds within the following item numbers have been examined during the current 4th interval with no relevant indications: B2.11, B2.12, & B3.110 (CNS1) & B2.11, B2.12, &

B3.110 (CNS2). Refer to Attachment 1, Tables 1-5 and 1-6 for current examination history.

The proposed alternative is to defer the ISI examinations for these Item Nos. for the pressurizer welds at CNS1/2 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 (CNS1) and August 18, 2035 (CNS2). This equates to an extension for CNS1 of 19 years, 10 months, 10 days and 20 years for CNS2 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 Section XI, ASME Section XI, Table IWB-2500-1 for the following examination categories and item numbers:

ASME Category Item No.

Description B-B B2.11 Pressurizer, shell-to-head welds, circumferential B-B B2.12 Pressurizer, shell-to-head welds, longitudinal B-D B3.110 Pressurizer, nozzle-to-vessel welds

Enclosure Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 6 of 23 For MNS1/2, the pressurizer welds received the required PSI examinations prior to service followed by ISI examinations through the 4th inspection interval. The proposed alternative is to defer the ISI examinations for these Item Nos. for the pressurizer welds 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 Section XI, ASME Section XI, Table IWB-2500-1 for the following examination categories and item numbers:

ASME Category Item No.

Description B-B B2.11 Pressurizer, shell-to-head welds, circumferential B-B B2.12 Pressurizer, shell-to-head welds, longitudinal B-D B3.110 Pressurizer, nozzle-to-vessel welds The pressurizer welds received the required PSI examinations prior to service followed by ISI examinations through the 5th inspection interval. 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 7/14/2024 for all three Oconee units The proposed alternative is to defer the ISI examinations for these Item Nos. for the pressurizer welds at ONS1/2/3 from only 6th interval requirements, which is currently scheduled to end 7/14/2034.

HNP For HNP, Duke Energy is requesting an inspection alternative to the examination requirements of ASME Section XI, ASME Section XI, Table IWB-2500-1 for the following examination categories and item numbers:

ASME Category Item No.

Description B-B B2.11 Pressurizer, shell-to-head welds, circumferential B-B B2.12 Pressurizer, shell-to-head welds, longitudinal B-D B3.110 Pressurizer, nozzle-to-vessel welds The pressurizer welds received the required PSI examinations prior to service followed by ISI examinations through the 1st period of the current 4th inspection interval. Welds within the following item number have been examined during the current 4th interval with

Enclosure Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 7 of 23 no relevant indications: B2.11, B2.12, & B3.110. Refer to Attachment 3, Table 3-5 for current examination history.

The proposed alternative is to defer the ISI examinations for these Item Nos. for the pressurizer welds 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 Section XI, ASME Section XI, Table IWB-2500-1 for the following examination categories and item numbers:

ASME Category Item No.

Description B-B B2.11 Pressurizer, shell-to-head welds, circumferential B-B B2.12 Pressurizer, shell-to-head welds, longitudinal The pressurizer welds received the required PSI examinations prior to service followed by ISI examinations through the 5th inspection interval. All required 5th interval ISI examinations for pressurizer welds were satisfied with no relevant indications.

The proposed alternative is to defer the ISI examinations for these Item Nos. for the pressurizer welds at RNP from only 6th interval requirements, which is currently scheduled to end February 19, 2033.

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, ONS1/2/3, HNP and RNP is shown in Attachments 1 through 5.

Applicability of the Degradation Mechanism Evaluation in Reference [9.1] to the Duke Energy PWR Units An evaluation of degradation mechanisms that could potentially impact the reliability of the pressurizer welds was performed in Reference [9.1]. 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 active degradation mechanisms identified that significantly affect the long-term structural integrity of the pressurizer welds covered in this request. This observation was acknowledged by the NRC in Section 2, page 3, second paragraph of the Reference [9.13] Safety Evaluation (SE) for Salem Units 1 & 2.

The materials and operating conditions for the plants considered in this Request for

Enclosure Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 8 of 23 Alternative are similar to those in Reference [9.1] and therefore the conclusions of that Report apply to the plants in this Request for Alternative. The fatigue-related mechanisms were considered in the PFM and DFM evaluations in Reference [9.1].

As part of the technical basis in Reference [9.1], 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 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 Reference [9.1] to the Duke Energy PWR Units Duke Energy Westinghouse PWR Units (CNS 1/2, MNS 1/2, HNP and RNP)

Finite element analysis (FEA) was performed in Reference [9.1] to determine the stresses in the pressurizer welds covered in this request. The analysis was performed using representative Westinghouse 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 the Duke Energy Westinghouse PWR units (CNS1/2, MNS1/2, HNP and RNP) is demonstrated in Attachments 1 through 4 and confirms that all plant-specific requirements are met. In particular, the key geometric parameters used in the Reference [9.1] stress analysis are compared to those of the Duke Energy Westinghouse PWR units in Tables 1 and 2:

Enclosure Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 9 of 23 Table 1. Duke Energy Westinghouse Units Pressurizer Shell Dimensions Plant Shell Inside Diameter (ID) (in)

Shell/Clad Thickness (in)

Shell Ro/t Shell Ri/t EPRI Report (Table 4-4 of

[9.1])

84(1) 3.75/0.063(1) 12.2(1) 11.2 CNS1 84 3.75/0.19 12.2 11.2 CNS2 84 3.75/0.19 12.2 11.2 MNS1 84 3.75/0.19 12.2 11.2 MNS2 84 3.75/0.19 12.2 11.2 HNP 84 3.75/0.063 12.2 11.2 RNP 84 4.1/0.188 11.24 10.24 Notes:

1.

Westinghouse pressurizer dimensions, associated with model for Lower Head.

Table 2. Duke Energy Westinghouse Units Pressurizer Nozzle Dimensions Plant Surge Nzl ID (in)

Surge Nzl Thk (in)

Surge Nzl Ri/t SRV Nzl ID (in)

SRV Nzl Thk (in)

SRV Nzl Ri/t EPRI Report (Table 4-5 of [9.1])

12.44(1) 3.28(1) 1.9(1) 5.625(2) 1.19(2) 2.363(2)

CNS1 11.508 3.761 1.53 5.62 2.70 1.04 CNS2 11.508 3.761 1.53 5.62 2.70 1.04 MNS1 11.508 3.761 1.53 5.62 2.70 1.04 MNS2 11.508 3.761 1.53 5.62 2.70 1.04 HNP 11.508 3.761 1.53 5.62 2.70 1.04 RNP N/A(3)

N/A(3)

N/A(3)

N/A(3)

N/A(3)

N/A(3)

Notes:

1.

Westinghouse pressurizer nozzle dimensions, associated with model for lower head.

2.

Combustion Engineering (CE) pressurizer nozzle dimensions, associated with model for upper head.

3.

RNP Pressurizer design does not contain any B3.110 welds.

As noted by the NRC in Section 5.1, page 7, fourth paragraph of the Salem Safety Evaluation (SE) [9.12], 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] report to obtain the plant-specific stresses for each unit and component.

In the selection of the transients in Section 5 of Reference [9.1] 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, ONS1/2/3, HNP and RNP are performed at normal operating conditions. No hydrostatic testing had been performed at CNS1/2, MNS1/2, ONS1/2/3, HNP or RNP since the plants went into operation.

Enclosure Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 10 of 23 Duke Energy B&W PWR Units (ONS1/2/3)

The geometric configuration of the pressurizer used in the Reference [9.1] stress analysis, while consistent with the Westinghouse/CE pressurizer designs, is not appropriate for the B&W pressurizer design. Therefore, a plant-specific stress analysis was performed for the ONS1/2/3 pressurizer and is included as Attachment 7. provides a comparison of some ONS1/2/3 parameters to the requirements of the Reference [9.1] EPRI report for completeness; parameters superseded by the evaluation in Attachment 7 are identified in the table. The combination of the information in Attachment 5 and the evaluation in Attachment 7 indicate that all plant-specific requirements are met.

The technical approach used in the stress analysis for the ONS1/2/3 pressurizers is consistent with Section 7 of the Reference [9.1] report using the ONS1/2/3 plant-specific geometry and operating conditions. Based on the results in the Reference [9.1] EPRI report, the bottom head is controlling from a stress point of view due to the insurge/outsurge transients experienced in that region. Hence, the plant-specific stress analyses for the B&W pressurizer design were performed for the bottom head. The stress results are presented in Section 6.0 of Attachment 7. Because of the relatively complicated geometry of the B&W pressurizer design (shown in Figure 1 of Attachment 7), thirteen (13) critical stress paths were chosen for subsequent fracture mechanics evaluations, compared to two in the Reference [9.1] EPRI report. The locations of the 13 critical stress paths are provided in Figure 13 of Attachment 7, which corresponds to Figure 7-9 of the Reference [9.1] EPRI Report. Typical transient stresses for the 13 stress paths are provided in Figures 14 through 26 of Attachment 7, which correspond to Figures 7-10 and 7-11 of the Reference [9.1] EPRI report.

Applicability of the Flaw Tolerance Evaluation in Reference [9.1] to the Duke Energy PWR Units Duke Energy Westinghouse PWR Units (CNS 1/2, MNS 1/2, HNP and RNP)

Flaw tolerance evaluations were performed in Reference [9.1] consisting of probabilistic fracture mechanics (PFM) evaluations and confirmatory deterministic fracture mechanics (DFM) evaluations. The Reference [9.1] report was developed consistent with the recommendations provided in EPRIs White Paper on suggested content for PFM submittals [9.11], NRC Regulatory Guide 1.245 for PFM submittals [9.20] and the associated technical basis [9.21]. Since the configuration considered in Reference [9.1]

is consistent with the Westinghouse pressurizer design, the results of the flaw tolerance evaluation are applicable to CNS 1/2, MNS1/2, HNP and RNP. 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 2.0 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 [9.1] 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-

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Page 11 of 23 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. The NRC staff found the use PROMISE Version 2.0 acceptable in Section 3.1, page 5, fourth paragraph of the Reference [9.12] SE for Salem.

A comparison of the PSI/ISI scenarios used in the sensitivity studies performed in Reference [9.1] to those at the Duke Energy Westinghouse 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.

CNS1/2 For the CNS1/2 original pressurizers, PSI examinations have been performed followed by ISI examinations over three 10-year intervals (the units are currently in their fourth ISI interval). The PSI/ISI scenario considered is therefore PSI plus three 10-year ISI examinations to be followed by a 30-year ISI deferral (PSI+10+20+30+60).

MNS1/2 For the MNS1/2 original pressurizers, PSI examinations have been performed followed by ISI examinations over three 10-year intervals (Unti 1 is currently in its 5th ISI Interval and Unit 2 is currently in its 4th ISI interval). The PSI/ISI scenario conservatively, considered is therefore PSI plus three 10-year ISI examinations to be followed by a 30-year ISI deferral (PSI+10+20+30+60).

HNP For the HNP original pressurizer, PSI examinations have been performed followed by ISI examinations over three 10-year intervals (the unit is currently in its fourth ISI interval).

The PSI/ISI scenario considered is therefore PSI plus three 10-year ISI examinations to be followed by a 30-year ISI deferral (PSI+10+20+30+60).

RNP For the RNP original pressurizer, PSI examinations have been performed followed by ISI examinations over four 10-year intervals (the unit is currently in its fifth ISI interval). The PSI/ISI scenario considered is therefore PSI plus four 10-year ISI examinations to be followed by a 30-year ISI deferral (PSI+10+20+30+40+70).

Limiting PSI/ISI Scenario for Westinghouse Units The most limiting PSI/ISI scenario for the Duke Energy Westinghouse PWR units is (PSI+10+20+30+60). This scenario was not specifically considered in the Reference

[9.1] PFM evaluations in combination with key variables, as evaluated by the NRC in Section 4.0 (page 6) of the Reference [9.13] Safety Evaluation. Therefore, a new PFM evaluation was performed for this limiting PSI/ISI scenario using PROMISE Version 2.0, the same version used for the evaluations in the EPRI report [9.1]. The evaluations were performed for the critical Case ID from Reference [9.1] (PRSHC-BW-2C) with a combination of the most dominant parameters (stress and fracture toughness) as identified by the NRC in Section 4.0 (page 6) and Section 10 (page 19) of Reference

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[9.13]. Since all welds under consideration are shell welds, a flaw density of 1.0 was used in the evaluation. This flaw density value was found acceptable by the NRC in Section 9.6 of Reference [9.13]. A fracture toughness of 200 ksiin with a standard deviation of 5 ksiin was used, as recommended by the NRC in Section 10 (page 19) of Reference [9.13]. A stress multiplier of 1.8 was used in the evaluation. This stress multiplier was conservatively chosen such that probability of rupture or leakage will be close the acceptance criteria of 1.0E-06 after 80 years. As discussed above, a conservative stress multiplier of 1.0 can be applied to the pressurizer components of the Duke Energy plants and therefore the stress multiplier of 1.8 used in the evaluation is very conservative. The results of the evaluation are presented in Table 3.

Table 3. Sensitivity to Combined Effects of Fracture Toughness, Stress, and Flaw Density for 80 Years for the Duke Energy Westinghouse Pressurizer Welds (Case ID PRSHC-BW-2C from Reference [9.1])

Time (year)

Probability per Year for Combined Case KIC = 200 ksiin.,

SD = 5 ksiin.

Stress Multiplier = 2.1 Flaw Density = 1 PSI+10+20+30+60 Rupture Leak 10 3.90E-07 1.00E-08 20 3.55E-07 5.00E-09 30 2.40E-07 3.33E-09 40 1.80E-07 2.50E-09 50 1.44E-07 2.00E-09 60 1.23E-07 1.67E-09 70 1.06E-07 1.43E-09 80 9.25E-08 1.25E-09 The plant-specific PFM evaluation for the limiting Duke Energy Westinghouse PWR pressurizer PSI/ISI scenario presented in Table 3 indicates 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 multiplier applied in Tables 3 is greater than the ratio of R/t of the Duke Energy plants shown in Tables 1 and 2 relative to the that of the model in the EPRI report and therefore the analysis in Table 3 is conservative. It should be noted that the evaluation incorporates conservative assumptions 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.

In the PFM evaluations in Reference [9.1], the PVRUF initial flaw size distribution was used. This distribution is applicable to thick vessels and not to relatively thin vessels like pressurizers. This issue was raised by the NRC in RAI No. 4 in Reference [9.14]. In response to this RAI, various initial flaw size distributions were used in a sensitivity study

[9.15] which showed that regardless of which distribution was used, the conclusions of

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Page 13 of 23 Reference [9.1] remain the same. This was found acceptable by the NRC in Section 9.1, page 15, last paragraph of the SE for Salem [9.13].

The DFM evaluation in Table 8-4 of Reference [9.1] provides verification of the above PFM results for the Duke Energy Westinghouse PWR units by 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.

Duke Energy B&W PWR Units (ONS1/2/3)

Plant-specific DFM and PFM evaluations were performed for ONS1/2/3 pressurizers using the results of the stress analyses in Attachment 7. The DFM and PFM evaluations are presented in Attachment 8.

The technical approach used in the DFM evaluation for ONS1/2/3 pressurizers is consistent with Section 8.2 in the Reference [9.1] EPRI report. The design inputs used in the DFM evaluation are summarized in Table 1 of Attachment 8. An initial flaw size of 5.2% of the wall thickness was assumed, equivalent to the most conservative ASME Code,Section XI acceptance standard for these components. The ASME Code,Section XI, Appendix A, Paragraph A-4300 fatigue crack growth (FCG) law was used in the evaluation using the through-wall stress distributions from the stress analyses in. In addition, the weld residual stress from Figure 8-1 in the Reference [9.1]

EPRI Report and the 30ksi clad residual stress discussed in Section 8.2.2.4 of the Reference [9.1] EPRI report were considered in the evaluation. The fracture mechanics models identified in Section 8.2.2.4 of the Reference [9.1] EPRI report were used to determine the length of time for the postulated initial flaw to grow to a depth of 80% of the wall thickness (assumed to equate to leakage in this evaluation) or the depth at which the allowable toughness (upper shelf value of KIC equal to 106 ksiinch reduced by a structural factor of 2 for primary stresses and 1.0 for secondary stresses) was reached, whichever was less.

The results of the DFM evaluation for the ONS1/2/3 pressurizer configuration is summarized in Table 4 of Attachment 8, which shows that for the DFM evaluation the period required for hypothetical postulated flaws to leak are very long (in excess of 200 years). This indicates that the ONS1/2/3 pressurizer welds are very flaw tolerant.

Because the DFM evaluation considered hypothetical postulated flaws, structural factors of 2.0 on primary loads and 1.0 on secondary loads, consistent with ASME Code,Section XI, Appendix G, were applied.

The PFM evaluations were performed consistent with the approach described in Section 8.3 of the Reference [9.1] EPRI report using PROMISE, Version 2.0. The design inputs used for the PFM evaluation are shown in Table 5 of Attachment 8. For the ONS1/2/3 original pressurizers, PSI examinations have been performed followed by ISI examinations over four 10-year intervals (the units are currently in their fifth ISI interval).

The PSI/ISI scenario considered is therefore PSI plus four 10-year ISI examinations to be followed by a 30-year ISI deferral (PSI+10+20+30+40+70).

Stress and fracture toughness which were identified as the key variables in the PFM evaluation in Reference [9.1]. As such, three sensitivity studies were performed as part of the PFM as follows:

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Page 14 of 23 1.

The fracture toughness was decreased to determine the minimum fracture toughness that will meet the acceptance criteria of 1.0x10-6.

2.

The stresses were increased to determine the maximum stress multiplier that will meet the acceptance criteria of 1.0x10-6.

3.

A sensitivity study of the combined effects of the fracture toughness and stress.

Inspection history coverage for all the welds under consideration for ONS1/2/3 are provided in Attachment 5. For Item Nos. B2.11 and B2.12, inspection coverage is greater than 90% (essentially 100%) for all welds. However, for Items B3.110 some welds have limited coverage. The minimum coverage for ONS1 is 25.8%, for ONS2 Unit 2 is 25.2%, and for ONS3 is 30.0%. A sensitivity study is performed with the limiting minimum coverage of 25.2% for ONS2. Evaluations were performed using this limiting coverage to determine the probabilities of rupture and leakage for the plant-specific inspection scenarios of (PSI+10+10+30+40+70) using the same input parameters as in Table 6 in Attachment 8. For comparison, evaluations were also performed for the current ASME Code,Section XI mandated 10-year inspection interval of (PSI+10+20+30+40+50+60+70).

The results of the PFM evaluation are presented in Table 6 of Attachment 8 for the ONS1/2/3 plant-specific inspection history. As shown in this table, the probabilities of rupture and leakage are all below the acceptance criteria of 1.0x10-6 after 80 years of plant operation by three orders of magnitude.

The results of the sensitivity studies are presented in Tables 7 through 9 of Attachment

8. From Table 7 of Attachment 8, the fracture toughness can be as low as 72 ksiin before the acceptance criterion of 1.0x10-6 is reached after 80 years of operation. From Table 8 of Attachment 8, a stress multiplier of 1.4 can be applied to all the stresses considered in the evaluation before the acceptance criterion is reached. Table 9 of Attachment shows that by applying a stress multiplier of 1.1 and reducing the fracture toughness to 80 ksiin, the probabilities of rupture and leakage are all below the acceptance criterion of 1.0x10-6 after 80 years of plant operation. These sensitivity studies demonstrate the additional margins that are inherent in the PFM evaluation.

The results of the sensitivity study on coverage are presented in Table 10 of Attachment

8. As shown in this table, considering the most limiting coverage for Item No. B3.110 and the ONS1/2/3 PSI/ISI scenario, the probabilities of rupture and leakage are below the acceptance criteria of 1.0x10-6 after 80 years of operation by three orders of magnitude.

Furthermore, when the probabilities of rupture and leakage for the alternative inspection schedule are compared to the present ASME Code,Section XI inspection schedule, there is no difference. This indicates that there is no change in risk from the current ASME Code,Section XI schedule to that of the alternative inspection schedule.

Inspection History As described in Section 8.3.4.1 of Reference [9.1], PSI refers to the superset of the examinations required by ASME Code,Section III during fabrication and required by ASME Code,Section XI prior to service. The Section III

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Page 15 of 23 fabrication examinations required for these components were robust, and any Section XI preservice examinations further contributed to thorough initial examinations.

CNS1/2 Inspection history for CNS1/2 (including examinations performed to date, examination findings, inspection coverage, and relief requests) is presented in Attachment 1. As shown in the attachment, some of the welds have limited exam coverage, with the minimum coverage being 48.90%. Examination coverage greater than 37.2% is acceptable per Section 10 of the Salem Safety Evaluation [9.12]. As shown in, 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 2. As shown in the attachment, some of the welds have limited exam coverage, with the minimum coverage being 73.60%. Examination coverage greater than 37.2% is acceptable per Section 10 of the Salem Safety Evaluation [9.12]. As shown in, relevant indications that exceeded the ASME Code,Section XI acceptance standards were found acceptable. A flaw evaluation for MNS1 weld 1PZR-1 (Circumferential Lower Shell-to-Head, B2.11) on the acceptability of the five detected circumferential subsurface indications was determined to be acceptable for continued service in accordance with IWB-3600 and the flaws are demonstrated acceptable for the intended service life of the vessel [9.18 & 9.19]. Re-examination per IWB-2420(b) was not required per ASME Code Case N-526. Previously recorded indications have remained the same and no new indications were identified during the most recent 4th Interval exam.

HNP Inspection history for HNP (including examinations performed to date, examination findings, inspection coverage, and relief requests) is presented in Attachment 3. As shown in the attachment, some of the welds have limited exam coverage, with the minimum coverage being 59.10%. Examination coverage greater than 37.2% is acceptable per Section 10 of the Salem Safety Evaluation [9.12]. As shown in, 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 4. As shown in the attachment, all welds have examinations coverage greater than 90%

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Page 16 of 23 (essentially 100%). As shown in Attachment 4, 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 5. As shown in the attachment, some of the welds have limited exam coverage, with the minimum coverage being 25.20%. This limited coverage was obtained using conventional UT techniques during the 4th Inspection Interval. This minimum coverage was evaluated under the Applicability of the Flaw Tolerance Evaluation in Reference

[9.1] to the Duke Energy PWR Units section above and found acceptable. Subsequently, for the 5th Inspection Interval, these B3.110 weld locations were modeled and performed using phased array UT techniques with examination coverage greater than 37.2% which was acceptable per Section 10 of the Salem Safety Evaluation [9.12]. Finally, as shown in Attachment 5, no flaws that exceeded the ASME Code,Section XI acceptance standards were identified during any 4th or 5th Interval examinations.

Industry Survey The inspection history for these components as obtained from an industry survey is presented in Attachment 6. The results of the survey indicate that these components are very flaw tolerant.

Conclusion It is concluded that the pressurizer pressure-retaining welds and full penetration welded nozzles are very flaw tolerant. PFM and DFM evaluations performed as part of the technical basis report [9.1], 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, ONS1/2/3, HNP and RNP is demonstrated in Attachments 1 through

5. The requested ISI deferrals provide an acceptable level of quality and safety in lieu of the current ASME 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 5 show the examination history for the pressurizer welds examined in the two most recent 10-year inspection intervals.

In addition to the required PSI examinations for these pressurizer welds, all Duke Energy units have performed multiple ISI examinations through the current 10-year inspection interval at each plant.

No flaws that exceeded the ASME Code,Section XI acceptance standards were identified during any examinations, as shown in Attachments 1 through 5.

Some examinations listed in Attachments 1, 2, 3 and 4 for CNS1/2, MNS1/2, and HNP have limited examination coverage (less than 90%). Coverage for all units completed during the most recent ISI examination was greater than 37.2%, which was determined

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Page 17 of 23 to be acceptable per Section 10 of the Salem Safety Evaluation [9.12]. Plant-specific evaluations have been performed to show that the probabilities of rupture and leakage for the ISI scenarios in this Request for Alternative for ONS1/2/3 are similar to those corresponding to performing the regular Section XI inspections every 10 years. This is consistent with Section 8.3.5 of Reference 9.1, which discusses 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 9.2, 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 remainder of the 4th (for MNS2) and 5th (for MNS1) 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.

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7.0 PRECEDENTS

The following submittal has been made by PSEG Nuclear to provide relief from the ASME Code,Section XI, Examination Category B-B (Item Nos. B2.11 and B2.12) volumetric examinations based on the Reference 9.1 technical basis report:

Letter from Paul R. Duke, Jr. (PSEG Nuclear) to USNRC, Proposed Alternative for Examination of ASME Section XI, Examination Category B-B, Item Number B2.11 and B2.12, dated August 5, 2020, ADAMS Accession No. ML20218A587

[9.12].

The USNRC issued a safety evaluation of the PSEG Nuclear request for alternative on April 12, 2021.

Letter from James G. Danna (USNRC) to Eric. Carr (PSEG Nuclear), Salem Generating Station Unit Nos. 1 and 2 - Authorization and Safety Evaluation for Alternative Request No. SC-I4R-200 (EPID L-2020-LLR-0103), dated April 12, 2021, ADAMS Accession No. ML20218A587 [9.13].

The following submittal has been made by Constellation (Formally known as Exelon Generation Company, LLC) to provide relief from the ASME Code,Section XI, Examination Category B-B (Item Nos. B2.11 and B2.12) and Examination Category B-D (Item No. B3.110) volumetric examinations based on the Reference 9.1 technical basis report:

Letter from D. Gudger (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds, dated May 12, 2021, ADAMS Accession No. ML21133A297 [9.16].

The USNRC issued a issued verbal authorization followed by a safety evaluation of the proposed alternative I4R-15 for Braidwood Station Units 1 and 2 and IR4-21 for Byron Station Units 1 and 2 on November 10, 2022.

Verbal Authorization by the Office of Nuclear Reactor Regulation Proposed Alternative I4R-15 for Braidwood Station Units 1 and 2; Proposed Alternative I4R-21 for Byron Station Units 1 and 2 Constellation Energy Generation Docket Nos. 50-456 50-457 50-454 50-455 April 15, 2022, ADAMS Accession No. ML22105A072 [9.17].

Letter from Scott P. Wall to (USNRC) to David P. Rhoades (Constellation Energy Generation), Braidwood Station, Units 1 and 2 and Byron Station, Unit Nos. 1 and 2 - Proposed Alternative to the Requirements of the American Society of Mechanical Engineers Boiler & Pressure Vessel Code (EPIDS L-2021-LLR-0035 and L-2021-LLR-0036), dated November 20, 2022, ADAMS Accession No. ML22307A246.

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Page 19 of 23 The USNRC issued a safety evaluation of the Constellation request for alternative on January 3, 2023.

Letter from Hipolito J. Gonzalez (USNRC) to Constellation Energy Generation, LLC, Calvert Cliffs Nuclear Power Plant, Units 1 and 2 - Authorization and Safety Evaluation for Alternative Request No. ISI-05-016 (EPID L-2021-LLR-0036), dated January 3, 2023, ADAMS Accession No. ML22195A025.

The following is a list of other Relief Requests and other precedents related to inspections of pressurizer welds and components:

Letter from M. G. Kowal (NRC) to M. A. Balduzi (Entergy Nuclear Operations, Inc.), Indian Point Nuclear Generating Unit No. 2 - Relief Request No. RR-01 (TAC No. MD4695), dated September 5, 2007, ADAMS Accession No. ML072130487.

Letter from T. L. Tate (NRC) to Vice President, Operations (Entergy Nuclear Operations, Inc.), Indian Point Nuclear Generating Unit No. 2 - Safety Evaluation for Relief Request No. IP2-ISI-RR-01 (CAC No. MF082), dated September 14, 2016, ADAMS Accession No. ML16179A178.

Letter from H. K. Chemoff (NRC) to D. A. Heacock (Dominion Nuclear Connecticut, Inc.), Millstone Power Station Unit No. 3 - Issuance of Relief Request IR-2-51 through IR-2-60 Regarding Second 10-Year Interval Inservice Inspection Program Plan (TAC Nos. ME3809 through ME3818), dated April 26, 2011, ADAMS Accession No. ML110691154.

Letter from R. L. Emch (NRC) to J. B. Beasley Jr. (Southern Nuclear Operating Company, Inc.), Second Ten-Year Interval Inservice Inspection Program Plan Requests for Relief 13, 14, 15, 21 and 33 for Vogtle Electric Generating Plant Units 1 and 2 (TAC No. MB0603 and MB0604), dated June 20, 2001, ADAMS Accession No. ML011640178.

Letter from N. DiFrancesco (NRC) to M. J. Pacilio (Exelon Nuclear), Braidwood Station Units 1 and 2 - Relief from Requirements of the ASME Code for the Third 10-Year Interval of Inservice Inspection (TAC Nos. ME9748 and ME9749), dated January 30, 2013, ADAMS Accession No. ML13016A515.

Letter from E. C. Marinos (NRC) to D. Jamil (Duke Power Company LLC)),

Catawba Nuclear Station, Unit 1 - Request for Relief 05-CN-004, Limited Weld Examinations During End-of-Cycle 15 Refueling Outage (TAC Nos. MC8337, MC9171, MC9172, MC9173, MC9174, MC9175, MC9176, MC9177, MC9178, and MC9179), dated September 25, 2006, ADAMS Accession No. ML062390020.

Letter from J. Boska (NRC) to K. Henderson (Duke Energy Carolinas, LLC)),

Catawba Nuclear Station, Units 1 and 2 - Proposed Relief Request 11-CN-001 for the Third 10-Year Inservice Inspection Interval (TAC Nos. ME7277, ME7278, ME7279, ME7280, ME7281, ME7282, AND ME7283), dated August 20, 2012, ADAMS Accession No. ML12228A723.

Letter from R. J. Pascarelli (NRC) to K. Henderson (Duke Energy Carolinas, LLC)), Catawba Nuclear Station, Units 1 and 2 - Proposed Relief Request 14-CN-001, American Society of Mechanical Engineers (ASME)Section XI Volumetric Examination Requirements (TAC Nos. MF3527 AND MF3528), dated October 30, 2014, ADAMS Accession No. ML14295A532.

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Page 20 of 23 Letter from R. T. Repko (Duke Energy Carolinas, LLC) to NRC, Duke Energy Carolinas, LLC (Duke Energy), McGuire Nuclear Station Units 1 and 2, Docket Nos. 50-369 and 50-370, Relief Request Serial # 11-MN-001, Limited Weld Examinations for Refueling Outage 1EOC20 and 2EOC19, dated September 21, 2011, ADAMS Accession No. ML11279A035.

Letter from J. A. Price (Dominion Nuclear Connecticut, Inc.) to NRC, Dominion Nuclear Connecticut, Inc., Millstone Power Station Unit 3, ASME Section XI Inservice Inspection Program, Relief Requests for Limited Coverage Examinations Performed in the Second 10-Year Inspection Interval, dated April 19, 2010, ADAMS Accession No. ML101130187.

Letter from D. H. Corlett (Progress Energy) to NRC, Shearon Harris Nuclear Power Plant, Unit No. 1, Docket No. 50-400/License No. NPF-63, Second Ten Year Interval Inservice Inspection Program - Final Documentation Including Requests for Relief in Accordance with 10 CFR 50.55a, dated February 5, 2009, ADAMS Accession No. ML090540055.

Letter from D. H. Corlett (Progress Energy) to NRC, Shearon Harris Nuclear Power Plant, Unit No. 1, Docket No. 50-400/Renewed License No. NPF-63, Response to Request for Additional information Regarding Relief Requests 2R1-018, 2R1-019, 2R1-020, 2R1-021, 2R1-022, 2R2-009, 2R2-010, AND 2R2-011 for the Second Ten Year Interval Inspection Program (TAC Nos. ME0608, ME0609, ME0610, ME0166, ME0612, ME0613, ME0614, AND ME0615), dated September 24, 2009, ADAMS Accession No. ML092740063.

In addition, other studies have been performed by the industry to extend the inspection interval for various components and have been accepted by the NRC.

Based on studies presented in Reference 9.3, the NRC approved extending PWR reactor vessel nozzle-to-shell welds from 10 to 20 years in Reference 9.4.

Based on work performed in BWRVIP-108 [9.5] and BWRVIP-241 [9.7], the NRC approved the reduction of BWR vessel 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 [9.6] and [9.8]. The work performed in BWRVIP-108 and BWRVIP-241 provided the technical basis for ASME Code Case N-702 [9.9],

which has been conditionally approved by the NRC in Revision 19 of Regulatory Guide 1.147 [9.10].

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

Enclosure Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 21 of 23 ID Inner diameter ISI Inservice Inspection MIC Microbiologically influenced corrosion MS Main Steam NPS Nominal pipe size NRC Nuclear Regulatory Commission NSSS Nuclear steam supply system Nzl Nozzle OD Outside diameter PFM Probabilistic fracture mechanics PSI Preservice inspection PWR Pressurized Water Reactor PZR Pressurizer SCC Stress corrosion cracking SRV Safety Relief Valve Thk Thickness WEC Westinghouse Electric Company

9.0 REFERENCES

9.1 Technical Bases for Inspection Requirements for PWR Pressurizer Head, Shell-to-Head, and Nozzle-to-Vessel Welds, EPRI, Palo Alto, CA: 2019. 3002015905.

9.2 American Society of Mechanical Engineers, Risk-Based Inspection: Development of Guidelines, Volume 2-Part 1 and Volume 2-Part 2, Light Water Reactor (LWR)

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

9.3 B. A. Bishop, C. Boggess, N. Palm, Risk-Informed extension of the Reactor Vessel In-Service Inspection Interval, WCAP-16168-NP-A, Rev. 3, October 2011.

9.4 US NRC, Revised Safety Evaluation by the Office of Nuclear Reactor Regulation; Topical Report WCAP-16168-NP-A, Revision 2, Risk-Informed Extension of the Reactor Vessel In-service Inspection Interval, Pressurized Water Reactor Owners Group, Project No. 694, July 26, 2011, ADAMS Accession No. ML111600303.

9.5 BWRVIP-108

BWR Vessels and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA 2002. 1003557.

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

9.7 BWRVIP-241

BWR Vessels and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA 2010. 1021005.

9.8 US NRC, Safety Evaluation of Proprietary EPRI Report, BWR Vessel and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling

Enclosure Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 22 of 23 Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii (BWRVIP-241),

April 19, 2013, ADAMS Accession Nos. ML13071A240 and ML13071A233.

9.9 Code Case N-702, Alternate Requirements for Boiling Water Reactor (BWR)

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

9.10 U. S. NRC Regulatory Guide 1.147, Revision 19, Inservice Inspection Code Case Acceptability, ASME Code Section XI, Division 1, dated October 2019.

9.11 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.

9.12 Letter from Paul R. Duke, Jr. (PSEG Nuclear) to USNRC, Proposed Alternative for Examination of ASME Section XI, Examination Category B-B, Item Number B2.11 and B2.12, dated August 5, 2020, ADAMS Accession No. ML20218A587.

9.13 Letter from James G. Danna (USNRC) to Eric Carr (PSEG Nuclear), Salem Generating Station Unit Nos. 1 and 2 - Authorization and Safety Evaluation for Alternative Request No. SC-I4R-200 (EPID L-2020-LLR-0103), dated June 10, 2021, ADAMS Accession No. ML21145A189.

9.14 Letter from James Kim (USNRC) to Paul R. Duke, Jr. (PSEG Nuclear), Requests for Additional Information Regarding Salem Generating Station Units Nos. 1 and 2 Regarding Alternative for Examination of ASME Section XI, Category B-B, Item Number B2.11 and B2.12, EPID L-2020-LRR-0103, dated February 11, 2021, ADAMS Accession No. ML21043A144.

9.15 Letter from Paul R. Duke, Jr. (PSEG Nuclear) to USNRC, Response to Request for Additional Information for Proposed Alternative for Examination of ASME Section XI, Examination Category B-B, Item Number B2.11 and B2.12, dated April 12, 2021, ADAMS Accession No. ML21102A024.

9.16 Letter from D. Gudger (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds, dated May 12, 2021, ADAMS Accession No. ML21133A297.

9.17 Verbal Authorization by the Office of Nuclear Reactor Regulation Proposed Alternative I4R-15 for Braidwood Station Units 1 and 2; Proposed Alternative I4R-21 for Byron Station Units 1 and 2 Constellation Energy Generation Docket Nos.

50-456 50-457 50-454 50-455 April 15, 2022, ADAMS Accession No. ML22105A072.

9.18 McGuire Nuclear Station, Unit 1, Inservice Inspection Report, End of Cycle 17 Refueling Outage, dated January 9, 2006, ADAMS Accession No. ML060180504.

9.19 Letter from Evangelos C. Marinos (USNRC) to G.R. Peterson (Duke Power),

McGuire Nuclear Station, Unit 1, Third 10-Year Inservice Inspection Interval Pressurizer Lower Head to Shell Weld Flaw Evaluation (TAC No. MD0245),

dated January 5, 2007, ADAMS Accession No. 063610248.

9.20 USNRC Regulatory Guide 1.245, Revision 0, Preparing Probabilistic Fracture Mechanics Submittals, January 2022.

Enclosure Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 23 of 23 9.21 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 10 ATTACHMENT 1 PLANT-SPECIFIC APPLICABILITY CNS1/2

- CNS1/2 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 2 of 10 Section 9 of Reference 1-1 provides requirements that must be demonstrated in order to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for CNS1/2 is provided in Table 1-1 and indicates that all plant-specific requirements are met. Therefore, the results and conclusions of the EPRI report is applicable to CNS1/2.

Table 1-1 Applicability of Reference 9.1 Representative Analyses to CNS1/2 Pressurizer Surge Nozzle and Bottom Head Welds (Item Nos. B2.11, B2.12, and B3.110)

Category Requirement from Reference 1-1 Applicability to CNS1/2 General Requirements The plant-specific pressurizer general transients and cycles must be bounded by those shown in Table 5-6 for a 60-year operating life. It should be noted that the number of cycles were extrapolated to 80 years in the evaluations.

As shown in Table 1-3, the Catawba Units 1 & 2 general transients are bounded by the transients listed in Table 5-6 of Reference 1-1.

The materials of the pressurizer shell and nozzles must be low alloy ferritic steels which conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

The Catawba Units 1 & 2 nozzles are fabricated of SA-508, Class 2 material, and the pressurizer shell/heads are fabricated from SA-533, Grade A, Class 2 material.

Both materials conform to the requirements of ASME Code,Section XI, Division 1, Appendix G, Paragraph G-2110.

Specific Requirements The plant-specific pressurizer surge nozzle and bottom head weld configurations must conform to those shown in Figure 1-1 (Item No. B2.11),

Figure 1-2 (Item No. 2.12), and Figures 1-3 and 1-4 (Item No. B3.110) of Reference 1-1.

The Catawba Units 1 & 2 weld configurations are shown in Figures 1-1 and 1-2 and show conformance with the figures shown in Reference 1-1.

The plant-specific dimensions of the pressurizer shell and the surge nozzle must be within the range of values listed in Table 9-1 of Reference 1-1.

As shown in Table 1-2, the Catawba Units 1 & 2 pressurizer shell and surge nozzle dimensions are within the range of values listed in Table 9-1 of Reference 1-1.

The plant-specific Insurge/Outsurge transient definitions (temperature difference between the pressurizer shell and the pressurizer surge As shown in Table 1-4, the Catawba Units 1 & 2 Insurge/Outsurge transients are bounded by the

- CNS1/2 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 3 of 10 Category Requirement from Reference 1-1 Applicability to CNS1/2 nozzle fluid temperature and associated number of cycles) must be bounded by those shown in Table 5-10 for a Westinghouse/CE plant, or Table 5-11 for a B&W plant of Reference 1-1.

transients listed in Table 5-10 of Reference 1-1.

Pressurizer Upper Head Welds (Item Nos. B2.11, B2.12, and B3.110)

Category Requirement from Reference 1-1 Applicability to CNS1/2 General Requirements The plant-specific pressurizer general transients and cycles must be bounded by those shown in Table 5-6 for a 60-year operating life. It should be noted that the number of cycles were extrapolated to 80 years in the evaluations.

As shown in Table 1-3, the Catawba Units 1 & 2 general transients are bounded by the transients listed in Table 5-6 of Reference 1-1 The materials of the pressurizer shell and nozzles must be low alloy ferritic steels which conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

The Catawba Units 1 & 2 nozzles are fabricated of SA-508, Class 2 material, and the pressurizer shell/heads are fabricated from SA-533, Grade A, Class 2 material.

Both materials conform to the requirements of ASME Code,Section XI, Division 1, Appendix G, Paragraph G-2110.

Specific Requirements The plant-specific pressurizer upper head weld configurations must conform to those shown in Figure 1-1 (Item No. B2.11), Figure 1-2 (Item No. B2.12) and Figures 1-3 and 1-4 (Item No.

B3.110) of Reference 1-1.

The Catawba Units 1 & 2 weld configurations are shown in Figures 1-1, 1-3 and 1-4, and show conformance with the figures shown in Reference 1-

1.

The plant-specific dimensions of the pressurizer shell and the upper head nozzles must be within the range of values listed in Table 9-1 of Reference 1-1.

As shown in Table 1-2, the Catawba Units 1 & 2 pressurizer shell and upper head nozzle dimensions are within the range of values listed in Table 9-1 of Reference 1-1 Table 1-2

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Page 4 of 10 Range of Geometric Parameters for which the Evaluation is Applicable in Comparison with CNS1/2 Component Geometric Parameter For a Westinghouse Plant CNS1/2 Dimensions Pressurizer Shell Inside Diameter (in)

Must be between 80 and 88 84 [1-3]

Surge Nozzle NPS of piping or component (e.g.,

reducer) attached to nozzle safe-end (in) (1)

Must be between 12 and 18 14

[1-10, 1-11]

Safety/Relief Nozzle NPS of piping or component (e.g.,

reducer) attached to nozzle safe-end (in) (1)

Must be between 4 and 8 6

[1-10, 1-11]

Spray Nozzle NPS of piping or component (e.g.,

reducer) attached to nozzle safe-end (in) (1)

Must be between 4 and 6 4

[1-10, 1-11]

Note:

(1) Depending on the plant-specific configuration, the NPS of the piping (or component) attached directly to the nozzle (i.e., no safe-end) or to a safe-end extension may be used.

Table 1-3 Comparison of CNS1/2 General Transients to Requirements in Reference 1-1 Transient Number of Cycles for 60 Years from Table 5-6 of Reference 1-1 CNS1/2 60-Year Projection Heatup /

Cooldown 300 167/159(1)

Loss of Load (Large Step Load Decrease, Loss of Power, Loss of Flow, Reactor Trip) 360 70/71(2)

Notes:

1. Heatup/Cooldown = RCS Heatup and RCS Cooldown from Tables 45 and 49 of [1-8].

2. Loss of Load = Reactor Trip (large & small deltaP) and Loss of Power =

Loss/Pwr/Blackout+NatCirc from Tables 45 and 49 of [1-8].

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Page 5 of 10 Table 1-4 Comparison of CNS1/2 Insurge/Outsurge Temperature Differences and Numbers of Cycles With those in Reference 1-1 T (oF)(1) 60-Year No. of Cycles From Table 5-10 of Reference 1-1 (For Westinghouse and CE Plants)

CNS1/2 Cycles Projected to 60 Years of Operation

[1-9]

330 600 0

320 3,000 212 103 1,500 0

Notes:

1. T is the temperature difference between the pressurizer fluid temperature and the fluid temperature in the surge nozzle.

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Page 6 of 10 Table 1-5 CNS1 Inspection History Item No.

Component ID Exam Date Interval/Period/

Outage Exam Results Coverage Relief Request B2.11 1PZR-W8A 11/15/2003 2nd/3rd/C1R14 NRI 96.50%

N/A B2.11 1PZR-W8A 5/14/2014 3rd/3rd/C1R21 NRI 98.80%

N/A B2.11 1PZR-W8E 4/30/2011 3rd/2nd/C1R19 NRI 92.10%

N/A B2.11 1PZR-W8E 10/20/2021 4th/2nd/C1R26 NRI 91.59%

N/A B2.12 1PZR-W9A 11/15/2003 2nd/3rd/C1R14 NRI 100%

N/A B2.12 1PZR-W9A 5/14/2014 3rd/3rd/C1R21 NRI 100%

N/A B2.12 1PZR-W9D 4/30/2011 3rd/2nd/C1R19 NRI 100%

N/A B2.12 1PZR-W9D 10/20/2021 4th/2nd/C1R26 NRI 48.90%

B3.110 1PZR-W1 5/17/2005 2nd/3rd/C1R15 NRI 77.20%

05-CN-004**

B3.110 1PZR-W1 4/28/2011 3rd/2nd/C1R19 NRI 81.40%

14-CN-001^

B3.110 1PZR-W2 5/7/2008 3rd/1st/C1R17 NRI 81.70%

11-CN-001^^

B3.110 1PZR-W2 5/1/2017 4th/1st/C1R23 NRI 81.70%

B3.110 1PZR-W3 5/7/2008 3rd/1st/C1R17 NRI 81.20%

11-CN-001^^

B3.110 1PZR-W3 5/1/2017 4th/1st/C1R23 NRI 81.20%

B3.110 1PZR-W4A 5/26/2005 2nd/3rd/C1R15 NRI 79.20%

05-CN-004**

B3.110 1PZR-W4A 4/29/2011 3rd/2nd/C1R19 NRI 78.70%

14-CN-001^

B3.110 1PZR-W4B 5/26/2005 2nd/3rd/C1R15 NRI 79.20%

05-CN-004**

B3.110 1PZR-W4B 4/29/2011 3rd/2nd/C1R19 NRI 78.70%

14-CN-001^

B3.110 1PZR-W4C 5/26/2005 2nd/3rd/C1R15 NRI 79.20%

14-CN-001^

B3.110 1PZR-W4C 4/29/2011 3rd/2nd/C1R19 NRI 78.70%

05-CN-004**

• Pending Relief Request to be submitted by the end of the 4th Interval.

• • NRC SER via ADAMS Accession Number ML062390020.

^NRC SER via ADAMS Accession Number ML14295A532.

^^NRC SER via ADAMS Accession Number ML12228A723.

Table 1-6. CNS2 Inspection History Item No.

Component ID Exam Date Interval/Period/

Outage Exam Results Coverage Relief Request B2.11 2PZR-W8A 9/23/2001 2nd/2nd/C2R11 NRI

> 90%

N/A B2.11 2PZR-W8A 9/28/2010 3rd/2nd/C2R17 NRI

> 90%

N/A B2.11 2PZR-W8E 9/16/2007 3rd/1st/C2R15 NRI 96.50%

N/A B2.11 2PZR-W8E 3/27/2018 4th/1st/C2R22 NRI 96.50%

N/A B2.12 2PZR-W9A 9/23/2001 2nd/2nd/C2R11 NRI 100%

N/A B2.12 2PZR-W9A 9/28/2010 3rd/2nd/C2R17 NRI 100%

N/A B2.12 2PZR-W9D 9/30/2013 3rd/3rd/C2R19 NRI 100%

N/A B2.12 2PZR-W9D 3/27/2018 4th/1st/C2R22 NRI 100%

N/A

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Page 7 of 10 Table 1-6. CNS2 Inspection History Item No.

Component ID Exam Date Interval/Period/

Outage Exam Results Coverage Relief Request B3.110 2PZR-W1 9/25/2010 3rd/2nd/C2R17 NRI 81.20%

14-CN-001**

B3.110 2PZR-W1 4/6/2021 4th/2nd/C2R24 NRI 81.30%

B3.110 2PZR-W2 9/16/2007 3rd/1st/C2R15 NRI 81.70%

08-CN-001^

B3.110 2PZR-W2 3/23/2018 4th/1st/C2R22 NRI 93.25%

N/A B3.110 2PZR-W3 9/24/2004 2nd/3rd/C2R13 NRI 81.20%

05-CN-003^^

B3.110 2PZR-W3 3/16/2012 3rd/2nd/C2R18 NRI 81.20%

14-CN-001**

B3.110 2PZR-W4A 9/24/2004 2nd/3rd/C2R13 NRI 81.20%

05-CN-003^^

B3.110 2PZR-W4A 3/16/2012 3rd/2nd/C2R18 NRI 81.20%

14-CN-001**

B3.110 2PZR-W4B 9/24/2004 2nd/3rd/C2R13 NRI 81.20%

05-CN-003^^

B3.110 2PZR-W4B 3/16/2012 3rd/2nd/C2R18 NRI 81.20%

14-CN-001**

B3.110 2PZR-W4C 9/16/2007 3rd/1st/C2R15 NRI 81.20%

08-CN-001^

B3.110 2PZR-W4C 3/23/2018 4th/1st/C2R22 NRI 94%

N/A

• Pending Relief Request to be submitted by the end of the 4th Interval.

• • NRC SER via ADAMS Accession Number ML14295A532.

^NRC SER via ADAMS Accession Number ML092570541.

^^NRC SER via ADAMS Accession Number ML053550370.

Figure 1-1. CNS1/2 Pressurizer Vessel [1-3)]

- CNS1/2 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 8 of 10 Figure 1-2. CNS1/2 Pressurizer Surge Nozzle Safe End Configuration and Geometry

[1-4, 1-7]

- CNS1/2 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 9 of 10 Figure 1-3. CNS1/2 Pressurizer Spray Nozzle [1-5]

Figure 1-4. CNS1/2 Pressurizer SRV Nozzle [1-6]

- CNS1/2 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 10 of 10 References 1-1.

Technical Bases for Inspection requirements for Pressurizer Vessel Head, Shell-to-Head and Nozzle-to-Vessel Welds. EPRI, Palo Alto, CA: 2019. 3002015905.

1-2.

Westinghouse Electric Corporation Drawing No. EDSK 379605B, Pressurizer (84 Series) - Upper Head and Shell Complex, Revision 2.

1-3.

Westinghouse Electric Corporation Drawing No. 1101J22, Pressurizer 1800 Cu. Ft

[50.96] Cu. M. - General Arrangement, Revision 3.

1-4.

Westinghouse Electric Corporation Drawing No. EDSK 379429B, Pressurizer (84 Series) - Surge Nozzle Safe End Configuration, Revision 2.

1-5.

Westinghouse Electric Corporation Drawing No. EDSK 379445B, Pressurizer (84 Series) - Spray Nozzle Detail (Fab HD), Revision 3.

1-6.

Westinghouse Electric Corporation Drawing No. EDSK 379443B, Pressurizer (84 Series) - Safety & Relief Noz Det (Fab HD), Revision 2.

1-7.

Westinghouse Electric Corporation Drawing No. EDSK 379442B, Pressurizer (84 Series) - Surge Nozzle Detail (Fab HD), Revision 2.

1-8.

CNC 1206.02-45-0031, SI Calculation No. FP-CNS-308 - Catawba SI:FatiguePro 4 Baseline Analysis Startup through 5/2/2020 (U1) and 3/30/2021 (U2), Revision 0.

1-9.

SI Calculation No. 2100561.301, Pressurizer insurge/outsurge transients, Revision 0.

1-10. CNS Drawing CN-1553-1.1, Flow Diagram, Reactor Coolant System (NC), Revision

22.

1-11. CNS Drawing CN-2553-1.1, Flow Diagram, Reactor Coolant System (NC), Revision

20.

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Page 1 of 11 ATTACHMENT 2 PLANT-SPECIFIC APPLICABILITY MNS1/2

- MNS1/2 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 2 of 11 Section 9 of Reference 2-1 provides requirements that must be demonstrated in order to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for MNS1/2 is provided in Table 2-1 and indicates that all plant-specific requirements are met. Therefore, the results and conclusions of the EPRI report is applicable to MNS1/2.

Table 2-1 Applicability of Reference 9.1 Representative Analyses to MNS1/2 Pressurizer Surge Nozzle and Bottom Head Welds (Item Nos. B2.11, B2.12, and B3.110)

Category Requirement from Reference 2-1 Applicability to MNS1/2 General Requirements The plant-specific pressurizer general transients and cycles must be bounded by those shown in Table 5-6 for a 60-year operating life. It should be noted that the number of cycles were extrapolated to 80 years in the evaluations.

As shown in Table 2-3, the McGuire Units 1 & 2 transients are bounded by the transients listed in Table 5-6 of Reference 2-1.

The materials of the pressurizer shell and nozzles must be low alloy ferritic steels which conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

The McGuire Units 1 & 2 nozzles are fabricated of SA-508, Class 2A material, and the pressurizer shell/heads are fabricated from SA-533, Grade A, Class 2 material.

Both materials conform to the requirements of ASME Code,Section XI, Division 1, Appendix G, Paragraph G-2110.

Specific Requirements The plant-specific pressurizer surge nozzle and bottom head weld configurations must conform to those shown in Figure 1-1 (Item No. B2.11),

Figure 1-2 (Item No. 2.12), and Figures 1-3 and 1-4 (Item No. B3.110) of Reference 2-1.

The McGuire Units 1 & 2 weld configurations are shown in Figures 2-1 and 2-2, and show conformance with the figures shown in Reference 2-1.

The plant-specific dimensions of the pressurizer shell and the surge nozzle must be within the range of values listed in Table 9-1 of Reference 2-1.

As shown in Table 2-2, the McGuire Units 1 & 2 pressurizer shell and surge nozzle dimensions are within the range of values listed in Table 9-1 of Reference 2-1.

The plant-specific Insurge/Outsurge transient definitions (temperature difference between the pressurizer shell and the pressurizer surge As shown in Table 2-4, the McGuire Units 1 & 2 Insurge/Outsurge transients are bounded by the

- MNS1/2 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 3 of 11 Category Requirement from Reference 2-1 Applicability to MNS1/2 nozzle fluid temperature and associated number of cycles) must be bounded by those shown in Table 5-10 for a Westinghouse/CE plant, or Table 5-11 for a B&W plant of Reference 2-1.

transients listed in Table 5-10 of Reference 2-1.

Pressurizer Upper Head Welds (Item Nos. B2.11, B2.12, and B3.110)

Category Requirement from Reference 2-1 Applicability to MNS1/2 General Requirements The plant-specific pressurizer general transients and cycles must be bounded by those shown in Table 5-6 for a 60-year operating life. It should be noted that the number of cycles were extrapolated to 80 years in the evaluations.

As shown in Table 2-3, the McGuire Units 1 & 2 general transients are bounded by the transients listed in Table 5-6 of Reference 2-1.

The materials of the pressurizer shell and nozzles must be low alloy ferritic steels which conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

The McGuire Units 1 & 2 nozzles are fabricated of SA-508, Class 2A material, and the pressurizer shell/heads are fabricated from SA-533, Grade A, Class 2 material.

Both materials conform to the requirements of ASME Code,Section XI, Division 1, Appendix G, Paragraph G-2110 Specific Requirements The plant-specific pressurizer upper head weld configurations must conform to those shown in Figure 1-1 (Item No. B2.11), Figure 1-2 (Item No. B2.12) and Figures 1-3 and 1-4 (Item No.

B3.110) of Reference 2-1.

The McGuire Units 1 & 2 weld configurations are shown in Figures 2-1, 2-3 and 2-4, and show conformance with the figures shown in Reference 2-

1.

The plant-specific dimensions of the pressurizer shell and the upper head nozzles must be within the range of values listed in Table 9-1 of Reference 2-1.

As shown in Table 2-2, the McGuire Units 1 & 2 pressurizer shell and upper head nozzle dimensions are within the range of values listed in Table 9-1 of Reference 2-1.

Table 2-2

- MNS1/2 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 4 of 11 Range of Geometric Parameters for which the Evaluation is Applicable in Comparison with MNS1/2 Component Geometric Parameter For a Westinghouse Plant MNS1/2 Dimensions Pressurizer Shell Inside Diameter (in)

Must be between 80 and 88 84 [2-7]

Surge Nozzle NPS of piping or component (e.g.,

reducer) attached to nozzle safe-end (in) (1)

Must be between 12 and 18 14 [2-8]

Safety/Relief Nozzle NPS of piping or component (e.g.,

reducer) attached to nozzle safe-end (in) (1)

Must be between 4 and 8 6 [2-8]

Spray Nozzle NPS of piping or component (e.g.,

reducer) attached to nozzle safe-end (in) (1)

Must be between 4 and 6 4 [2-8]

Note:

(1) Depending on the plant-specific configuration, the NPS of the piping (or component) attached directly to the nozzle (i.e., no safe-end) or to a safe-end extension may be used.

Table 2-3 Comparison of MNS1/2 General Transients to Requirements in Reference 2-1 Transient Number of Cycles for 60 Years from Table 5-6 of Reference 2-1 MNS1/2 60-Year Projection Heatup /

Cooldown 300 92/81(1)

Loss of Load (Large Step Load Decrease, Loss of Power, Loss of Flow, Reactor Trip) 360 105/80(2)

Notes:

1. Heatup/Cooldown = RCS Heatup and RCS Cooldown from Tables 5-1 and 5-12 of [2-11].

2. Loss of Load = Reactor Trip (large & small deltaP) and Loss of Power =

Loss/Pwr/Blackout+NatCirc from Tables 5-1 and 5-12 of [2-11].

- MNS1/2 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 5 of 11 Table 2-4 Comparison of MNS1/2 Insurge/Outsurge Temperature Differences and Numbers of Cycles With those in Reference 2-1 T (oF) (1) 60-Year No. of Cycles From Table 5-10 of Reference 2-1 (For Westinghouse and CE Plants)

MNS1/2 Cycles Projected to 60 Years of Operation

[2-12]

330 600 0

320 3,000 261 103 1,500 0

Notes:

1. T is the temperature difference between the pressurizer fluid temperature and the fluid temperature in the surge nozzle.

- MNS1/2 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 6 of 11 Table 2-5 MNS1 Inspection History Item No.

Component ID Exam Date Interval/Period/

Outage Exam Results Coverage Relief Request B2.11 1PZR-1 10/5/2005 3rd/2nd/M1R17 RI**

93.70%

N/A B2.11 1PZR-1 3/26/2016 4th/2nd/M1R24 RI**

100%

N/A B2.11 1PZR-5 10/8/2005 3rd/2nd/M1R17 RI***

93.70%

N/A B2.11 1PZR-5 9/21/2014 4th/1st/M1R23 NRI***

92.60%

N/A B2.12 1PZR-6 10/5/2005 3rd/2nd/M1R17 NRI 100%

N/A B2.12 1PZR-6 3/26/2016 4th/2nd/M1R24 NRI 100%

N/A B2.12 1PZR-9 10/8/2005 3rd/2nd/M1R17 NRI 100%

N/A B2.12 1PZR-9 9/28/2020 4th/3rd/M1R27 NRI 100%

N/A B3.110 1PZR-10 3/15/2007 3rd/2nd/M1R18 NRI 81.20%

08-MN-001^

B3.110 1PZR-10 3/30/2019 4th/3rd/M1R26 NRI 81.30%

B3.110 1PZR-12 3/14/2004 3rd/1st/M1R16 NRI 73.60%

05-MN-001^^

B3.110 1PZR-12 9/28/2020 4th/3rd/M1R27 NRI 91.40%

N/A B3.110 1PZR-13 9/25/2008 3rd/2nd/M1R19 NRI 91.925%

N/A B3.110 1PZR-13 9/29/2017 4th/2nd/M1R25 NRI 97.75%

N/A B3.110 1PZR-14 9/25/2008 3rd/2nd/M1R19 NRI 91.925%

N/A B3.110 1PZR-14 9/29/2017 4th/2nd/M1R25 NRI 97.75%

N/A B3.110 1PZR-15 3/14/2004 3rd/1st/M1R16 NRI 73.60%

05-MN-001^^

B3.110 1PZR-15 9/28/2020 4th/3rd/M1R27 NRI 91.40%

N/A B3.110 1PZR-16 3/14/2004 3rd/1st/M1R16 NRI 73.60%

05-MN-001^^

B3.110 1PZR-16 9/28/2020 4th/3rd/M1R27 NRI 91.40%

N/A

• Pending Relief Request to be submitted by the end of the 4th Interval.

• • A flaw evaluation on the acceptability of the five detected circumferential subsurface indications was determined to be acceptable for continued service in accordance with IWB-3600 and the flaws are demonstrated acceptable for the intended service life of the vessel [9.18 & 9.19]. Re-examination per IWB-2420(b) was not required per ASME Code Case N-526. Previously recorded indications have remained the same and no new indications were identified during the most recent 4th Interval exam.

• • • Two subsurface indications were identified during the 3rd Interval exam were found to be acceptable per IWB-3500. Previously recorded indications were identified below recordable criteria during the most recent 4th Interval exam.

^NRC SER via ADAMS Accession Number ML091050008.

^^NRC SER via ADAMS Accession Number ML061530387.

- MNS1/2 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 7 of 11 Table 2-6 MNS2 Inspection History Item No.

Component ID Exam Date Interval/Period/

Outage Exam Results Coverage Relief Request B2.11 2PZR-1 9/27/2012 3rd/3rd/M2R21 NRI 98.8%

N/A B2.11 2PZR-1 9/24/2021 4th/3rd/M2R27 NRI 98%

N/A B2.11 2PZR-5 3/2/2011 3rd/2nd/M2R20 RI**

93.90%

N/A B2.11 2PZR-5 4/3/2017 4th/1st/M2R24 RI**

92.60%

N/A B2.12 2PZR-6 9/27/2012 3rd/3rd/M2R21 NRI 100%

N/A B2.12 2PZR-6 9/24/2021 3rd/3rd/M2R27 NRI 100%

N/A B2.12 2PZR-9 3/2/2011 3rd/2nd/M2R20 NRI 100%

N/A B2.12 2PZR-9 4/3/2017 4th/1st/M2R24 NRI 100%

N/A B3.110 2PZR-10 9/23/2006 3rd/1st/M2R17 NRI 81.20%

08-MN-001^

B3.110 2PZR-10 9/26/2018 4th/2nd/M2R25 NRI 81.20%

B3.110 2PZR-12 9/18/2006 3rd/1st/M2R17 NRI 81.70%

08-MN-001^

B3.110 2PZR-12 9/21/2021 4th/3rd/M2R27 NRI 94.90%

N/A B3.110 2PZR-13 9/7/2009 3rd/2nd/M2R19 NRI 78.70%

11-MN-001^^

B3.110 2PZR-13 9/21/2021 4th/3rd/M2R27 NRI 91.10%

N/A B3.110 2PZR-14 9/7/2009 3rd/2nd/M2R19 NRI 78.70%

11-MN-001^^

B3.110 2PZR-14 9/21/2021 4th/3rd/M2R27 NRI 91.10%

N/A B3.110 2PZR-15 9/7/2009 3rd/2nd/M2R19 NRI 78.70%

11-MN-001^^

B3.110 2PZR-15 9/21/2021 4th/3rd/M2R27 NRI 91.10%

N/A B3.110 2PZR-16 9/18/2006 3rd/1st/M2R17 NRI 81.20%

08-MN-001^

B3.110 2PZR-16 9/21/2021 4th/3rd/M2R27 NRI 91.10%

N/A

• Pending Relief Request to be submitted by the end of the 4th Interval.

• • Slag indications were found acceptable and no changes in the dimensions of the indications were noted in both the 3rd and 4th Interval exams.

^NRC SER via ADAMS Accession Number ML091050008.

^^NRC SER via ADAMS Accession Number ML12250A401.

- MNS1/2 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 8 of 11 Figure 2-1. MNS1/2 Pressurizer Vessel [2-8]

- MNS1/2 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 9 of 11 Figure 2-2. MNS1/2 Pressurizer Surge Nozzle Safe End Configuration and Geometry

[2-5, 2-6]

- MNS1/2 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 10 of 11 Figure 2-3. MNS1/2 Pressurizer Spray Nozzle [2-9]

Figure 2-4. MNS1/2 Pressurizer SRV Nozzle [2-10]

- MNS1/2 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 11 of 11 References 2-1.

Technical Bases for Inspection requirements for Pressurizer Vessel Head, Shell-to-Head and Nozzle-to-Vessel Welds. EPRI, Palo Alto, CA: 2019. 3002015905.

2-2.

Model D Series 84 Pressurizer Stress Report - Surge Nozzle Analysis. Westinghouse Electric Corporation, Tampa, Florida: 1981. MCM 1201.01-0362.

2-3.

Model D Series 84 Pressurizer Stress Report - Spray Nozzle Analysis. Westinghouse Electric Corporation, Tampa, Florida: 1981. MCM 1201.01-0361.

2-4.

Model D Series 84 Pressurizer Stress Report - Safety and Relief Nozzle Analysis.

Westinghouse Electric Corporation, Tampa, Florida: 1981. MCM 1201.01-0365.

2-5.

Westinghouse Electric Corporation Drawing No. EDSK 379442B, Pressurizer (84 Series) - Surge Nozzle Detail (Fab HD), Revision 2.

2-6.

Westinghouse Electric Corporation Drawing No. EDSK 379430B, Pressurizer (84 Series) - Surge Nozzle Safe End Configuration, Revision 2.

2-7.

Westinghouse Electric Corporation Drawing No. 6522D80, 1800 Cu. Ft. Pressurizer (6 Relief & Safety) - As Built Comparison of Dimensions, MCM 1201.01-0170 Rev. 1.

2-8.

Westinghouse Electric Corporation Drawing No. 6523D20, 1800 Cu. Ft. Pressurizer (6 Relief & Safety) - As Built Comparison of Dimensions, MCM 2201.01-16.

2-9.

Westinghouse Electric Corporation Drawing No. 379446B, Pressurizer (84 Series) -

Spray Nozzle Detail (Fab HD), Revision 3.

2-10. Westinghouse Electric Corporation Drawing No. 379443B, Pressurizer (84 Series) -

Safety & Relief Noz. Det. (Fab HD), Revision 2.

2-11. MCC-1206.02-45-0040, McGuire SI:FatiguePro 4.0 Baseline Analysis Startup through 9/24/2017 (U1) and 9/16/2018 (U2), Revision 0.

2-12. SI Calculation No. 2100561.301, Pressurizer insurge/outsurge transients, Revision 0.

- HNP Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 1 of 10 ATTACHMENT 3 PLANT-SPECIFIC APPLICABILITY HNP

- HNP Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 2 of 10 Section 9 of Reference 3-1 provides requirements that must be demonstrated in order to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for HNP is provided in Table 3-1 and indicates that all plant-specific requirements are met. Therefore, the results and conclusions of the EPRI report is applicable to HNP.

Table 3-1 Applicability of Reference 9.1 Representative Analyses to HNP Pressurizer Surge Nozzle and Bottom Head Welds (Item Nos. B2.11, B2.12, and B3.110)

Category Requirement from Reference 3-1 Applicability to HNP General Requirements The plant-specific pressurizer general transients and cycles must be bounded by those shown in Table 5-6 for a 60-year operating life. It should be noted that the number of cycles were extrapolated to 80 years in the evaluations.

As shown in Table 3-3, the Harris general transients are bounded by the transients listed in Table 5-6 of Reference 3-1.

The materials of the pressurizer shell and nozzles must be low alloy ferritic steels which conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

The Harris nozzles are fabricated of SA-508, Class 2A material, and the pressurizer shell/heads are fabricated from SA-533, Grade A, Class 2 material.

Both materials conform to the requirements of ASME Code,Section XI, Division 1, Appendix G, Paragraph G-2110.

Specific Requirements The plant-specific pressurizer surge nozzle and bottom head weld configurations must conform to those shown in Figure 1-1 (Item No. B2.11),

Figure 1-2 (Item No. 2.12), and Figures 1-3 and 1-4 (Item No. B3.110) of Reference 3-1.

The Harris weld configurations are shown in Figures 3-1 and 3-2, and show conformance with the figures shown in Reference 3-1.

The plant-specific dimensions of the pressurizer shell and the surge nozzle must be within the range of values listed in Table 9-1 of Reference 3-1.

As shown in Table 3-2, the Harris pressurizer shell and surge nozzle dimensions are within the range of values listed in Table 9-1 of Reference 3-1.

The plant specific Insurge/Outsurge transient definitions (temperature difference between the pressurizer shell and the pressurizer surge As shown in Table 3-4, the Harris Insurge/Outsurge transients are

- HNP Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 3 of 10 Category Requirement from Reference 3-1 Applicability to HNP nozzle fluid temperature and associated number of cycles) must be bounded by those shown in Table 5-10 for a Westinghouse/CE plant, or Table 5-11 for a B&W plant of Reference 3-1.

bounded by the transients listed in Table 5-10 of Reference 3-1.

Pressurizer Upper Head Welds (Item Nos. B2.11, B2.12, and B3.110)

Category Requirement from Reference 3-1 Applicability to HNP General Requirements The plant-specific pressurizer general transients and cycles must be bounded by those shown in Table 5-6 for a 60-year operating life. It should be noted that the number of cycles were extrapolated to 80 years in the evaluations.

As shown in Table 3-3, the Harris general transients are bounded by the transients listed in Table 5-6 of Reference 3-1.

The materials of the pressurizer shell and nozzles must be low alloy ferritic steels which conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

The Harris nozzles are fabricated of SA-508, Class 2A material, and the pressurizer shell/heads are fabricated from SA-533, Grade A, Class 2 material.

Both materials conform to the requirements of ASME Code,Section XI, Division 1, Appendix G, Paragraph G-2110.

Specific Requirements The plant-specific pressurizer upper head weld configurations must conform to those shown in Figure 1-1 (Item No. B2.11), Figure 1-2 (Item No. B2.12) and Figures 1-3 and 1-4 (Item No.

B3.110) of Reference 3-1.

The Harris weld configurations are shown in Figures 3-1, 3-3 and 3-4, and show conformance with the figures shown in Reference 3-1.

The plant-specific dimensions of the pressurizer shell and the upper head nozzles must be within the range of values listed in Table 9-1 of Reference 3-1.

As shown in Table 3-2, the Harris pressurizer shell and upper head nozzle dimensions are within the range of values listed in Table 9-1 of Reference 3-1.

- HNP Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 4 of 10 Table 3-2 Range of Geometric Parameters for which the Evaluation is Applicable in Comparison with HNP Component Geometric Parameter For a Westinghouse Plant HNP Dimensions Pressurizer Shell Inside Diameter (in)

Must be between 80 and 88 84 [3-2]

Surge Nozzle NPS of piping or component (e.g.,

reducer) attached to nozzle safe-end (in) (1)

Must be between 12 and 18 14 [3-3], [3-10]

Safety/Relief Nozzle NPS of piping or component (e.g.,

reducer) attached to nozzle safe-end (in) (1)

Must be between 4 and 8 6 [3-4], [3-10]

Spray Nozzle NPS of piping or component (e.g.,

reducer) attached to nozzle safe-end (in) (1)

Must be between 4 and 6 4 [3-5], [3-10]

Note:

(2) Depending on the plant-specific configuration, the NPS of the piping (or component) attached directly to the nozzle (i.e., no safe-end) or to a safe-end extension may be used.

Table 3-3 Comparison of HNP General Transients to Requirements in Reference 3-1 Transient Number of Cycles for 60 Years from Table 5-6 of Reference 3-1 HNP 60-Year Projection Heatup /

Cooldown 300 84(1)

Loss of Load (Large Step Load Decrease, Loss of Power, Loss of Flow, Reactor Trip) 360 3(2)

Notes:

1. Heatup/Cooldown = Plant Heatup and Plant Cooldown from Table 5-2 of [3-8].

2. Loss of Load = Loss of Load and Loss of Power = Loss of Power from Table 5-2 of [3-8]

- HNP Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 5 of 10 Table 3-4 Comparison of HNP Insurge/Outsurge Temperature Differences and Numbers of Cycles With those in Reference 3-1 T (oF) (1) 60-Year No. of Cycles From Table 5-10 of Reference 3-1 (For Westinghouse and CE Plants)

HNP Cycles Projected to 60 Years of Operation

[3-9]

330 600 0

320 3,000 2,352 103 1,500 0

Notes:

1. T is the temperature difference between the pressurizer fluid temperature and the fluid temperature in the surge nozzle.

- HNP Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 6 of 10 Table 3-5 HNP Inspection History Item No.

Component ID Exam Date Interval/Period/

Outage Exam Results Coverage Relief Request B2.11 II-PZR-01STHW-01 4/30/2003 2nd/2nd/H1R11 RI*

95%

N/A B2.11 II-PZR-01STHW-01 4/22/2015 3rd/3rd/H1R19 RI*

95%

N/A B2.11 II-PZR-01STHW-04 4/25/2009 3rd/1st/H1R15 RI**

100%

N/A B2.11 II-PZR-01STHW-04 10/24/2019 4th/1st/H1R22 NRI 100%

N/A B2.12 II-PZR-01-LSW-05 4/30/2003 2nd/2nd/H1R11 NRI 100%

N/A B2.12 II-PZR-01-LSW-05 4/22/2015 3rd/3rd/H1R19 NRI 100%

N/A B2.12 II-PZR-01-LSW-07 4/25/2009 3rd/1st/H1R15 NRI 100%

N/A B2.12 II-PZR-01-LSW-07 10/24/2019 4th/1st/H1R22 NRI 100%

N/A B3.110 II-PZR-01NTH-08 4/21/2000 2nd/1st/H1R9 NRI 70%

2R1-021^

B3.110 II-PZR-01NTH-08 5/2/2012 3rd/2nd/H1R17 NRI 59.10%

I3R-19 B3.110 II-PZR-01NTH-09 10/30/1998 2nd/1st/H1R8 NRI 67.14%

2R1-021^

B3.110 II-PZR-01NTH-09 4/30/2012 3rd/2nd/H1R17 NRI 66.31%

I3R-19^^

B3.110 II-PZR-01NTH-10 4/28/2009 3rd/1st/H1R15 NRI 67.14%

I3R-19^^

B3.110 II-PZR-01NTH-10 10/23/2019 4th/1st/H1R22 NRI 90.40%

N/A B3.110 II-PZR-01NTH-11 4/28/2009 3rd/1st/H1R15 NRI 67.14%

I3R-19^^

B3.110 II-PZR-01NTH-11 10/23/2019 4th/1st/H1R22 NRI 90.40%

N/A B3.110 II-PZR-01NTH-12 4/28/2009 3rd/1st/H1R15 NRI 67.14%

I3R-19^^

B3.110 II-PZR-01NTH-12 10/23/2019 4th/1st/H1R22 NRI 90.40%

N/A B3.110 II-PZR-01NTH-13 10/30/1998 2nd/1st/H1R8 NRI 67.14%

2R1-021^

B3.110 II-PZR-01NTH-13 5/3/2012 3rd/2nd/H1R17 NRI 69.26%

I3R-19^^

• Subsurface indications were found acceptable per IWB-3500 and no changes in the dimensions of the indications were noted in both the 3rd and 4th Interval exams.

• • Previously identified indications accepted per IWB-3500. Previously recorded indications were identified below recordable criteria during the most recent 4th Interval exam.

^NRC SER via ADAMS Accession Number ML093561419.

^^NRC SER via ADAMS Accession Number ML20080G950.

- HNP Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 7 of 10 Figure 3-1. HNP Pressurizer Vessel [3-2]

- HNP Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 8 of 10 Figure 3-2. HNP Pressurizer Surge Nozzle Safe End Configuration and Geometry

[3-3, 3-7]

- HNP Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 9 of 10 Figure 3-3. HNP Pressurizer Spray Nozzle [3-5]

Figure 3-4. HNP Pressurizer SRV Nozzle [3-4]

- HNP Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 10 of 10 References 3-1.

Technical Bases for Inspection requirements for Pressurizer Vessel Head, Shell-to-Head and Nozzle-to-Vessel Welds. EPRI, Palo Alto, CA: 2019. 3002015905.

3-2.

Model D Series 84 Pressurizer Stress Report for Carolina Power and Light Shearon Harris Unit 1. Westinghouse Electric Corporation. Pensacola, FL. Doc. No. 1364-052710. Revision 003.

3-3.

Westinghouse Electric Corporation Drawing No. EDSK 380419B, Pressurizer (1400 Cu Ft) - Pressurizer, Revision 2.

3-4.

Westinghouse Electric Corporation Drawing No. EDSK 380430B, Pressurizer (84 Series) - Surge Nozzle Safe End Configuration, Revision 0.

3-5.

Westinghouse Electric Corporation Drawing No. EDSK 380763B, Pressurizer (84 Series) - Safety & Relief Noz. Det. (Fab HD), Revision 0.

3-6.

Westinghouse Electric Corporation Drawing No. EDSK 380764B, Pressurizer (84 Series) - Spray Nozzle Detail (Fab HD), Revision 0.

3-7.

Westinghouse Electric Corporation Drawing No. EDSK 379442B, Pressurizer (84 Series) - Surge Nozzle Detail (Fab HD), Revision 2.

3-8.

SI Calculation No. FP-HNP-315, Harris SI:FatiguePro 4.0 Baseline Analysis Startup through October 12, 2019, Revision 0.

3-9.

SI Calculation No. 2100561.301, Pressurizer insurge/outsurge transients, Revision 0.

3-10. HNP Drawing CAR-2165-G-0801, Flow Diagram Reactor Coolant System Sheet 2, Revision 26.

- RNP Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 1 of 7 ATTACHMENT 4 PLANT-SPECIFIC APPLICABILITY RNP

- RNP Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 2 of 7 Section 9 of Reference 4-1 provides requirements that must be demonstrated in order to apply the representative stress and flaw tolerance analyses to a specific plant. Plant-specific evaluation of these requirements for RNP is provided in Table 4-1 and indicates that all plant-specific requirements are met. Therefore, the results and conclusions of the EPRI report is applicable to RNP.

Table 4-1 Applicability of Reference 9.1 Representative Analyses to RNP Pressurizer Bottom Head Welds (Item Nos. B2.11 and B2.12)

Category Requirement from Reference 4-1 Applicability to RNP General Requirements The plant-specific pressurizer general transients and cycles must be bounded by those shown in Table 5-6 for a 60-year operating life. It should be noted that the number of cycles were extrapolated to 80 years in the evaluations.

As shown in Table 4-3, the Robinson general transients are bounded by the transients listed in Table 5-6 of Reference 4-1.

The materials of the pressurizer shell and nozzles must be low alloy ferritic steels which conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

Nozzle requirements are N/A for Robinson (not seeking relief for B3.110). The Robinson pressurizer shell/heads are fabricated from SA-302, Grade B.

The shell/head material conforms to the requirements of ASME Code,Section XI, Division 1, Appendix G, Paragraph G-2110.

Specific Requirements The plant-specific pressurizer surge nozzle and bottom head weld configurations must conform to those shown in Figure 1-1 (Item No. B2.11) and Figure 1-2 (Item No. 2.12) of Reference 4-

1.

The Robinson surge nozzle weld configuration requirements are N/A (not seeking relief for B3.110). The Robinson pressurizer bottom head weld configurations are shown in Figures 4-1 and show conformance with the figures shown in Reference 4-

1.

The plant-specific dimensions of the pressurizer shell and the surge nozzle must be within the range of values listed in Table 9-1 of Reference 4-1.

As shown in Table 4-2, the Robinson pressurizer shell dimensions are within the range of values listed in Table 9-1 of Reference 4-1.

- RNP Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 3 of 7 Category Requirement from Reference 4-1 Applicability to RNP The Robinson surge nozzle weld configuration requirements are N/A (not seeking relief for B3.110).

The plant-specific Insurge/Outsurge transient definitions (temperature difference between the pressurizer shell and the pressurizer surge nozzle fluid temperature and associated number of cycles) must be bounded by those shown in Table 5-10 for a Westinghouse/CE plant, or Table 5-11 for a B&W plant of Reference 4-1.

As shown in Table 4-4, the Robinson Insurge/Outsurge transients are bounded by the transients listed in Table 5-10 of Reference 4-1.

Pressurizer Upper Head Welds (Item Nos. B2.11 and B2.12)

Category Requirement from Reference 4-1 Applicability to RNP General Requirements The plant-specific pressurizer general transients and cycles must be bounded by those shown in Table 5-6 for a 60-year operating life. It should be noted that the number of cycles were extrapolated to 80 years in the evaluations.

As shown in Table 4-3, the Robinson general transients are bounded by the transients listed in Table 5-6 of Reference 4-1.

The materials of the pressurizer shell and nozzles must be low alloy ferritic steels which conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

The Robinson pressurizer shell/heads are fabricated from SA-302, Grade B.

The Robinson SRV nozzle weld configuration requirements are N/A (not seeking relief for B3.110).

The shell/head material conforms to the requirements of ASME Code,Section XI, Division 1, Appendix G, Paragraph G-2110 Specific Requirements The plant-specific pressurizer upper head weld configurations must conform to those shown in Figure 1-1 (Item No. B2.11) and Figure 1-2 (Item No. B2.12) and Figure 1-3 and 1-4 (Item No. B3.110) of Reference 4-1.

The Robinson weld configurations are shown in Figures 4-1, 4-3 and 4-4, and show conformance with the figures shown in Reference 4-1.

The Robinson SRV nozzle weld configuration requirements are N/A (not seeking relief for B3.110).

- RNP Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 4 of 7 Category Requirement from Reference 4-1 Applicability to RNP The plant-specific dimensions of the pressurizer shell and the upper head nozzles must be within the range of values listed in Table 9-1 of Reference 4-1.

As shown in Table 4-2, the Robinson pressurizer shell dimensions are within the range of values listed in Table 9-1 of Reference 4-1.

The Robinson SRV nozzle weld configuration requirements are N/A (The design of Robinson Pressurizer does not contain any welds subject to B3.110 requirements).

Table 4-2 Range of Geometric Parameters for which the Evaluation is Applicable in Comparison with RNP Component Geometric Parameter For a Westinghouse Plant RNP Dimensions Pressurizer Shell Inside Diameter (in)

Must be between 80 and 88 84 [4-1]

Surge Nozzle NPS of piping or component (e.g.,

reducer) attached to nozzle safe-end (in) (1)

Must be between 12 and 18 N/A(2)

Safety/Relief Nozzle NPS of piping or component (e.g.,

reducer) attached to nozzle safe-end (in) (1)

Must be between 4 and 8 N/A(2)

Spray Nozzle NPS of piping or component (e.g.,

reducer) attached to nozzle safe-end (in) (1)

Must be between 4 and 6 N/A(2)

Note:

(1) Depending on the plant-specific configuration, the NPS of the piping (or component) attached directly to the nozzle (i.e., no safe-end) or to a safe-end extension may be used.

(2) RNP is not seeking relief for B3.110 locations. The design of the RNP Pressurizer does not contain any welds subject to B3.110 requirements.

- RNP Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 5 of 7 Table 4-3 Comparison of RNP General Transients to Requirements in Reference 4-1 Transient Number of Cycles for 60 Years from Reference 4-1 RNP 60-Year Projection Heatup /

Cooldown 300 137/126(1)

Loss of Load (Large Step Load Decrease, Loss of Power, Loss of Flow, Reactor Trip) 360 246(2)

Notes:

1. Heatup/Cooldown = Plant Heatup/Plant Cooldown from [4-6] projected to 60 years.

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

Table 4-4 Comparison of RNP Insurge/Outsurge Temperature Differences and Numbers of Cycles With those in Reference 4-1 T (oF) (1) 60-Year No. of Cycles From Table 5-10 of Reference 4-1 (For Westinghouse and CE Plants)

RNP Cycles Projected to 60 Years of Operation

[4-5]

330 600 0

320 3,000 416 103 1,500 0

Notes:

1. T is the temperature difference between the pressurizer fluid temperature and the fluid temperature in the surge nozzle.

- RNP Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 6 of 7 Table 4-5 RNP Inspection History Item No.

Component ID Exam Date Interval/Period/

Outage Exam Results Coverage Relief Request B2.11 103/05 2/26/2012 4th/3rd/R2R27 NRI 98%

N/A B2.11 103/05 11/28/2022 5th/3rd/R2R33 NRI 98.5%

N/A B2.11 103/09 10/6/2008 4th/2nd/R2R25 NRI 98.90%

N/A B2.11 103/09 10/8/2018 5th/2nd/R2R31 NRI 99%

N/A B2.12 103/01 10/6/2008 4th/2nd/R2R25 NRI 92.30%

N/A B2.12 103/01 11/28/2022 5th/3rd/R2R33 NRI 100%

N/A B2.12 103/04 2/26/2012 4th/3rd/R2R27 NRI 100%

N/A B2.12 103/04 10/8/2018 5th/2nd/R2R31 NRI 91.70%

N/A Figure 4-1. RNP Pressurizer Vessel [4-2]

References 4-1.

Technical Bases for Inspection requirements for Pressurizer Vessel Head, Shell-to-Head and Nozzle-to-Vessel Welds. EPRI, Palo Alto, CA: 2019. 3002015905.

4-2.

Westinghouse Electric Corporation Drawing No. 4417D30, 1300 Cu. Ft. Pressurizer (4 Safety & Relief Nozzles) - Outline.

4-3.

Westinghouse Electric Corporation Drawing No. 681J271, 1300 Cu. Ft. Pressurizer (4 Safety & Relief Nozzles) - General Assembly & Final Machining.

- RNP Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 7 of 7 4-4.

Updated RNP transient counts.docx.

4-5.

SI Calculation No. 2100561.301, Pressurizer insurge/outsurge transients, Revision 0.

- ONS1/2/3 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 1 of 11 ATTACHMENT 5 PLANT-SPECIFIC APPLICABILITY ONS1/2/3

- ONS1/2/3 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 2 of 11 Section 9 of Reference 5-1 provides requirements that must be demonstrated in order to apply the representative stress and flaw tolerance analyses to a specific plant.

However, the pressurizer configuration for ONS1/2/3 is of B&W design and is considerably different in terms of both geometry and materials from the configuration evaluated in Reference 5-1. Therefore, in lieu of comparison to the requirements of Section 9 of Reference 5-1, plant-specific stress analyses and fracture mechanics (DFM and PFM) evaluations were performed for ONS1/2/3 pressurizers and are presented in Attachments 7 and 8, respectively.

Table 5-1 (which provides a comparison of some ONS1/2/3 parameters to the requirements of Section 9 of Reference 5-1) is provided below for completeness; parameters superseded by the evaluations in Attachments 7 and 8 are identified in the table. The combination of Table 5-1 and the evaluations in Attachments 7 and 8 indicate that all plant-specific requirements are met. Therefore, the results and conclusions of the EPRI report are applicable to ONS1/2/3.

The plant-specific geometry of the ONS1/2/3 pressurizers used in the analyses in Attachments 7 and 8 is shown in Figures 5-1 through 5-4. ONS1/2/3 thermal transients and stratification loads used in the analyses in Attachments 7 and 8 are shown in Tables 5-3 and 5-4, respectively. The inspection history for ONS1, ONS2, and ONS3 are provided in Tables 5-5, 5-6, and 5-7, respectively.

Table 5-1 Applicability of Reference 5-1 Representative Analyses to ONS1/2/3 Pressurizer Surge Nozzle and Bottom Head Welds (Item Nos. B2.11, B2.12, and B3.110)

Category Requirement from Reference 5-1 Applicability to ONS1/2/3 General Requirements The plant-specific pressurizer general transients and cycles must be bounded by those shown in Table 5-6 for a 60-year operating life. It should be noted that the number of cycles were extrapolated to 80 years in the evaluations.

As shown in Table 5-3, the ONS1/2/3 transients are bounded by the transients listed in Table 5-6 of Reference 5-1.

The materials of the pressurizer shell and nozzles must be low alloy ferritic steels which conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

Refer to evaluations in Attachments 7 and 8.

Specific Requirements The plant-specific pressurizer surge nozzle and bottom head weld configurations must conform to those shown in Figure 1-1 (Item No. B2.11),

Figure 1-2 (Item No. 2.12), and Figures 1-3 and 1-4 (Item No. B3.110) of Reference 5-1.

The ONS1/2/3 weld configurations are shown in Figures 5-1 and 5-2 and show conformance with the figures shown in Reference 5-1.

- ONS1/2/3 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 3 of 11 Category Requirement from Reference 5-1 Applicability to ONS1/2/3 The plant-specific dimensions of the pressurizer shell and the surge nozzle must be within the range of values listed in Table 9-1 of Reference 5-1.

As shown in Table 5-2, the ONS1/2/3 pressurizer shell and surge nozzle dimensions are within the range of values listed in Table 9-1 of Reference 5-1.

The plant-specific Insurge/Outsurge transient definitions (temperature difference between the pressurizer shell and the pressurizer surge nozzle fluid temperature and associated number of cycles) must be bounded by those shown in Table 5-10 for a Westinghouse/CE plant, or Table 5-11 for a B&W plant of Reference 5-1.

As shown in Table 5-4, the ONS1/2/3 transients are bounded by the transients listed in Table 5-11 of Reference 5-1.

Pressurizer Upper Head Welds (Item Nos. B2.11, B2.12, and B3.110)

Category Requirement from Reference 5-1 Applicability to ONS1/2/3 General Requirements The plant-specific pressurizer general transients and cycles must be bounded by those shown in Table 5-6 for a 60-year operating life. It should be noted that the number of cycles were extrapolated to 80 years in the evaluations.

As shown in Table 5-3, the ONS1/2/3 general transients are bounded by the transients listed in Table 5-6 of Reference 5-1.

The materials of the pressurizer shell and nozzles must be low alloy ferritic steels which conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

Refer to evaluations in Attachments 7 and 8.

Specific Requirements The plant-specific pressurizer upper head weld configurations must conform to those shown in Figure 1-1 (Item No. B2.11), Figure 1-2 (Item No. B2.12) and Figures 1-3 and 1-4 (Item No.

B3.110) of Reference 5-1.

The ONS1/2/3 weld configurations are shown in Figures 5-1, 5-3, and 5-4, and show conformance with the figures shown in Reference 5-1.

The plant-specific dimensions of the pressurizer shell and the upper head nozzles must be within the range of values listed in Table 9-1 of Reference 5-1.

As shown in Table 5-2, the ONS1/2/3 pressurizer shell and upper head nozzles are within the range of values shown in Table 9-1 of Reference 5-1.

- ONS1/2/3 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 4 of 11 Table 5-2 Range of Geometric Parameters for which the Evaluation is Applicable in Comparison with ONS1/2/3 Component Geometric Parameter For a B&W Plant ONS1/2/3 Dimensions Pressurizer Shell Inside Diameter (in)

Must be between 80 and 88 84 [5-2, 5-3, 5-4]

Surge Nozzle NPS of piping or component (e.g.,

reducer) attached to nozzle safe-end (in) (1)

Must be between 8 and 12 10

[5-10, 5-11, 5-12]

Safety/Relief Nozzle NPS of piping or component (e.g.,

reducer) attached to nozzle safe-end (in) (1)

Must be between 4 and 6 4

[5-10, 5-11, 5-12]

Spray Nozzle NPS of piping or component (e.g.,

reducer) attached to nozzle safe-end (in) (1)

Must be between 3 and 6 4

[5-10, 5-11, 5-12]

Note:

(1) Depending on the plant-specific configuration, the NPS of the piping (or component) attached directly to the nozzle (i.e., no safe-end) or to a safe-end extension may be used.

Table 5-3 ONS1/2/3 Plant-Specific Transient Cycles Transient ONS1/2/3 60-Year Projection Heatup /

Cooldown 132/134/104(1)

Loss of Load (Large Step Load Decrease, Loss of Power, Loss of Flow, Reactor Trip) 62/33/41(2)

Notes:

1. Heatup/Cooldown = RCS Heatup and RCS Cooldown from Tables 13, 14 and 15 of [5-8]

scaled down from 80 to 60 years. 134 cycles conservatively used in the DFM and PFM evaluations.

2. Loss of Load = Rx Trip No Loss of Flow from Tables 13, 14 and 15 of [5-8] scaled down from 80 to 60 years. 62 cycles conservatively used in the DFM and PFM evaluations.

- ONS1/2/3 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 5 of 11 Table 5-4 ONS1/2/3 Plant Specific Insurge/Outsurge Temperature Differences and Numbers of Cycles(1)

T (oF)(2) 60-Year No. of Cycles for Evaluation 400 0

350 0

300 720 250 360 200 1440 Note:

(1) From Reference [5-9].

(2) T is the temperature difference between the pressurizer fluid temperature and the fluid temperature in the surge nozzle.

Table 5-5 ONS1 Inspection History Item No.

Component ID Exam Date Interval/Period/

Outage Exam Results Coverage Relief Request B2.11 1-PZR-WP76 4/18/2005 4th/1st/O1R22 NRI 100%

N/A B2.11 1-PZR-WP76 11/21/2016 5th/1st/O1R29 NRI 100%

N/A B2.11 1-PZR-WP28 11/10/2012 4th/3rd/O1R27 NRI 92%

N/A B2.11 1-PZR-WP28 11/08/2022 5th/3rd/O1R32 NRI 99.35%

N/A B2.12 1-PZR-WP1-1 4/18/2005 4th/1st/O1R22 NRI 100%

N/A B2.12 1-PZR-WP1-1 11/21/2016 5th/1st/O1R29 NRI 100%

N/A B2.12 1-PZR-WP7-1 11/10/2012 4th/3rd/O1R27 NRI 90.30%

N/A B2.12 1-PZR-WP7-1 11/08/2022 5th/3rd/O1R32 NRI 100%

N/A B3.110 1-PZR-WP15 11/12/2006 4th/1st/O1R23 NRI 56.92%

07-ON-002^

B3.110 1-PZR-WP15 11/16/2016 5th/1st/O1R29 NRI 41.72%

B3.110 1-PZR-WP33-1 10/13/2006 4th/1st/O1R23 NRI 62.80%

07-ON-002^

B3.110 1-PZR-WP33-1 11/19/2016 5th/1st/O1R29 NRI 28.80%

B3.110 1-PZR-WP33-2 10/13/2006 4th/1st/O1R23 NRI 62.80%

07-ON-002^

B3.110 1-PZR-WP33-2 11/19/2016 5th/1st/O1R29 NRI 25.80%

B3.110 1-PZR-WP33-3 10/13/2006 4th/1st/O1R23 NRI 62.80%

07-ON-002^

- ONS1/2/3 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 6 of 11 Table 5-5 ONS1 Inspection History Item No.

Component ID Exam Date Interval/Period/

Outage Exam Results Coverage Relief Request B3.110 1-PZR-WP33-3 11/19/2016 5th/1st/O1R29 NRI 28.70%

B3.110 1-PZR-WP34 10/13/2006 4th/1st/O1R23 NRI 67.48%

07-ON-002^

B3.110 1-PZR-WP34 11/19/2016 5th/1st/O1R29 NRI 45.60%

• Pending Relief Request to be submitted by the end of the 4th Interval.

^NRC SER via ADAMS Accession Number ML082480215.

Table 5-6 ONS2 Inspection History Item No.

Component ID Exam Date Interval/Period/

Outage Exam Results Coverage Relief Request B2.11 2-PZR-WP76 10/28/2005 4th/1st/O2R21 NRI 100%

N/A B2.11 2-PZR-WP76 10/25/2015 5th/1st/O2R27 NRI 96.4%

N/A B2.11 2-PZR-WP28 5/13/2010 4th/2nd/O2R24 RI**

98.50%

N/A B2.11 2-PZR-WP28 11/20/2019 5th/2nd/O2R29 NRI 99.21%

N/A B2.12 2-PZR-WP1-1 10/28/2005 4th/1st/O2R27 NRI 100%

N/A B2.12 2-PZR-WP1-1 10/25/2015 5th/1st/O2R21 NRI 100%

N/A B2.12 2-PZR-WP7-1 5/13/2010 4th/2nd/O2R24 NRI 100%

N/A B2.12 2-PZR-WP7-1 11/20/2019 5th/2nd/O2R29 NRI 100%

N/A B3.110 2-PZR-WP15 5/15/2007 4th/1st/O2R22 NRI 41.70%

10-ON-002^

B3.110 2-PZR-WP15 11/7/2017 5th/1st/O2R28 NRI 68.25%

B3.110 2-PZR-WP34 5/3/2010 4th/2nd/O2R24 NRI 76.10%

11-ON-001^^

B3.110 2-PZR-WP34 11/18/2019 5th/2nd/O2R29 NRI 78.80%

B3.110 2-PZR-WP33-3 5/3/2010 4th/2nd/O2R24 NRI 71.20%

11-ON-001^^

B3.110 2-PZR-WP33-3 11/18/2019 5th/2nd/O2R29 NRI 71.40%

B3.110 2-PZR-WP33-2 10/27/2011 4th/3rd/O2R25 NRI 25.20%

12-ON-001&

B3.110 2-PZR-WP33-2 11/18/2019 5th/2nd/O2R29 NRI 71.40%

B3.110 2-PZR-WP33-1 5/3/2010 4th/2nd/O2R24 NRI 71.20%

11-ON-001^^

B3.110 2-PZR-WP33-1 11/18/2019 5th/2nd/O2R29 NRI 71.40%

• Pending Relief Request to be submitted by the end of the 4th Interval.

• • Subsurface flaws were found acceptable per IWB-3500 during the 4th Interval exam. Previously recorded indications were identified below recordable criteria during 5th Interval exam.

^NRC SER via ADAMS Accession Number ML12111A006.

^^NRC SER via ADAMS Accession Number ML12334A549.

&NRC SER via ADAMS Accession Number ML13365A023.

- ONS1/2/3 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 7 of 11 Table 5-7 ONS3 Inspection History Item No.

Component ID Exam Date Interval/Period/

Outage Exam Results Coverage Relief Request B2.11 3-PZR-WP76 11/2/2007 4th/1st/O3R23 NRI 91.1%

N/A B2.11 3-PZR-WP76 5/1/2018 5th/1st/O3R29 NRI 100%

N/A B2.11 3-PZR-WP28 5/12/2003 3rd/3rd/O3R20 NRI 97.70%

N/A B2.11 3-PZR-WP28 5/1/2014 4th/3rd/O3R27 NRI 93.80%

N/A B2.12 3-PZR-WP1-1 11/2/2007 4th/1st/O3R23 NRI 100%

N/A B2.12 3-PZR-WP1-1 5/1/2018 5th/1st/O3R29 NRI 100%

N/A B2.12 3-PZR-WP7-1 5/12/2003 3rd/3rd/O3R20 NRI 97.50%

N/A B2.12 3-PZR-WP7-1 5/1/2014 4th/3rd/O3R27 NRI 90.30%

N/A B3.110 3-PZR-WP15 11/1/2007 4th/1st/O3R23 NRI 41.70%

10-ON-002^

B3.110 3-PZR-WP15 5/3/2018 5th/1st/O3R29 NRI 67.30%

B3.110 3-PZR-WP34 11/1/2007 4th/1st/O3R23 NRI 46.10%

10-ON-002^

B3.110 3-PZR-WP34 4/30/2018 5th/1st/O3R29 NRI 75.40%

B3.110 3-PZR-WP33-3 11/1/2007 4th/1st/O3R23 NRI 30.00%

10-ON-002^

B3.110 3-PZR-WP33-3 4/30/2018 5th/1st/O3R29 NRI 68.90%

B3.110 3-PZR-WP33-2 11/1/2007 4th/1st/O3R23 NRI 30.00%

10-ON-002^

B3.110 3-PZR-WP33-2 4/30/2018 5th/1st/O3R29 NRI 70.40%

B3.110 3-PZR-WP33-1 11/1/2007 4th/1st/O3R23 NRI 30.00%

10-ON-002^

B3.110 3-PZR-WP33-1 4/30/2018 5th/1st/O3R29 NRI 61%

• Pending Relief Request to be submitted by the end of the 4th Interval.

^NRC SER via ADAMS Accession Number ML12111A006.

- ONS1/2/3 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 8 of 11 Figure 5-1. ONS1/2/3 Pressurizer Vessel [5-6]

- ONS1/2/3 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 9 of 11 Figure 5-2. ONS1/2/3 Pressurizer Surge Nozzle [5-6]

Figure 5-3. ONS1/2/3 Pressurizer Spray Nozzle [5-6]

- ONS1/2/3 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 10 of 11 Figure 5-4. ONS1/2/3 Pressurizer SRV Nozzle [5-6]

- ONS1/2/3 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 11 of 11 References 5-1.

Technical Bases for Inspection requirements for Pressurizer Vessel Head, Shell-to-Head and Nozzle-to-Vessel Welds. EPRI, Palo Alto, CA: 2019. 3002015905.

5-2.

Oconee Unit 1 Drawing ISI-OCN1-002, Pressurizer Weld Outline, Revision 2.

5-3.

Oconee Unit 2 Drawing ISI-OCN2-002, Pressurizer Weld Outline, Revision 2.

5-4.

Oconee Unit 3 Drawing ISI-OCN3-002, Pressurizer Weld Outline, Revision 1.

5-5.

Oconee Nuclear Station UFSAR Appendix 5A, Table 5-22. Pressurizer Design Data.

December 31, 2006.

5-6.

Babcock & Wilcox Company Drawing No. 129268E, Radiographic Outline, Revision 6.

Oconee Drawing No. OM 201-1878.

5-7.

OSC-8745.01, Supporting Vendor Analysis for Alloy 600 Mitigation - Oconee Nuclear Station Units 1, 2, & 3-Pressurizer Spray Overlay. SI Calculation ONS-15Q-322, Design and UT Examination of Preemptive Repairs/Replacements of Pressurizer and Hot Leg Alloy 600 Components, Revision 2.

5-8.

SI Calculation FP-ONS-304P, Oconee SI:FatiguePro 4 Baseline Analysis, Startup through 11/3/2020 (U1), 11/21/2019 (U2) and 4/24/2020 (U3), Revision 0.

5-9.

SI Calculation No. 2100561.301, Pressurizer insurge/outsurge transients, Revision 0.

5-10. ONS Drawing OFD-100A-1.2, Flow Diagram of Reactor Coolant System (Pressurizer),

Revision 31.

5-11. ONS Drawing OFD-100A-2.2, Flow Diagram of Reactor Coolant System (Pressurizer),

Revision 31.

5-12. ONS Drawing OFD-100A-3.2, Flow Diagram of Reactor Coolant System (Pressurizer),

Revision 31.

- Industry Survey Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 1 of 3 ATTACHMENT 6 RESULTS OF INDUSTRY SURVEY

- Industry Survey Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 2 of 3 Overall Industry Inspection Summary The results of an industry survey of past inspections of pressurizer welds are summarized in Reference 6-1. Table 6-1 provides a summary of the combined survey results for Item Nos. B2.11, B2.12, B2.21, B2.22 and B3.110. The results identify that pressurizer examination of the items adversely impact outage activities including worker exposure, personnel safety, and radwaste. A total of 74 domestic and international PWR units responded to the survey and provided information representing all PWR plant designs currently in operation in the U.S. This included 2-loop, 3-loop, and 4-loop PWR designs from each of the PWR nuclear steam supply system (NSSS) vendors (i.e.,

Babcock and Wilcox (B&W), Combustion Engineering (CE), and Westinghouse). A total of 1,162 examinations for the components of the affected Item Nos. were conducted on PWR pressurizer components.

A small number of flaws were identified during these examinations which required flaw evaluation. None of these flaws were found to be service induced. Out of a total of 1,162 examinations identified by the plants that responded to the survey that have been performed on the above item numbers, only four examinations (for Item No. B2.11), at two units of a single plant site, identified flaws exceeding the acceptance criteria of ASME Code,Section XI. Flaw evaluations were performed to show acceptability of these indications and follow up examinations showed no change in flaw sizes since the original inspections. No other indications were identified in any in-scope components.

- Industry Survey Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

Page 3 of 3 Table 6-1 Summary of Survey Results Item No.

No. of Examinations No. of Reportable Indications B2.11 269 4 (1)

B2.12 269 0

B2.21 4

0 B2.22 30 0

B3.110 590 0

Note:

(1) Flaw evaluations were performed to show acceptability of these indications and follow up examinations showed no change in flaw sizes since the original inspections. References 6-1.

Technical Bases for Inspection requirements for Pressurizer Vessel Head, Shell-to-Head and Nozzle-to-Vessel Welds. EPRI, Palo Alto, CA: 2019. 3002015905.

- SI Calculation 2100561.302, Rev. 1 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

ATTACHMENT 7 SI CALCULATION 2100561.302, REV. 1 (40 PAGES)

FINITE ELEMENT MODEL DEVELOPMENT AND THERMAL/MECHANICAL STRESS ANALYSIS OF BABCOCK & WILCOX PWR PRESSURIZER SURGE NOZZLE AND BOTTOM HEAD

CALCULATION PACKAGE File No.: 2100561.302 Project No.: 2100561 Quality Program Type:

Nuclear Commercial PROJECT NAME:

Duke PWR Fleet SG & Pzr Inspection Optimization Relief Requests CONTRACT NO.:

GSA # 03021365 CLIENT:

Duke Energy Corporation PLANT:

Oconee Nuclear Station, Units 1, 2 and 3 CALCULATION TITLE:

Finite Element Model Development and Thermal/Mechanical Stress Analysis of Babcock & Wilcox PWR Pressurizer Surge Nozzle and Bottom Head Document Revision Affected Pages Revision Description Project Manager Approval Signature & Date Preparer(s) &

Checker(s)

Signatures & Date 0

1 - 37 A A-2 Computer Files Initial Issue Scott Chesworth 3/28/2022 Preparer Richard Bax 3/3/2022 Checker Minghao Qin 3/3/2022 1

1 - 38 A A-2 Computer Files Incorporating Duke Energy Comments Scott Chesworth 8/5/2022 Preparer Richard Bax 8/5/2022 Checker Minghao Qin 8/5/2022

File No.: 2100561.302 Revision: 1 Page 2 of 38 F0306-01R4 Table of Contents 1.0 OBJECTIVE.............................................................................................................. 4 2.0 TECHNICAL APPROACH......................................................................................... 4 2.1 Finite Element Model..................................................................................... 4 2.2 Pressure / Thermal Stress Analyses.............................................................. 4 3.0 DESIGN INPUTS....................................................................................................... 4 4.0 ASSUMPTIONS........................................................................................................ 5 5.0 CALCULATIONS....................................................................................................... 5 5.1 Finite Element Model..................................................................................... 5 5.1.1 Material Properties......................................................................................... 6 5.2 Pressure / Thermal Stress Analysis............................................................... 6 5.2.1 Unit Internal Pressure Loading Analysis......................................................... 6 5.2.2 Thermal Heat Transfer Analyses.................................................................... 6 5.2.3 Thermal Stress Analyses............................................................................... 7 6.0 RESULTS OF ANALYSIS.......................................................................................... 7

7.0 REFERENCES

.......................................................................................................... 7 COMPUTER FILES LISTING...................................................................... A-1 List of Tables Table 1. Material Properties for Carbon Steel, C>0.3% (SA-508, Class 1 or SA-516, Grade 70)............................................................................................................ 9 Table 2. Material Properties for Stainless Steel (SA-240, Type 304).................................. 10 Table 3. Thermal Transients for Pressurizer Surge Nozzle................................................. 11 Table 4. Insurge/Outsurge Transients for Pressurizer Surge Nozzle.................................. 12

File No.: 2100561.302 Revision: 1 Page 3 of 38 F0306-01R4 List of Figures Figure 1. Modeled Dimensions........................................................................................... 13 Figure 2. Weld Locations.................................................................................................... 14 Figure 3. 3-D Finite Element Model.................................................................................... 15 Figure 4. Applied Boundary Conditions and Unit Internal Pressure.................................... 16 Figure 5. Example of Applied Thermal Boundary Conditions for Thermal Transient Analyses............................................................................................................ 17 Figure 6. Example of Applied Thermal Boundary Conditions for Insurge/Outsurge Transient Analyses............................................................................................ 18 Figure 7. Applied Mechanical Boundary Conditions for Thermal Stress Analyses.............. 19 Figure 8. Stress Contours Due to Unit Internal Pressure.................................................... 20 Figure 9. Temperature Contour Heatup/Cooldown Transient (Time = 10,494 seconds)...... 21 Figure 10. Stress Contours of Heatup/Cooldown Transient (Time = 10,494 seconds)........ 22 Figure 11. Temperature Contour Insurge/Outsurge Group 3 Transient (Time = 900 seconds)............................................................................................................ 23 Figure 12. Stress Contours of Insurge/Outsurge Group 3 Transient (Time = 900 seconds)............................................................................................................ 24 Figure 13. Path Locations.................................................................................................. 25 Figure 14. Through-Wall Stress Distribution at Path 1........................................................ 26 Figure 15. Through-Wall Stress Distribution at Path 2........................................................ 27 Figure 16. Through-Wall Stress Distribution at Path 3........................................................ 28 Figure 17. Through-Wall Stress Distribution at Path 4........................................................ 29 Figure 18. Through-Wall Stress Distribution at Path 5........................................................ 30 Figure 19. Through-Wall Stress Distribution at Path 6........................................................ 31 Figure 20. Through-Wall Stress Distribution at Path 7........................................................ 32 Figure 21. Through-Wall Stress Distribution at Path 8........................................................ 33 Figure 22. Through-Wall Stress Distribution at Path 9........................................................ 34 Figure 23. Through-Wall Stress Distribution at Path 10...................................................... 35 Figure 24. Through-Wall Stress Distribution at Path 11...................................................... 36 Figure 25. Through-Wall Stress Distribution at Path 12...................................................... 37 Figure 26. Through-Wall Stress Distribution at Path 13...................................................... 38

File No.: 2100561.302 Revision: 1 Page 4 of 38 F0306-01R4 1.0 OBJECTIVE The objective of this calculation is to develop a finite element model of a typical Babcock and Wilcox (B&W) designed pressurizer water reactor (PWR) pressurizer (PZR) surge nozzle and bottom head and determine stresses due to thermal transients and unit pressure. The through-wall stresses are extracted at the surge nozzle-to-bottom head, bottom head-to-shell weld locations and heater bundle shell thickness region to shell and bottom head welds and are stored in computer files that will be used in a separate fracture mechanics evaluation.

2.0 TECHNICAL APPROACH 2.1 Finite Element Model A finite element model (FEM) is developed using the ANSYS finite element analysis software package [1].

The FEM is a 3-dimensional (3-D) model of a typical B&W pressurizer surge nozzle, bottom head, and lower shell region. The model includes a local portion of the pressurizer lower shell and cladding (which includes the shell thickness increase at the heater bundle region), the pressurizer bottom head and cladding, and the surge nozzle and cladding.

2.2 Pressure / Thermal Stress Analyses Stress analyses are performed for thermal transients and a unit internal pressure. For thermal loads due to thermal transients, thermal analyses are performed to determine the temperature distribution time-histories for each transient. The temperature distributions are then used as input to perform stress analyses for each transient. For internal pressure, an arbitrary unit internal pressure is applied. Due to the linear elastic nature of modeling, the stress results from the unit pressure can be scaled to correspond to actual pressure values as needed. Stress results are saved for use in future evaluations and are listed in Appendix A.

3.0 DESIGN INPUTS A typical B&W PZR surge nozzle, bottom head and lower shell configuration is used as a representative component for the finite element model. The geometry of the surge nozzle and bottom head is derived from Oconee Finite Element Model of Pressurizer Surge Nozzle with Weld Overlay Repair Per Design Dimensions [2]. The dimensions of the lower shell, including the shell thickness increase at the heater bundle region is based on a B&W Pressurizer General Arrangement Drawing [3, 9]. The base metal thickness of the increased thickness region at the heater bundle region is defined as 13.563 inches in Reference [6]. However, Reference [7] indicates that the minimum wall thickness at the heater bundle region is 12.5 inches. For this evaluation, the greater wall thickness of 13.563 inches will be used since it will generate higher thermal stresses which are expected to dominate given the relatively thick component.

The general transients for analysis were previously defined in Table 5-6 Reference [4], while the typical insurge/outsurge transients for Westinghouse (W) and Combustion Engineering (CE) were defined in Table 5-9 of Reference [4]. In Reference [4], the B&W insurge/outsurge transients were derived based on applying scaling factors to the Group 3 insurge/outsurge transients of the Westinghouse/CE insurge/outsurge transients. The same approach is adopted in this calculation and therefore only the Group 3 insurge/outsurge transients for the W/CE pressurizer designs documented in Table 5-9 of Reference [4] will be evaluated.

For the thermal transient stress analyses performed herein, only the transient definitions from Table 5-6 (for the general transients) and Table 5-11 (for the B&W insurge/outsurge transients) of Reference [4]

are required. The number of cycles provided in these tables are not considered herein but will be considered in subsequent deterministic and probabilistic fracture mechanics evaluations.

File No.: 2100561.302 Revision: 1 Page 5 of 38 F0306-01R4 The welds of interest for the lower pressurizer region are identified in ASME Code,Section XI, Table IWB-2500-1 [5], and are as follows:

Item No. B2.11 (Pressurizer Shell Circumferential Weld)

Item No. B2.12 (Pressurizer Shell Longitudinal Weld)

Item No. B3.110 (Pressurizer Nozzle-to-Vessel Welds) 4.0 ASSUMPTIONS A number of assumptions are made during development of the finite element model and the thermal /

pressure stress evaluation, which are listed as follows:

The nozzle-to-lower head, lower head-to-shell and the shell welds are not specifically modeled.

The material properties between the base metals and the weld materials are similar enough that the effect of this assumption will be minimal.

Per References [3, 9] the total circumferential extent of the thickness increase for the heater bundle region is 108°. However, the circumferential extent of the full thickness of this region is not defined and is estimated to be 78°.

Heat transfer coefficients during thermal transients are assumed based on the flow condition for the inside surface of the nozzle and bottom head.

All thermal transients are assumed to start and end at a steady-state uniform temperature.

The stress-free reference temperature for thermal stress calculation is assumed to be an ambient temperature of 70°F, which is used for thermal strain calculations. This assumption is typical for stress analyses in similar components.

All outside surfaces are assumed to be fully insulated and the insulation itself is treated as perfect, with zero heat transfer capability. This assumption is typical for stress analyses in similar components.

Pressure stresses are calculated at a stress-free temperature of 70°F and do not include any thermal stress effects.

The density and Poissons ratio are assumed temperature independent.

The W/CE insurge/outsurge transient Group 3 is used and scaled based on the methodology outlined in Reference [4] to obtain the B&W insurge/outsurge transients.

5.0 CALCULATIONS 5.1 Finite Element Model A finite element model of the PZR surge nozzle, bottom head and lower shell is developed using the ANSYS finite element analysis software package [1], with dimensions shown in Figure 1.

Because of the axisymmetric nature of this configuration, a 3-D model is constructed using 3-D structural solid, SOLID185, elements. The thermal equivalent element for the thermal transient analyses is SOLID70. The weld locations are shown in Figure 2 and are based on Reference [6]. The constructed model is shown in Figure 3.

File No.: 2100561.302 Revision: 1 Page 6 of 38 F0306-01R4 5.1.1 Material Properties The Oconee pressurizer nozzles are fabricated using SA-508, Grade 1, Class 1 (Carbon Steel), while the pressurizer heads and vessels are fabricated from SA-212, Grade B (Unit 1) or SA-516, Grade 70 (Unit 2 and 3) (Carbon Steel) [2, Table 1]. A typical stainless steel (SA-240, Type 304) is assumed for all cladding material.

The material properties were obtained from the relevant tables in the 2013 Edition of ASME Code,Section II, Part D [8]. Temperature dependent material properties used in the finite element analysis are listed in Table 1 and Table 2.

5.2 Pressure / Thermal Stress Analysis 5.2.1 Unit Internal Pressure Loading Analysis A unit internal pressure of 1,000 psi is applied to the interior surfaces of the model. The resulting stresses will be scaled to the appropriate plant pressure conditions for subsequent fracture mechanics evaluations. An induced end-cap load is applied to the free end of the surge nozzle in the form of tensile axial pressures, as calculated below.

Pnozzle-cap =

2 22 =

10008.752 11.528.752 = 1,375 psi

where, Pnozzle-cap =

End cap pressure on nozzle free end (psi)

P

=

Internal pressure (psi)

ID

=

Inside diameter of nozzle free end (in)

OD

=

Outside diameter of nozzle free end (in)

Symmetric boundary conditions are applied to the axial and circumferential free ends of the pressurizer shell while axial displacement couples are applied to the free end of the surge nozzle. The applied pressure load and boundary conditions for this case are shown in Figure 4.

5.2.2 Thermal Heat Transfer Analyses Thermal transient parameters were developed in Reference [4] for use in this analysis. The thermal transients listed in Table 3 and Table 4, are applied to the interior surface nodes of the nozzle and head.

The heat transfer coefficients for the inside surface of the surge nozzle and shell are also listed in Table 3 and Table 4.

Per Section 4.0, no heat transfer coefficients or temperatures are applied to the assumed insulated outside surfaces. Figure 5 shows representative plots of the thermal loads applied for Heatup/Cooldown transient.

For the insurge and outsurge transients defined in Table 4 a pool height, temperature and heat transfer coefficient are also specified. These represents the pool of colder water that builds up in the pressurizer during the insurge and then its draining during the outsurge.

Examining Table 4 shows that the pool heat transfer coefficient is relatively low (50 Btu/hr-ft²-°F), due to the mixing process of the surge inflow with the relatively stagnant pressurizer fluid. However, in order to introduce added conservativism, the pool heat transfer will be left at the higher heat transfer value at insurge initiation for the entire length of the insurge / outsurge event. Figure 6 shows representative plots of the thermal loads applied for Insurge/Outsurge transient Group 3.

File No.: 2100561.302 Revision: 1 Page 7 of 38 F0306-01R4 An additional time of 3,600 seconds is added to the end of each transient to ensure that any lagging peak stresses are captured, followed by a steady state load step (at an arbitrary 400 seconds after the 3,600 seconds of additional time).

5.2.3 Thermal Stress Analyses Symmetric boundary conditions are applied to the axial and circumferential free ends of the pressurizer shell, while axial displacement couples are applied on the free end of the surge nozzle. The stress-free reference temperature for the thermal strain calculation is assumed to be 70°F. Figure 7 shows a plot of the boundary conditions applied for the thermal stress analyses.

6.0 RESULTS OF ANALYSIS A finite element model of a typical B&W pressurizer surge nozzle, bottom head, lower shell and heater bundle shell thickness region has been developed. The stress results due to thermal transients and unit internal pressure have been run and are stored to be used for future fracture mechanics evaluations.

Example hoop stress and axial stress contour plots for unit internal pressure are shown in Figure 8.

Representative temperature contour and stress contour plots for the Heatup/Cooldown are shown in Figure 9 and Figure 10, respectively. The instance shown in Figure 9 and Figure 10 is when the maximum stress intensity occurs. Representative temperature contour and stress contour plots for the Insurge/Outsurge Transient Group 3 at 900 second are shown in Figure 11 and Figure 12, respectively.

Figure 13 shows the path locations where stresses are extracted. Path 1 through Path 3 are chosen as representative of the nozzle-to-lower head weld. Path 4 though Path 7 are chosen as a representative of the shell-to-lower head weld and heater bundle shell thickness region-to-lower head weld. Path 8 through Path 11 are chosen as a representative of the lower shell-to-upper shell weld and heater bundle shell thickness region-to-upper shell weld. Path 12 is representative of the heater bundle shell thickness region-to-lower shell axial weld and Path 13 is representative of the heater bundle shell thickness region middle circumferential weld. The input and output files used in this evaluation are listed in Appendix A.

All stresses are extracted in a cylindrical coordinate system, about the pressurizer surge nozzle.

Through-wall stress distributions for Path 1 through Path 13 are shown in Figure 14 and Figure 26, respectively.

7.0 REFERENCES

1. ANSYS Mechanical APDL (UP20170403) and Workbench (March 31, 2017), Release 18.1, SAS IP, Inc.

2. SI Calculation No. ONS-15Q-312, Rev. 1, Finite Element Model of Pressurizer Surge Nozzle with Weld Overlay Repair Per Design Dimensions.

3. Babcock and Wilcox Drawing No. 25476, Rev. 7, Pressurizer General Arrangement, SI File No.

ANO-39Q-212.

4. Technical Bases for Inspection Requirements for PWR Pressurizer Head, Shell-to-Head, and Nozzle-to-Vessel Welds. EPRI, Palo Alto, CA: 2019. 3002015905.

5. ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2017 Edition.

6. Oconee Drawing No. ISI-OCNI-002, Rev. 2, Pressurizer Weld Outline, SI File No.

2100561.205.

File No.: 2100561.302 Revision: 1 Page 8 of 38 F0306-01R4

7. Oconee Drawing No. OM 201-1878, Rev. 6, Radiographic Outline, SI File No. 2100561.205.

8. ASME Boiler and Pressure Vessel Code,Section II Materials, Part D - Properties, 2013 Edition.

9. Letter from Austin Keller (Duke Energy) to Scott Chesworth (SI) on August 3, 2022,

Subject:

Duke Energy Inputs for SIA Calculation 2100561.302, SI File No. 2100561.205.

File No.: 2100561.302 Revision: 1 Page 9 of 38 F0306-01R4 Table 1. Material Properties for Carbon Steel, C>0.3% (SA-508, Class 1 or SA-516, Grade 70)

Temperature

(°F)

Modulus of Elasticity (E)

(106 psi)

Coefficient of Thermal Expansion ()

(10-6 in/in/°F)

Thermal Conductivity (K)

(10-4 BTU/in-s-°F)

Specific Heat (C)(4)

(BTU/lb-°F) 70 29.2 6.4 8.08 0.103 100 29.1(1) 6.5 8.03 0.106 150 28.8(1) 6.6 7.92 0.110 200 28.6 6.7 7.80 0.114 250 28.4(1) 6.8 7.64 0.117 300 28.1 6.9 7.48 0.119 350 27.9(1) 7.0 7.31 0.122 400 27.7 7.1 7.15 0.124 450 27.4(1) 7.2 6.97 0.126 500 27.1 7.3 6.81 0.128 550 26.8(1) 7.3 6.64 0.131 600 26.4 7.4 6.48 0.134 650 25.9(1) 7.5 6.32 0.136 700 25.3 7.6 6.16 0.140 Notes:

1. Linearly interpolated.

2. Density () = 0.280 lb/in3 [8, Table PRD], assumed temperature independent.

3. Poissons Ratio () = 0.3 [8, Table PRD], assumed temperature independent.

4. Calculated per Note 1 of Table TCD [8].

5. SA-508, Class 1 has been reclassified as SA-508, Grade 1, Class 1 in Reference [8].

File No.: 2100561.302 Revision: 1 Page 10 of 38 F0306-01R4 Table 2. Material Properties for Stainless Steel (SA-240, Type 304)

Temperature

(°F)

Modulus of Elasticity (E)

(106 psi)

Coefficient of Thermal Expansion ()

(10-6 in/in/°F)

Thermal Conductivity (K)

(10-4 BTU/in-s-°F)

Specific Heat (C)(4)

(BTU/lb-°F) 70 28.3 8.5 1.99 0.114 100 28.1(1) 8.6 2.01 0.114 150 27.8(1) 8.8 2.08 0.117 200 27.5 8.9 2.15 0.119 250 27.3(1) 9.1 2.22 0.121 300 27.0 9.2 2.27 0.122 350 26.7(1) 9.4 2.34 0.124 400 26.4 9.5 2.41 0.126 450 26.2(1) 9.6 2.45 0.127 500 25.9 9.7 2.52 0.129 550 25.6(1) 9.8 2.57 0.129 600 25.3 9.9 2.62 0.130 650 25.1(1) 9.9 2.69 0.131 700 24.8 10.0 2.73 0.132 Notes:

1. Linearly interpolated

2. Density () = 0.290 lb/in3 [8, Table PRD], assumed temperature independent.

3. Poissons Ratio () = 0.31 [8, Table PRD], assumed temperature independent.

4. Calculated per Note 1 of Table TCD [8].

File No.: 2100561.302 Revision: 1 Page 11 of 38 F0306-01R4 Table 3. Thermal Transients for Pressurizer Surge Nozzle Transient

Time, sec

Tpzr,

°F

Tnoz,

°F

Press, psig Heat Transfer Coefficients, BTU/hr-ft2-°F Surge Nozzle Thermal Sleeve Region in Surge Nozzle (2)

Bottom Head

& Lower Shell Heatup /

Cooldown 0

70 70 0

300 100 300 10494 653 653 2300 39294 653 653 2300 49788 (3) 70 70 0

Loss of Load 0

653 653 Max. 2710 Min. 1685 15000 150 300 10 672 644.5 20 672 617 40 672 614.7 71 610 549.2 240 594 514.0 600 560 480 2000 560 560 Notes:

1. Above table is reproduced from Table 5-6 of Reference [4].

2. The B&W surge nozzle does not include a thermal sleeve and the heat transfer coefficient for the Thermal Sleeve Region in the Surge Nozzle is ignored.

3. Note that an end time of 51,048 seconds was run instead. This results in a cooldown rate of 178.45°F/hr vs. a design of 200°F/hr. The cooldown rate of 178.45°F/hr. used in the analysis (vs. a design of 200°F/hr.) only affects the cooldown ramp; the heatup ramp is not impacted. The change in cooldown rate is expected to have minimal impact on the stresses.

File No.: 2100561.302 Revision: 1 Page 12 of 38 F0306-01R4 Table 4. Insurge/Outsurge Transients for Pressurizer Surge Nozzle Transient Group Time sec Pzr Temp oF Pzr Press psig Pool Temp oF Pool Height ft Fluid Temp at Pzr Nozzle oF HPzr Btu/hr-ft2-°F HPool Btu/hr-ft2-°F HNoz Btu/hr-ft2-°F HTS Btu/hr-ft2-°F (3)

Group 3 0

550 1000 550 0

550 300 200 300 100 900 550 1000 390 12 220 300 200 300 100 5200 550 1000 400 12 220 300 50 (2) 300 100 5201 550 1000 400 12 400 400 50 (2) 300 100 6200 550 1000 400 0

400 400 50 (2) 300 100 6201 550 1000 550 0

550 400 50 (2) 300 100 Notes:

1. Above table is reproduced from Table 5-9 of Reference [4].

2. Pool heat transfer coefficient conservatively increased to 200 Btu/hr-ft²-°F.

3. The B&W surge nozzle does not include a thermal sleeve and the heat transfer coefficient for the Thermal Sleeve Region in the Surge Nozzle is ignored.

File No.: 2100561.302 Revision: 1 Page 13 of 38 F0306-01R4 Figure 1. Modeled Dimensions (Dimensions Based on Figure 1 of Reference [2], Reference [3] and Reference [6].)

File No.: 2100561.302 Revision: 1 Page 14 of 38 F0306-01R4 Figure 2. Weld Locations (Weld Location based on Reference [6].)

File No.: 2100561.302 Revision: 1 Page 15 of 38 F0306-01R4 Figure 3. 3-D Finite Element Model

File No.: 2100561.302 Revision: 1 Page 16 of 38 F0306-01R4 Figure 4. Applied Boundary Conditions and Unit Internal Pressure (Units for pressure is psi.)

File No.: 2100561.302 Revision: 1 Page 17 of 38 F0306-01R4 Heat Transfer Coefficient Bulk Temperature Figure 5. Example of Applied Thermal Boundary Conditions for Thermal Transient Analyses Heatup/Cooldown transient shown; loads applied at time = 51,048 seconds.

(Units for HTC is BTU/sec-in2-°F, TBULK is °F.)

File No.: 2100561.302 Revision: 1 Page 18 of 38 F0306-01R4 Heat Transfer Coefficient Bulk Temperature Figure 6. Example of Applied Thermal Boundary Conditions for Insurge/Outsurge Transient Analyses Insurge/Outsurge transient Group 3 shown, loads applied at time = 900 seconds.

(Units for HTC is BTU/sec-in2-°F, TBULK is °F.)

File No.: 2100561.302 Revision: 1 Page 19 of 38 F0306-01R4 Figure 7. Applied Mechanical Boundary Conditions for Thermal Stress Analyses

File No.: 2100561.302 Revision: 1 Page 20 of 38 F0306-01R4 X-direction Stress (Radial Stress to the Nozzle) Y-direction Stress (Axial Stress to the Nozzle)

Z-direction Stress (Hoop Stress to the Nozzle)

Figure 8. Stress Contours Due to Unit Internal Pressure (Units for stress is psi.)

File No.: 2100561.302 Revision: 1 Page 21 of 38 F0306-01R4 Figure 9. Temperature Contour Heatup/Cooldown Transient (Time = 10,494 seconds)

(Units for temperature is °F.)

File No.: 2100561.302 Revision: 1 Page 22 of 38 F0306-01R4 X-direction Stress (Radial Stress to the Nozzle) Y-direction Stress (Axial Stress to the Nozzle)

Z-direction Stress (Hoop Stress to the Nozzle)

Figure 10. Stress Contours of Heatup/Cooldown Transient (Time = 10,494 seconds)

(Units for stress is psi.)

File No.: 2100561.302 Revision: 1 Page 23 of 38 F0306-01R4 Figure 11. Temperature Contour Insurge/Outsurge Group 3 Transient (Time = 900 seconds)

(Units for temperature is °F.)

File No.: 2100561.302 Revision: 1 Page 24 of 38 F0306-01R4 X-direction Stress (Radial Stress to the Nozzle) Y-direction Stress (Axial Stress to the Nozzle)

Z-direction Stress (Hoop Stress to the Nozzle)

Figure 12. Stress Contours of Insurge/Outsurge Group 3 Transient (Time = 900 seconds)

(Units for stress is psi.)

File No.: 2100561.302 Revision: 1 Page 25 of 38 F0306-01R4 Figure 13. Path Locations

File No.: 2100561.302 Revision: 1 Page 26 of 38 F0306-01R4 Figure 14. Through-Wall Stress Distribution at Path 1

File No.: 2100561.302 Revision: 1 Page 27 of 38 F0306-01R4 Figure 15. Through-Wall Stress Distribution at Path 2

File No.: 2100561.302 Revision: 1 Page 28 of 38 F0306-01R4 Figure 16. Through-Wall Stress Distribution at Path 3

File No.: 2100561.302 Revision: 1 Page 29 of 38 F0306-01R4 Figure 17. Through-Wall Stress Distribution at Path 4

File No.: 2100561.302 Revision: 1 Page 30 of 38 F0306-01R4 Figure 18. Through-Wall Stress Distribution at Path 5

File No.: 2100561.302 Revision: 1 Page 31 of 38 F0306-01R4 Figure 19. Through-Wall Stress Distribution at Path 6

File No.: 2100561.302 Revision: 1 Page 32 of 38 F0306-01R4 Figure 20. Through-Wall Stress Distribution at Path 7

File No.: 2100561.302 Revision: 1 Page 33 of 38 F0306-01R4 Figure 21. Through-Wall Stress Distribution at Path 8

File No.: 2100561.302 Revision: 1 Page 34 of 38 F0306-01R4 Figure 22. Through-Wall Stress Distribution at Path 9

File No.: 2100561.302 Revision: 1 Page 35 of 38 F0306-01R4 Figure 23. Through-Wall Stress Distribution at Path 10

File No.: 2100561.302 Revision: 1 Page 36 of 38 F0306-01R4 Figure 24. Through-Wall Stress Distribution at Path 11

File No.: 2100561.302 Revision: 1 Page 37 of 38 F0306-01R4 Figure 25. Through-Wall Stress Distribution at Path 12

File No.: 2100561.302 Revision: 1 Page 38 of 38 F0306-01R4 Figure 26. Through-Wall Stress Distribution at Path 13

File No.: 2100561.302 Revision: 1 Page A-1 of A-2 F0306-01R4 COMPUTER FILES LISTING

File No.: 2100561.302 Revision: 1 Page A-2 of A-2 F0306-01R4 File Name Description BW-Surge-Geom.INP Input file to construct model for a B&W PWR pressurizer surge nozzle, bottom head and lower shell.

PRESS.INP Unit internal pressure input file HUCD.INP Plant Heatup/Cooldown thermal analysis file LOL.INP Loss of Load thermal analysis file IO_GP3.INP Insurge/Outsurge Group 3 thermal analysis file STRESS.INP Input file for stress analyses for thermal transients CMNTR.mac ANSYS macro used to develop temperature load files during stress analysis.

$$_mntr.inp Thermal analysis load step input file for thermal transients, $$ =

HUCD for Heatup/Cooldown, LOL for Loss of Load, and IO_GP3 for Insurge/Outsurge Group 3 Post.INP Input file for post-processing GenStress.mac ANSYS Macro for stress extraction GETPATH.TXT Input file for defining stress Path 1 through Path 13

$$_MAP_P%.CSV Output files containing mapped stresses, $$ = STR_HUCD for Heatup/Cooldown, STR_LOL for Loss of Load, STR_IO_GP3 for Insurge/Outsurge Group 3, and PRESS for unit internal pressure loading, % = Paths 1 - 13 Results.xlsx Excel file to create Figure 14 through Figure 26.

- SI Calculation 2100561.303, Rev. 2 Proposed Alternative RA-22-0257 in Accordance with 10 CFR 50.55a(z)(1)

ATTACHMENT 8 SI CALCULATION 2100561.303, REV. 2 (40 PAGES)

DETERMINISTIC AND PROBABILISTIC FRACTURE MECHANICS ANALYSES OF OCONEE UNITS 1, 2 AND 3 BABCOCK & WILCOX PWR PRESSURIZER SURGE NOZZLE AND BOTTOM HEAD

CALCULATION PACKAGE File No.: 2100561.303 Project No.: 2100561 Quality Program Type:

Nuclear Commercial PROJECT NAME:

Duke PWR Fleet Steam Generator & Pressurizer Inspection Optimization Relief Requests CONTRACT NO.:

GSA # 03021365 CLIENT:

Duke Energy Corporation PLANT:

Oconee Nuclear Station, Units 1, 2 and 3 CALCULATION TITLE:

Deterministic and Probabilistic Fracture Mechanics Analyses of Oconee Units 1, 2 and 3 Babcock & Wilcox PWR Pressurizer Surge Nozzle and Bottom Head Document Revision Affected Pages Revision Description Project Manager Approval Signature & Date Preparer(s) &

Checker(s)

Signatures & Date 0

1 - 38 A A-2 Initial Issue Scott Chesworth 5/27/2022 Dilip Dedhia 5/27/2022 Nathaniel G. Cofie 5/27/2022 1

11 Revised Reference [5]

information; removed Proprietary notation from all locations Scott Chesworth 6/2/2022 Scott Chesworth 6/2/2022 Nathaniel G. Cofie 6/2/2022 2

6, 11 Corrected Reference [1];

added clarifying wording to Sections 2.2.3.1 and 4.0 Scott Chesworth 8/8/2022 Scott Chesworth 8/8/2022 Nathaniel G. Cofie 8/8/2022

File No.: 2100561.303 Revision: 2 Page 2 of 38 F0306-01R4 Table of Contents 1.0 OBJECTIVE................................................................................................................ 5 2.0 DFM EVALUATION..................................................................................................... 5 2.1 Technical Approach......................................................................................... 5 2.2 Design Inputs.................................................................................................. 5 2.3 Results of Deterministic Fracture Mechanics Evaluation................................ 9 3.0 PFM EVALUATION..................................................................................................... 9 3.1 Technical Approach......................................................................................... 9 3.2 Design Inputs.................................................................................................. 9 3.3 Inspection Coverage..................................................................................... 10 3.4 Results of PFM Evaluation............................................................................ 10

4.0 CONCLUSION

S........................................................................................................ 10

5.0 REFERENCES

.......................................................................................................... 11 COMPUTER FILES LISTING........................................................................ A-1

File No.: 2100561.303 Revision: 2 Page 3 of 38 F0306-01R4 List of Tables Table 1: Summary of DFM Design Inputs............................................................................. 13 Table 2: ONS 1/2/3 Plant Specific Transient Cycles Used in the DM and PFM Evaluations 13 Table 3: Summary of ONS 1/2/3 Plant Specific Insurge/Outsurge Temperature Differences and Numbers of Cycles Used in the DFM Evaluations(1)............................................ 14 Table 4: Results of the DFM Evaluation................................................................................ 15 Table 5: PFM Inputs for ONS 1/2/3 Inspection Scenario...................................................... 16 Table 6: Probability of Rupture (per year) and Probability of Leakage (per year) for 80 Years for Oconee Units 1, 2 and 3 Plant Specific Inspection Scenario (PSI+10+20+30+40+70)

............................................................................................................................ 17 Table 7: Sensitivity of the Probability of Rupture (per year) and Probability of Leakage (per year) for 80 Years for ONS 1/2/3 Plant Specific Inspection Scenario (PSI+10+20+30+40+70) with Fracture Toughness............................................. 18 Table 8: Sensitivity of the Probability of Rupture (per year) and Probability of Leakage (per year) for 80 Years for ONS 1/2/3 Plant Specific Inspection Scenario (PSI+10+20+30+40+70) with Stress................................................................... 19 Table 9: Sensitivity of the Probability of Rupture (per year) and Probability of Leakage (per year) for 80 Years for ONS 1/2/3 Plant Specific Inspection Scenario (PSI+10+20+30+40+70) with Fracture Toughness and Stress........................... 20 Table 10: Sensitivity Study for ISI Examination Coverage for ONS 1/2/3 B3.110 Welds...... 21

File No.: 2100561.303 Revision: 2 Page 4 of 38 F0306-01R4 List of Figures Figure 1. Modeled Dimensions (From Reference [2]).......................................................... 22 Figure 2: Path Locations (From Reference [2]).................................................................... 23 Figure 3 Through-Wall Stress Distribution for Path P1 [2]................................................... 24 Figure 4. Through-Wall Stress Distribution for Path P2 [2].................................................. 25 Figure 5. Through-Wall Stress Distribution for Path P3 [2].................................................. 26 Figure 6. Through-Wall Stress Distribution for Path P4 [2].................................................. 27 Figure 7. Through-Wall Stress Distribution for Path P5 [2].................................................. 28 Figure 8. Through-Wall Stress Distribution for Path P6 [2].................................................. 29 Figure 9. Through-Wall Stress Distribution for Path P7 [2].................................................. 30 Figure 10. Through-Wall Stress Distribution for Path P8 [2]................................................ 31 Figure 11. Through-Wall Stress Distribution for Path P9 [2]................................................ 32 Figure 12. Through-Wall Stress Distribution for Path P10 [2].............................................. 33 Figure 13. Through-Wall Stress Distribution for Path P11 [2].............................................. 34 Figure 14. Through-Wall Stress Distribution for Path P12 [2].............................................. 35 Figure 15. Through-Wall Stress Distribution for Path P13 [2].............................................. 36 Figure 16. Weld Residual Stress Distribution....................................................................... 37 Figure 17. Semi-Elliptical Axial Crack in a Cylinder Model.................................................. 37 Figure 18. Semi-Elliptical Circumferential Crack in a Cylinder Model.................................. 37 Figure 19. The Effect of Temperature on the Fracture Toughness, JIc, of SA-516 Grade 70 Steel

[11]...................................................................................................................... 38

File No.: 2100561.303 Revision: 2 Page 5 of 38 F0306-01R4 1.0 OBJECTIVE EPRI Report 3002015906 [1] developed the technical basis for optimizing the examination of PWR pressurizer shell and nozzle weld components. The pressurizer configuration in the Reference [1]

report is representative of Westinghouse and CE designs. The configuration of the B&W design pressurizers at Oconee Units 1, 2 and 3 (ONS 1/2/3) is considerably different from the Westinghouse/CE designs in terms of both geometry and materials; therefore, an evaluation is required to address the B&W pressurizer configuration. To this end, stress analyses of the ONS 1/2/3 design were performed in Reference [2] for the pressurizer bottom head and in Reference [3] for the pressurizer top head.

The objective of this calculation is to perform deterministic fracture mechanics (DFM) and probabilistic fracture mechanics (PFM) analyses of the B&W pressurizer components using the results of the stress analysis in Reference [2] and other relevant design inputs. The DFM evaluation will determine how long a postulated flaw will take to reach the ASME Code allowable flaw size, while the PFM evaluation will determine the probabilities of failure (leak and rupture) at the component locations.

The ASME Code,Section XI item numbers associated with the ONS 1/2/3 pressurizer components are:

Item No. B2.11 - Pressurizer, shell-to-head welds, circumferential Item No. B2.12 - Pressurizer, shell-to-head welds, longitudinal Item No. B3.110 - Pressurizer, nozzle-to-vessel welds 2.0 DFM EVALUATION 2.1 Technical Approach The technical approach used in the DFM evaluation is to postulate an initial flaw size equivalent to the relevant ASME Code,Section XI acceptance standard [4]. The ASME Code,Section XI fatigue crack growth (FCG) law, with the through-wall stress distributions from References [2, 3] and appropriate fracture mechanics models, is then used to determine the length of time for the postulated initial flaw to grow to a depth of 80% of the wall thickness (assumed to equate to leakage in this evaluation) or the depth at which the allowable toughness (KIC reduced by a structural factor of 2.0 for primary stresses and 1.0 for secondary stresses) is reached, whichever is less.

2.2 Design Inputs The design inputs used in the DFM evaluation are summarized in Table 1 and discussed in the following sections.

2.2.1 Geometry The geometries of the B&W pressurizer components considered in the evaluation are presented in Figure 1 [2]. The specific components of interest are the nozzle-to-head welds and head-to-shell welds for the pressurizer bottom head (Note: comparison of the stresses in References [2, 3] indicate the stresses of the bottom head in Reference [2] are controlling). The dimensions at these welds are provided in Figure 1. Figure 2 [2] shows the stress paths where these welds are located.

File No.: 2100561.303 Revision: 2 Page 6 of 38 F0306-01R4 2.2.2 Initial Crack Size and Shape For all components, an initial crack size of 5.2% of the wall thickness (which corresponds to the most conservative flaw acceptance standard for these components from Table IWB-3510-1 of ASME Code,Section XI [4]) was used in the DFM evaluation. This initial crack depth is the maximum value from these two tables with an associated crack aspect ratio (half crack length-to-crack depth) of 1.0. This crack shape results in the most conservative initial stress intensity factor (K) at the deepest point of the crack. The aspect ratio is then subsequently allowed to vary during the crack growth process.

2.2.3 Applied Stresses 2.2.3.1 Operating Transient Stresses Comparison of the stresses in References [2, 3] indicate the stresses of the bottom head in Reference

[2] are controlling and therefore will be used in this evaluation. The applied stresses consist of through-wall stresses due to pressure and the thermal transients described in Reference [2]. Typical through-wall stress distributions for stress paths used in the evaluation from Reference [2] are reproduced in Figures 3 to 15. Figure 2 shows the stress paths where these stresses were extracted. Plant-specific number of cycles for ONS 1/2/3 general transients [5] and insurge/outsurge transients [6] shown in Tables 2 and 3 (respectively) were used in the evaluations. It should be noted that for the PFM evaluation, the conservative number of cycles associated with transients defined in Reference [2] was used in lieu of the plant-specific transients.

2.2.3.2 Weld Residual Stresses Pressure vessel welds typically receive post-weld heat treatment (PWHT) to reduce the effects of weld residual stresses. In this evaluation, weld residual stresses remaining after PWHT were characterized in the form of a cosine distribution with a peak stress of 8 ksi [7] as shown in Figure 16, consistent with what was used in Reference [1].

2.2.4 Fracture Mechanics Models In this evaluation, all pre-existing flaws were conservatively assumed to be surface flaws. Two different fracture mechanics models were used for axial and circumferential flaws. For an axial flaw, the stress intensity factor (K) solution for an internal, semi-elliptical crack from API-579/ASME-FFS-1 [8] was used.

This model is shown in Figure 17. The aspect ratio (a/c) is allowed to vary during crack growth.

For a circumferential flaw, the K solution for an internal, semi-elliptical crack from API-579/ASME-FFS-1

[8] was used. This model is shown in Figure 18. The aspect ratio (a/c) was allowed to vary during crack growth. These models are consistent with those used in Reference [1].

These fracture mechanics models were incorporated into an SI-developed software, SI-TIFFANY [9],

that determines the K distribution due to through-wall stress profiles for both circumferential and axial cracks. The outputs of SI-TIFFANY are the maximum and minimum K distributions as well as the K distribution for each transient.

File No.: 2100561.303 Revision: 2 Page 7 of 38 F0306-01R4 2.2.5 Fracture Toughness The materials under consideration are low carbon ferritic steels (SA-516 Grade 70 and SA-212). The fracture toughness provided in ASME Code,Section XI, Appendix A [4] only applies to low alloy ferritic steels such as SA-533 Grade B Class 1, SA-508 Class 2, and SA-508 Class 3 (typically used in the fabrication of Westinghouse and CE pressurizers) and is therefore not applicable to the B&W pressurizer design. The fracture toughness for low carbon ferritic steel piping components is provided in Appendix C of ASME Code,Section XI. Though the pressurizers are not piping components, guidance will be taken from Appendix C in addition to information in the open literature in determining a reasonable lower bound fracture toughness for use in this evaluation. Lower bound values of fracture toughness for ferritic piping are provided in ASME Code,Section XI, Appendix C for both circumferential and axial flaws.

Table C-8321-1 of Appendix C provides fracture toughness values for two categories of ferritic materials operating in the upper shelf region for circumferential flaws. The first material category includes seamless or welded wrought ferritic steel pipe and pipe fittings that have a material yield strength not greater than 40 ksi (280 MPa) and welds made with E7015, E7016, and E7018 electrodes in the as-welded or PWHT condition. The fracture toughness (JIc) value for circumferential flaws for this material category is 600 in.-lb/in.2 (105 kN/m) at temperatures equal to or greater than the upper shelf temperature. The second material category is for all other ferritic shielded metal arc and submerged arc weld with a specified minimum tensile strength not greater than 80 ksi (550 MPa) in the as-welded or post-weld heat treated condition for circumferential flaws. The JIc value for this material category is 350 in.-lb/in.2 (61 kN/m) at temperatures equal to or greater than the upper shelf temperature.

Similarly, the JIc value for axial flaws for from Table C-8322-1 of Appendix C for base metals and weldments is 300 in.-lb/in.2 (53 kN/m) at temperatures equal to or greater than the upper shelf temperature. However, per the technical basis document for ferritic piping [10], this value of fracture toughness is associated with small diameter piping (4-inch and 6-inch NPS) for SA-106 Grade B material and was determined in the C-L direction for through-wall flaws. Since the pressurizer components are much larger in diameter and part-wall flaws are considered in this evaluation, this fracture toughness is not considered appropriate for use in the evaluation of the pressurizer components. Therefore, the lower bound fracture toughness value in the circumferential direction was used in determining the flaw acceptance criteria for all flaws in the pressurizer components in this evaluation.

From Table C-8321-1 of Appendix C, circumferential flaws in ferritic steel base metals and weldments that have thicknesses equal to or greater than one inch and an operating temperature greater than 52oF (11°C) are in the upper shelf region. The thickness of one inch or greater is consistent with the dimensions of the pressurizer shown in Figure 1.

Using the relation from Table C-8321 of the ASME Code, the value of the plane strain fracture toughness (KIc) can be determined from:

J1c = 1000 (KIc)2 / E Eq. 1 where E = E / (1 - 2)

E = modulus of elasticity = 29,200 ksi at 70°F (from Table 5-2)

File No.: 2100561.303 Revision: 2 Page 8 of 38 F0306-01R4

= Poisson ratio = 0.3 Using Equation 1, the values of JIc of 350 in.-lb/in.2 translates into KIc value of 106 ksiin. This lower value bound value of KIc will be used in this evaluation.

The fracture toughness determined above is applicable for piping components. The pressurizer under consideration in this study is comprised of a cylindrical shell and nozzle. As such, the reasonableness of applying a fracture toughness value for piping to the components is investigated for SA-516 Grade 70 and SA-212.

The fracture toughness for SA-516 Grade 70, typical of what is used to fabricate pressurizer shells has been investigated by several researchers [11, 12, 13]. A plot of the fracture toughness, JIc, versus temperature for SA-516, Grade 70 from Reference [11] is shown in Figure 19. Per Reference [11], the JKe data (black solid data) is only valid in the temperature range of -130oC to -160oC. As indicated in this figure, JIc remains constant at temperatures above 0°C (32°F) at a value of 120 kN/m (685 in.-

lb/in2). This upper shelf toughness is greater than the JIc value of 350 in.-lb/in.2 established above.

Studies performed in References [12, 13] determined the average value of KIc at room temperature for SA-516 Grade 70 steel base material and weldments to be 129 MPam (117 ksiin), which is greater than the 106 ksiin established above. In addition, data presented in Reference [11] for various ferritic steels show that the fracture toughness (JIc) of SA-516 Grade 70 material at 550oF is at least 800 in.-

lb/in2. Therefore, the lower bound upper shelf fracture toughness of 106 ksiin derived from Appendix C of Section XI of the ASME Code for circumferential flaws in piping is a conservative lower bound for use in this evaluation for the SA-106 pressurizer material at ONS 1/2/3.

SA-212 is an older version of SA-516, Grade 70. It first occurred in the 1940 Edition of ASME Code,Section II. In the 1968 edition of Section II, the SA-212 Specification was deleted and was replaced with two specifications, SA-515 (Specification for Pressure Vessel Plates, Carbon Steel, for Intermediate and Higher Temperature Service) and the SA-212 steel plate melted to fine grain practice was replaced with SA-516 (Specification for Pressure Vessel Plates, Carbon Steel, for Moderate and Lower Temperature Service). SA-212 was the material of choice for fabrication of railroad tank cars. Data presented n Table 4-11 of Reference [14] indicate that the upper shelf energy for SA-212 range from 45 to 68 ft lbs.

in the longitudinal direction and 30 to 34 ft lbs. in the transverse direction. Data from Table A1 of Reference [15] also show that the minimum upper shelf energy at temperatures greater than 212oF is 71.4 ft-lbs. and from Table A4, it is 45.1 ft-lbs. Using Equation 3.2 from Reference [16]:

(KIc)2/E = 2(CVN)3/2 (psi-in., ft-lb.)

Eq. 2 where:

KIc = fracture toughness E = modulus of elasticity = 29,200 ksi at 70°F CVN = energy absorbed Using a reasonable lower bound value of CVN of 45 ft-lbs. from the data in References [14, 15] and the KIc-CVN correlation from Equation 2 results in KIc of 132.3 ksiin. Therefore, the lower bound upper shelf fracture toughness of 106 ksiin derived from Appendix C of Section XI of the ASME Code for

File No.: 2100561.303 Revision: 2 Page 9 of 38 F0306-01R4 circumferential flaws in piping is a conservative lower bound for use in this evaluation in application to SA-212 pressurizer material at ONS 1/2/3.

2.2.6 Fatigue Crack Growth Law The FCG law for ferritic steels, as defined in ASME Code,Section XI, Appendix A, Paragraph A-4300

[4], was used in the evaluation.

2.3 Results of Deterministic Fracture Mechanics Evaluation The results of the DFM evaluation are summarized in Table 2. The table shows that the periods required for hypothetical postulated flaws to reach the allowable fracture toughness or 80% of thickness are very long, which indicates that all the evaluated components are very flaw tolerant. Because the DFM evaluation considered hypothetical postulated flaws, structural factors of 2.0 on primary loads and 1.0 on secondary loads, consistent with ASME Code,Section XI, Appendix G, were applied. Hence, a structural factor of 2.0 was applied to the applied K due to pressure stress and a structural factor of 1.0 was applied to the thermal and residual stresses. The resulting K with the structural factors were compared to the allowable fracture toughness of 106 ksiin to determine the allowable operating period.

Table 5 shows that the number of years to reach the fracture toughness with a structural factor of 2.0 on primary stresses and 1.0 on secondary stresses for all locations. As seen in this table, it takes a minimum of 208 years at the limiting location (Case ID PRSHC-BW-4C) to reach the fracture toughness.

3.0 PFM EVALUATION 3.1 Technical Approach The PFM evaluation was performed consistent with the evaluation presented in the EPRI Report [1].

PROMISE, Version 2.0 [17] was used to perform the PFM evaluations. The evaluation considered the ONS 1/2/3 plant-specific inspection history.

3.2 Design Inputs The design inputs used for the PFM evaluation are shown in Table 5 for ONS 1/2/3 plant specific inspection history and are consistent with those used in Reference 1. ONS 1/2/3 has performed pre-service (PSI) inspection at year zero followed by four successive 10-year in-service inspections (ISI).

This is to be followed by 30-year inspection deferral which is being considered in the Relief Request being developed by Duke Energy. This plant specific history was considered in the PFM evaluation.

Stress and fracture toughness which were identified as the key variables in the PFM evaluation in Reference [1]. As such, three sensitivity studies were performed as part of the PFM as follows:

1.

The fracture toughness was decreased to determine the minimum fracture toughness that will meet the acceptance criteria of 1.0x10-6.

2.

The stresses were increased to determine the maximum stress multiplier that will meet the acceptance criteria of 1.0x10-6.

3.

A sensitivity study of the combined effects of the fracture toughness and stress.

File No.: 2100561.303 Revision: 2 Page 10 of 38 F0306-01R4 3.3 Inspection Coverage Inspection coverage for all the welds under consideration for ONS 1/2/3 is provided in Reference [18].

For Item Nos. B2.11 and B2.12, the inspection coverage is greater than 90% (essentially 100%) for all welds. However, for Items B3.110, some welds have limited coverage. The minimum coverage for Unit 1 is 25.8%, for Unit 2 is 25.2 and for Unit 3 is 30.0%. A sensitivity study is performed with the limiting minimum coverage of 25.2% for Unit 2. Evaluations were performed using this limiting coverage to determine the probabilities of rupture and leakage for the plant-specific inspection scenarios of (PSI+10+10+30+40+70) using the same input parameters as in Table 6. For comparison, evaluations were also performed for the current ASME Code,Section XI mandated 10-year inspection interval of (PSI+10+20+30+40+50+60+70).

3.4 Results of PFM Evaluation The results of the PFM evaluation are presented in Table 6 for ONS 1/2/3 plant specific inspection history. As shown in this table, the probabilities of rupture and leakage are all below the acceptance criteria of 1.0x10-6 after 80 years of plant operation by three orders of magnitude.

The results of the sensitivity studies are presented in Tables 7 through 9. From Table 7, the fracture toughness can be as low as 72 ksiin before the acceptance criterion of 1.0x10-6 is reached after 80 years of operation. From Table 8, a stress multiplier of 1.4 can be applied to all the stresses considered in the evaluation before the acceptance criterion is reached. Table 9 shows that by applying a stress multiplier of 1.1 and reducing the fracture toughness to 80 ksiin, the probabilities of rupture and leakage are all below the acceptance criterion of 1.0x10-6 after 80 years of plant operation.

The results of the sensitivity study on coverage are presented in Table 10. As shown in Table 10, considering the most limiting coverage for Item No. B3.110 and the ONS 1/2/3 PSI/ISI scenario, the probabilities of rupture and leakage are below the acceptance criteria of 1.0x10-6 after 80 years of operation by three orders of magnitude. Furthermore, when the probabilities of rupture and leakage for the alternative inspection schedule are compared to the present ASME Code,Section XI inspection schedule, there is no difference. This indicates that there is no change in risk from the current ASME Code,Section XI schedule to that of the alternative inspection schedule.

4.0 CONCLUSION

S From the PFM and DFM evaluations, the following conclusions are made:

The DFM evaluation demonstrated that a very long operating period (greater than 200 years) is necessary for a postulated initial flaw (with a depth equal to ASME Code,Section XI acceptance standards) to reach the allowable fracture toughness or propagate through 80% of the wall thickness (assumed as leakage for this study). This indicates that all in-scope components at ONS 1/2/3 are very flaw tolerant.

For the ONS 1/2/3 specific inspection history, the PFM evaluation showed that the probabilities of rupture and leakage are significantly below the acceptance criterion of 1.0x10-6 failures per year after 80 years of operation.

File No.: 2100561.303 Revision: 2 Page 11 of 38 F0306-01R4 In the PFM evaluations, conservative number of cycles were used (300 cycles analyzed versus projected actual cycles of 134 cycles for 60 years, or 400 cycles analyzed versus projected actual cycles of 179 for 80 years). In the DFM evaluation, the results indicate that the plant can operate safely for 208 years (the equivalent of 464 cycles). This compares with the actual projected cycles of 134 for 60 years and 179 for 80 years. Hence there are sufficient margins to accommodate any deviations from the projected cycles.

Sensitivity studies involving stress and fracture toughness indicated that when all stresses are increased by a factor of 1.4 or the fracture toughness reduced from 106 ksiin to 72 ksiin, the acceptance criteria are met for both rupture and leakage.

A sensitivity study involving increasing all the stresses by 10% and reducing the fracture toughness from 106 ksiin to 80 ksiin also showed that the probability of rupture and leakage are below the acceptance criteria.

For Item Nos. B2.11 and B2.12, coverage is greater than 90% for all welds and therefore essentially 100% coverage is achieved. For Item No. B3.110, the minimum coverage is 25.2%.

An evaluation using this coverage results in acceptable probabilities of rupture and leakage for this Item No. Furthermore, when compared to the mandated ASME Code,Section XI inspection schedule, there is no change in risk.

5.0 REFERENCES

1. Technical Bases for Inspection Requirements for PWR Pressurizer Vessel Head, Shell-to-Head and Nozzle-to-Vessel Welds. EPRI, Palo Alto, CA: 2019. 3002015905, ADAMS Accession No. ML21021A271.

2. SI Calculation 2100561.302, Rev. 1, Finite Element Model Development and Thermal/Mechanical Stress Analysis of Babcock & Wilcox PWR Pressurizer Surge Nozzle and Bottom Head.

3. SI Calculation 2100561.304, Rev. 0, Finite Element Model Development and Thermal/Mechanical Stress Analysis of Babcock & Wilcox PWR Pressurizer Safety Relief Nozzle and Top Head.

4. ASME Boiler and Pressure Vessel Code,Section XI, 2017 Edition.

5. SI Calculation FP-ONS-304P, Oconee SI:FatiguePro 4 Baseline Analysis, Startup through 11/3/2020 (U1), 11/21/2019 (U2) and 4/24/2020 (U3), Revision 0. Only non-Proprietary information from this reference was used.

6. SI Calculation 2100561.301, Rev. 0, Duke Plants Insurge/Outsurge Transients.

7. Simonen, F. A. and Johnson, K. I., Effects of Residual Stresses and Underclad Flaws on the Reliability of Reactor Pressure Vessels, PVP-Vol. 251, Reliability and Risk in Pressure Vessels and Piping, ASME PVP Conference, 1993.

File No.: 2100561.303 Revision: 2 Page 12 of 38 F0306-01R4

8. API Standard 579-1/ASME FFS-1, Fitness-for-Service, Second Edition, June 2016.

9. SI-TIFFANY 3.1, Structural Integrity Associates, September 2018.

10. Novetech Corporation, Evaluation of Flaws in Ferritic Piping, EPRI NP-6045, October,1988.

11. C-S Seok, Effect of Temperature on the Fracture Toughness of A-516 Grade 70 Steel, Korean Society of Mechanical Engineers (KSME) International Journal, Vol. 14, No. 1, pp 11 - 18, 2000.

12. V. Mehta, Evaluation of the Fracture Parameters for SA-516 Grade 70 Material, IOSR Journal of Mechanical and Civil Engineering (IOSR-JMCE), e-ISSN: 2278-1684, p-ISSN: 2320-334X, Volume 13, Issue 3 Ver. III (May-Jun. 2016), PP 38-45.

13. V. Mehta, Experimental Study of the Fracture Parameters for the SMAW Joints of SA 516 -

Grade 70 Material, International Journal of Research and Scientific Innovation (IJRSI), Volume IV, Issue VI, June 2017, ISSN 2321-2705.

14. A. Zahoor, Materials and Fracture Mechanics Assessments of Railroad Tank Cars, NISTIR 6266, U. S. Dept. of Commerce, Technology Administration, Materials Performance Group, Metallurgical Division, National Institute of Standards and Technology, Gaithersburg, MD 20899, September 1998.

15. J. G. Early, Metallurgical Analysis of ASTM A212-B Steel Tank Car Head Plate, Report No.

FRA/ORC 81/32, PB 81-205098, National Bureau of Standards, Washington DC 20545, April 1981.

16. R. Roberts and C. Newton, Interpretive Report on Small Scale Test Correlations with KIc Data, Welding Research Council Bulletin 265, February 1981.

17. Structural Integrity Associates Report DEV1806.402, PROMISE 2.0 Theory and Users Manual, Revision 1.

18. Duke Examination Results.xlsx, SI File No. 2100561.207.

File No.: 2100561.303 Revision: 2 Page 13 of 38 F0306-01R4 Table 1: Summary of DFM Design Inputs Input Value Geometry From Section 4 Initial Crack Size 5.2% of the thickness, c/a = 1 Fracture toughness 106 ksiin Fatigue crack growth law ASME Code,Section XI Appendix A, Paragraph A-4300 Operating Transient Stresses From Reference [2]

Operating Cycles From Tables 2 and 3 Residual stresses Cosine curve with 8 ksi peak (Figure 16)

Table 2: ONS 1/2/3 Plant Specific Transient Cycles Used in the DM and PFM Evaluations Transient ONS1/2/3 60-Year Projection Heatup /

Cooldown 132/134/104(1)

Loss of Load (Large Step Load Decrease, Loss of Power, Loss of Flow, Reactor Trip) 62/33/41(2)

Notes:

1. Heatup/Cooldown = RCS Heatup and RCS Cooldown from Tables 13, 14 and 15 of [5] scaled down from 80 to 60 years. 134 cycles conservatively used in the DFM evaluations.

2. Loss of Load = Rx Trip No Loss of Flow from Tables 13, 14 and 15 of [5] scaled down from 80 to 60 years. 62 cycles conservatively used in the DFM evaluations.

File No.: 2100561.303 Revision: 2 Page 14 of 38 F0306-01R4 Table 3: Summary of ONS 1/2/3 Plant Specific Insurge/Outsurge Temperature Differences and Numbers of Cycles Used in the DFM Evaluations(1)

T (oF)(2) 60-Year No. of Cycles for Evaluation 400 0

350 0

300 720 250 360 200 1440 Note:

(1) From Reference [6]

(2) T is the temperature difference between the pressurizer fluid temperature and the fluid temperature in the surge nozzle.

File No.: 2100561.303 Revision: 2 Page 15 of 38 F0306-01R4 Table 4: Results of the DFM Evaluation Item No.

Component Description Case Identification(1)

Years to Reach KIc of 106 ksi SFprimary = 2.0 SFsecondary = 1.0 B3.110 Pressurizer surge nozzle-to-vessel weld PRSNV-BW-1A 925 PRSNV-BW-1C 1783 PRSNV-BW-2A 980 PRSNV-BW-2C 1975 PRSNV-BW-3A 1601 PRSNV-BW-3C 2037 B2.11 B2.12 B2.21 B2.22 Pressurizer head-to-shall welds (Circ)

Pressurizer head-to-shell welds (Long)

Pressurizer head welds (Circ)

Pressurizer head welds (Meridional)

PRSHC-BW-4A 671 PRSHC-BW-4C 208 PRSHC-BW-5A 1648 PRSHC-BW-5C 288 PRSHC-BW-6A 1817 PRSHC-BW-6C 396 PRSHC-BW-7A 1633 PRSHC-BW-7C 402 PRSHC-BW-8A 290 PRSHC-BW-8C 468 PRSHC-BW-9A 232 PRSHC-BW-9C 478 PRSHC-BW-10A 231 PRSHC-BW-10C 899 PRSHC-BW-11A 334 PRSHC-BW-11C 1200 PRSHC-BW-12A 208 PRSHC-BW-12C 336 PRSHC-BW-13A 240 PRSHC-BW-13C 249 Note 1:

The Case Identification terminology is as follows: PR for Pressurizer; SNV for surge nozzle-to-vessel and SHC for head or shell-to-head; BW for B&W design; P1 through P13 represent the crack paths (see Figure 2); C for circumferential part-through-wall crack; and A for axial part-through-wall crack.

File No.: 2100561.303 Revision: 2 Page 16 of 38 F0306-01R4 Table 5: PFM Inputs for ONS 1/2/3 Inspection Scenario No. of Realizations Epistemic = 1, Aleatory = 10 million No. of cracks per weld 1, constant Crack depth distribution PVRUF Crack length distribution NUREG/CR-6817-R1 Fracture toughness (ksiin)

Normal (106,5)

Inspection coverage 100%

PSI Yes ISI 10, 20, 30, 40 and 70 years POD Curve BWRVIP-108, Figure 8-6 Fatigue crack growth law and threshold A-4300, log-normal, Second Parameter = 0.467 Operating Transient Stresses and Cycles From Reference [2]

Uncertainties on transients None Weld residual stresses (ksi)

Cosine Curve (8, 8), constant (not random)

File No.: 2100561.303 Revision: 2 Page 17 of 38 F0306-01R4 Table 6: Probability of Rupture (per year) and Probability of Leakage (per year) for 80 Years for Oconee Units 1, 2 and 3 Plant Specific Inspection Scenario (PSI+10+20+30+40+70)

Item No.

Component Description Case Identification Probability of Rupture after 80 Years Probability of Leakage after 80 Years KIc = 106 ksi Stress Multiplier = 1.0 Flaw Density =1.0 B3.110 Pressurizer surge nozzle-to-vessel weld PRSNV-BW-1A 1.25E-09 1.25E-09 PRSNV-BW-1C 1.25E-09 1.25E-09 PRSNV-BW-2A 1.25E-09 1.25E-09 PRSNV-BW-2C 1.25E-09 1.25E-09 PRSNV-BW-3A 1.25E-09 1.25E-09 PRSNV-BW-3C 1.25E-09 1.25E-09 B2.11 B2.12 B2.21 B2.22 Pressurizer head-to-shell welds (Circ)

Pressurizer head-to-shell welds (Long)

Pressurizer head welds (Circ)

Pressurizer head welds (Meridional)

PRSHC-BW-4A 1.25E-09 1.25E-09 PRSHC-BW-4C 2.50E-09 1.25E-09 PRSHC-BW-5A 1.25E-09 1.25E-09 PRSHC-BW-5C 1.25E-09 1.25E-09 PRSHC-BW-6A 1.25E-09 1.25E-09 PRSHC-BW-6C 1.25E-09 1.25E-09 PRSHC-BW-7A 1.25E-09 1.25E-09 PRSHC-BW-7C 1.25E-09 1.25E-09 PRSHC-BW-8A 1.25E-09 1.25E-09 PRSHC-BW-8C 1.25E-09 1.25E-09 PRSHC-BW-9A 1.25E-09 1.25E-09 PRSHC-BW-9C 1.25E-09 1.25E-09 PRSHC-BW-10A 1.25E-09 1.25E-09 PRSHC-BW-10C 1.25E-09 1.25E-09 PRSHC-BW-11A 1.25E-09 1.25E-09 PRSHC-BW-11C 1.25E-09 1.25E-09 PRSHC-BW-12A 2.50E-09 1.25E-09 PRSHC-BW-12C 1.25E-09 1.25E-09 PRSHC-BW-13A 1.25E-09 1.25E-09 PRSHC-BW-13C 1.25E-09 1.25E-09

File No.: 2100561.303 Revision: 2 Page 18 of 38 F0306-01R4 Table 7: Sensitivity of the Probability of Rupture (per year) and Probability of Leakage (per year) for 80 Years for ONS 1/2/3 Plant Specific Inspection Scenario (PSI+10+20+30+40+70) with Fracture Toughness Item No.

Component Description Case Identification Probability of Rupture after 80 Years Probability of Leakage after 80 Years Probability of Rupture after 80 Years Probability of Leakage after 80 Years KIc = 106 ksiin Stress Multiplier = 1.0 Flaw Density = 1.0 KIc = 72 ksiin Stress Multiplier = 1.0 Flaw Density = 1.0 B3.110 Pressurizer surge nozzle-to-vessel weld PRSNV-BW-1A 1.25E-09 1.25E-09 6.25E-09 1.25E-09 PRSNV-BW-1C 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSNV-BW-2A 1.25E-09 1.25E-09 3.75E-09 1.25E-09 PRSNV-BW-2C 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSNV-BW-3A 1.25E-09 1.25E-09 2.50E-09 1.25E-09 PRSNV-BW-3C 1.25E-09 1.25E-09 1.25E-09 1.25E-09 B2.11 B2.12 B2.21 B2.22 Pressurizer head-to-shell welds (Circ)

Pressurizer head-to-shell welds (Long)

Pressurizer head welds (Circ)

Pressurizer head welds (Meridional)

PRSHC-BW-4A 1.25E-09 1.25E-09 1.00E-08 1.25E-09 PRSHC-BW-4C 2.50E-09 1.25E-09 6.85E-07 1.25E-09 PRSHC-BW-5A 1.25E-09 1.25E-09 5.00E-09 1.25E-09 PRSHC-BW-5C 1.25E-09 1.25E-09 1.23E-07 1.25E-09 PRSHC-BW-6A 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSHC-BW-6C 1.25E-09 1.25E-09 2.63E-08 1.25E-09 PRSHC-BW-7A 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSHC-BW-7C 1.25E-09 1.25E-09 2.25E-08 1.25E-09 PRSHC-BW-8A 1.25E-09 1.25E-09 1.13E-08 1.25E-09 PRSHC-BW-8C 1.25E-09 1.25E-09 5.00E-09 1.25E-09 PRSHC-BW-9A 1.25E-09 1.25E-09 2.00E-08 1.25E-09 PRSHC-BW-9C 1.25E-09 1.25E-09 6.25E-09 1.25E-09 PRSHC-BW-10A 1.25E-09 1.25E-09 1.63E-08 1.25E-09 PRSHC-BW-10C 1.25E-09 1.25E-09 5.00E-09 1.25E-09 PRSHC-BW-11A 1.25E-09 1.25E-09 1.00E-08 1.25E-09 PRSHC-BW-11C 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSHC-BW-12A 2.50E-09 1.25E-09 2.65E-07 1.25E-09 PRSHC-BW-12C 1.25E-09 1.25E-09 2.25E-08 1.25E-09 PRSHC-BW-13A 1.25E-09 1.25E-09 1.41E-07 1.25E-09 PRSHC-BW-13C 1.25E-09 1.25E-09 1.25E-09 1.25E-09

File No.: 2100561.303 Revision: 2 Page 19 of 38 F0306-01R4 Table 8: Sensitivity of the Probability of Rupture (per year) and Probability of Leakage (per year) for 80 Years for ONS 1/2/3 Plant Specific Inspection Scenario (PSI+10+20+30+40+70) with Stress Item No.

Component Description Case Identification Probability of Rupture after 80 Years Probability of Leakage after 80 Years Probability of Rupture after 80 Years Probability of Leakage after 80 Years KIc = 106 ksiin Stress Multiplier = 1.0 Flaw Density = 1.0 KIc = 106 ksiin Stress Multiplier = 1.4 Flaw Density = 1.0 B3.110 Pressurizer surge nozzle-to-vessel weld PRSNV-BW-1A 1.25E-09 1.25E-09 2.50E-09 1.25E-09 PRSNV-BW-1C 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSNV-BW-2A 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSNV-BW-2C 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSNV-BW-3A 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSNV-BW-3C 1.25E-09 1.25E-09 1.25E-09 1.25E-09 B2.11 B2.12 B2.21 B2.22 Pressurizer head-to-shell welds (Circ)

Pressurizer head-to-shell welds (Long)

Pressurizer head welds (Circ)

Pressurizer head welds (Meridional)

PRSHC-BW-4A 1.25E-09 1.25E-09 3.75E-09 1.25E-09 PRSHC-BW-4C 2.50E-09 1.25E-09 8.56E-07 1.25E-09 PRSHC-BW-5A 1.25E-09 1.25E-09 2.50E-09 1.25E-09 PRSHC-BW-5C 1.25E-09 1.25E-09 5.00E-08 1.25E-09 PRSHC-BW-6A 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSHC-BW-6C 1.25E-09 1.25E-09 1.13E-08 1.25E-09 PRSHC-BW-7A 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSHC-BW-7C 1.25E-09 1.25E-09 1.13E-08 1.25E-09 PRSHC-BW-8A 1.25E-09 1.25E-09 7.50E-09 1.25E-09 PRSHC-BW-8C 1.25E-09 1.25E-09 3.75E-09 1.25E-09 PRSHC-BW-9A 1.25E-09 1.25E-09 2.63E-08 1.25E-09 PRSHC-BW-9C 1.25E-09 1.25E-09 2.50E-09 1.25E-09 PRSHC-BW-10A 1.25E-09 1.25E-09 2.13E-08 1.25E-09 PRSHC-BW-10C 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSHC-BW-11A 1.25E-09 1.25E-09 3.75E-09 1.25E-09 PRSHC-BW-11C 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSHC-BW-12A 2.50E-09 1.25E-09 4.90E-07 1.25E-09 PRSHC-BW-12C 1.25E-09 1.25E-09 2.50E-08 1.25E-09 PRSHC-BW-13A 1.25E-09 1.25E-09 1.36E-07 1.25E-09 PRSHC-BW-13C 1.25E-09 1.25E-09 1.25E-09 1.25E-09

File No.: 2100561.303 Revision: 2 Page 20 of 38 F0306-01R4 Table 9: Sensitivity of the Probability of Rupture (per year) and Probability of Leakage (per year) for 80 Years for ONS 1/2/3 Plant Specific Inspection Scenario (PSI+10+20+30+40+70) with Fracture Toughness and Stress Item No.

Component Description Case Identification Probability of Rupture after 80 Years Probability of Leakage after 80 Years KIc = 80 ksi Stress Multiplier = 1.1 Flaw Density = 1.0 B3.110 Pressurizer surge nozzle-to-vessel weld PRSNV-BW-1A 5.00E-09 1.25E-09 PRSNV-BW-1C 1.25E-09 1.25E-09 PRSNV-BW-2A 1.25E-09 1.25E-09 PRSNV-BW-2C 1.25E-09 1.25E-09 PRSNV-BW-3A 1.25E-09 1.25E-09 PRSNV-BW-3C 1.25E-09 1.25E-09 B2.11 B2.12 B2.21 B2.22 Pressurizer head-to-shell welds (Circ)

Pressurizer head-to-shell welds (Long)

Pressurizer head welds (Circ)

Pressurizer head welds (Meridional)

PRSHC-BW-4A 8.75E-09 1.25E-09 PRSHC-BW-4C 6.53E-07 1.25E-09 PRSHC-BW-5A 2.50E-09 1.25E-09 PRSHC-BW-5C 8.88E-08 1.25E-09 PRSHC-BW-6A 1.25E-09 1.25E-09 PRSHC-BW-6C 2.63E-08 1.25E-09 PRSHC-BW-7A 2.50E-09 1.25E-09 PRSHC-BW-7C 1.88E-08 1.25E-09 PRSHC-BW-8A 7.50E-09 1.25E-09 PRSHC-BW-8C 6.25E-09 1.25E-09 PRSHC-BW-9A 2.00E-08 1.25E-09 PRSHC-BW-9C 3.75E-09 1.25E-09 PRSHC-BW-10A 1.88E-08 1.25E-09 PRSHC-BW-10C 2.50E-09 1.25E-09 PRSHC-BW-11A 6.25E-09 1.25E-09 PRSHC-BW-11C 1.25E-09 1.25E-09 PRSHC-BW-12A 3.05E-07 1.25E-09 PRSHC-BW-12C 2.63E-08 1.25E-09 PRSHC-BW-13A 1.08E-07 1.25E-09 PRSHC-BW-13C 1.25E-09 1.25E-09

File No.: 2100561.303 Revision: 2 Page 21 of 38 F0306-01R4 Table 10: Sensitivity Study for ISI Examination Coverage for ONS 1/2/3 B3.110 Welds Stress Path ID ASME Section XI Inspection Interval PSI/ISI Scenario: 0,10,20,30,40,50,60,70 Flaw density = 1.0 KIC = 106 ksiin, SD = 5 ksiin Stress multiplier = 1.0 Coverage = 25.2%

Alternate Inspection Interval PSI/ISI Scenario: 0,10,20,30,40,70 Flaw density = 1.0 KIC = 106 ksiin, SD = 5 ksiin Stress multiplier = 1.0 Coverage = 25.2%

Probability of Rupture Probability of Leakage Probability of Rupture Probability of Leakage PRSNV-BW-1A 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSNV-BW-1C 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSNV-BW-2A 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSNV-BW-2C 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSNV-BW-3A 1.25E-09 1.25E-09 1.25E-09 1.25E-09 PRSNV-BW-3C 1.25E-09 1.25E-09 1.25E-09 1.25E-09

File No.: 2100561.303 Revision: 2 Page 22 of 38 F0306-01R4 Figure 1. Modeled Dimensions (From Reference [2])

File No.: 2100561.303 Revision: 2 Page 23 of 38 F0306-01R4 Figure 2: Path Locations (From Reference [2])

File No.: 2100561.303 Revision: 2 Page 24 of 38 F0306-01R4 Figure 3 Through-Wall Stress Distribution for Path P1 [2]

File No.: 2100561.303 Revision: 2 Page 25 of 38 F0306-01R4 Figure 4. Through-Wall Stress Distribution for Path P2 [2]

File No.: 2100561.303 Revision: 2 Page 26 of 38 F0306-01R4 Figure 5. Through-Wall Stress Distribution for Path P3 [2]

File No.: 2100561.303 Revision: 2 Page 27 of 38 F0306-01R4 Figure 6. Through-Wall Stress Distribution for Path P4 [2]

File No.: 2100561.303 Revision: 2 Page 28 of 38 F0306-01R4 Figure 7. Through-Wall Stress Distribution for Path P5 [2]

File No.: 2100561.303 Revision: 2 Page 29 of 38 F0306-01R4 Figure 8. Through-Wall Stress Distribution for Path P6 [2]

File No.: 2100561.303 Revision: 2 Page 30 of 38 F0306-01R4 Figure 9. Through-Wall Stress Distribution for Path P7 [2]

File No.: 2100561.303 Revision: 2 Page 31 of 38 F0306-01R4 Figure 10. Through-Wall Stress Distribution for Path P8 [2]

File No.: 2100561.303 Revision: 2 Page 32 of 38 F0306-01R4 Figure 11. Through-Wall Stress Distribution for Path P9 [2]

File No.: 2100561.303 Revision: 2 Page 33 of 38 F0306-01R4 Figure 12. Through-Wall Stress Distribution for Path P10 [2]

File No.: 2100561.303 Revision: 2 Page 34 of 38 F0306-01R4 Figure 13. Through-Wall Stress Distribution for Path P11 [2]

File No.: 2100561.303 Revision: 2 Page 35 of 38 F0306-01R4 Figure 14. Through-Wall Stress Distribution for Path P12 [2]

File No.: 2100561.303 Revision: 2 Page 36 of 38 F0306-01R4 Figure 15. Through-Wall Stress Distribution for Path P13 [2]

File No.: 2100561.303 Revision: 2 Page 37 of 38 F0306-01R4 Figure 16. Weld Residual Stress Distribution Figure 17. Semi-Elliptical Axial Crack in a Cylinder Model Figure 18. Semi-Elliptical Circumferential Crack in a Cylinder Model

-10

-5 0

5 10 0

0.2 0.4 0.6 0.8 1

x/t Stress (ksi) p

File No.: 2100561.303 Revision: 2 Page 38 of 38 F0306-01R4 Figure 19. The Effect of Temperature on the Fracture Toughness, JIc, of SA-516 Grade 70 Steel [11]

File No.: 2100561.303 Revision: 2 Page A-1 of A-2 F0306-01R4 COMPUTER FILES LISTING

File No.: 2100561.303 Revision: 2 Page A-2 of A-2 F0306-01R4 Zip File Name Description Table6.zip 26 input and 26 output files for the results in Table 6. The filenames are the same as Case Identification.

Table7.zip 26 input and 26 output files for the results in Table 7. The filenames are the same as Case Identification.

Table8.zip 26 input and 26 output files for the results in Table 8. The filenames are the same as Case Identification.

Table9.zip 26 input and 26 output files for the results in Table 9. The filenames are the same as Case Identification.

Table10A.zip 12 input and 12 output files for the results in Table 10 for PSI+10+20+30+40+50+60+70. The filenames are the same as Case Identification.

Table10B.zip 12 input and 12 output files for the results in Table 10 for PSI+10+20+30+40+70. The filenames are the same as Case Identification.

Table4.zip 26 input and 26 output files for the results in Table 4. The filenames are the same as Case Identification.