RS-24-024, Response to Request for Additional Information Regarding Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds

Response to Request for Additional Information Regarding Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds

ML24082A185
Person / Time
Site:	Braidwood, Byron
Issue date:	03/22/2024
From:	Humphrey M Constellation Energy Generation
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
RS-24-024
Download: ML24082A185 (1)

Text

4300 Winfield Road Warrenville, IL 60555

www.constellation.com

RS-24-024 10 CFR 50.55a

March 22, 2024

ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Braidwood Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. 50-456 and 50-457

Byron Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. 50-454 and 50-455

SUBJECT:

Response to Request for Additional Information regarding Proposed Alternative for Exam ination of Pressurizer Circum ferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds (L-2023-LLR-0062)

REFERENCES:

1. Letter from Kevin Lueshen (Constellation Energy Generation, 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 November 1, 2023 (ML23305A069)

2. Electronic Mail from Scott Wall (Nuclear Regulatory Commission) to Christian Williams (Constellation Energy Generation, LLC) Constellation Energy Generation, LLC - Fleet Request - Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds (L-2023-LLR-0062) Dated February 22, 2024 (ML24053A338).

By letter dated November 1, 2023 (Reference 1), Constellation Energy Generation, LLC (CEG, the licensee) requested a proposed alternative to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI, at Braidwood Station, Units 1 and 2 (Braidwood), and Byron Station, Units 1 and 2 (Byron).

Specifically, the proposed alternatives are related to the volumetric examination of pressurizer (PZR) circumferential and longitudinal shell-to-head welds and nozzle-to-shell welds. The proposed alternative originally requested to extend the inspection interval frequency from 10 years to the end of the 5 th inservice inspection interval.

By electronic mail dated February 22, 2024 (Reference 2), the NRC requested additional information that was necessary to complete its review.

U.S. Nuclear Regulatory Commission March 22, 2024 Page 2

Specifically, CEG was requested to demonstrate how the proposed performance monitoring plan ensured that a sample of at least 25% of ASME Code required examinations are conducted.

The proposed performance monitoring plan included in Attachment 2 of Reference 1 has been modified to ensure that a sample of at least 25% of the ASME Code required examinations are conducted. This includes the addition of three (3) weld examinations.

The revised performance monitoring plan is provided in the attached revision to the proposed alternative. Additionally, the duration of the proposed alternative is being revised from the 5th inservice inspection interval to the end of the current operating licenses for Byron Unit 1 and 2 and Braidwood Unit 1 and 2. This revised performance monitoring plan will ensure that a sample of at least 25% of the ASME Code required examinations are conducted each inspection interval between Byron Unit 1 and 2 and Braidwood Unit 1 and 2.

This submittal contains a regulatory commitment provided in Attachment 1.

The Summary of Commitments provided as Attachment 1 to this letter shall replace of Reference 1 in its entirety.

The 10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 provided as to this letter shall replace Attachment 2 of Reference 1 in its entirety.

If you should have any questions regarding this submittal, please contact Christian Williams at Christian.Williams@Constellation.com.

Respectfully, Humphrey, Mark Digitally signed by Humphrey, Mark D.

D. Date: 2024.03.22 11:52:50 -05'00' Mark Humphrey Senior Manager, Licensing Constellation Energy Generation, LLC

Attachments: 1) Summary of Commitments

2) 10 CFR 50.55a Proposed Alternative 14R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative 14R-21 for Byron Station, Units 1 and 2, Revision 1

cc: NRC Regional Administrator - Region III NRC Senior Resident Inspector - Braidwood Station NRC Senior Resident Inspector - Byron Station NRC Project Manager - Braidwood Station NRC Project Manager - Byron Station Illinois Emergency Management Agency - Division of Nuclear Safety Attachment 1

Summary of Commitments

Summary of Commitments Page 1 of 1

Attachment 1

Summary of Commitments

The following table identifies commitments made in this document. (Any other actions discussed in the submittal represent intended or planned actions. They are described to the NRC for the NRCs information and are not regulatory commitments.)

COMMITMENT TYPE COMMITMENT COMMITTED DATE ONE-TIME Programmatic OR OUTAGE ACTION (Yes/No)

(Yes/No)

As part of the performance monitoring plan, The required Yes No during the 5th interval at Byron Station, Unit examinations will be 2, all ASME Section XI Category B-B and completed in the 5th Category B-D examinations applicable to interval at Byron Unit the pressurizer will be performed to the 2 and the 6th interval maximum extent possible. The components at Braidwood Unit 1 to be examined are provided in Section 1 of or Unit 2.

the proposed alternative.

During the 6th interval at Braidwood Station Unit 1 or Unit 2, all ASME Section XI Category B-B and Category B-D examinations applicable to the pressurizer will be performed to the maximum extent possible. The components eligible for examination are provided in Section 1 of the proposed alternative.

Any new unacceptable indications identified as part of the performance monitoring plan will result in welds being examined at the other three units as described in the proposed alternative.

Attachment 2

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 1 of 53)

Proposed Alternative for Examination of Pressurizer Shell-to-Head and Nozzle-to-Shell Welds (Examination Categories B-B and B-D)

In Accordance with 10 CFR 50.55a(z)(1)

1 ASME Code Component(s) Affected

Code Class: Class 1

Description:

Pressurizer Shell-to-Head and Nozzle-to-Vessel Welds Examination Categories: Category B-B, Pressure-Retaining Welds in Vessels Other Than Reactor Vessels 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

Braidwood Component IDs:

Unit ASME ASME Category Item Component ID Component Description 1 B-B B2.11 1PZR-01-08A Shell - Lower Head 1 B-B B2.11 1PZR-01-08E Shell - Upper Head 2 B-B B2.11 2PZR-01-08A Shell - Lower Head 2 B-B B2.11 2PZR-01-08E Shell - Upper Head 1 B-B B2.12 1PZR-01-09A Shell Longitudinal Weld 1 B-B B2.12 1PZR-01-09D Shell Longitudinal Weld 2 B-B B2.12 2PZR-01-09A Shell Longitudinal Weld 2 B-B B2.12 2PZR-01-09D Shell Longitudinal Weld 1 B-D B3.110 1PZR-01-N1 Surge Nozzle 1 B-D B3.110 1PZR-01-N2 Spray Nozzle 1 B-D B3.110 1PZR-01-N3 Relief Nozzle 1 B-D B3.110 1PZR-01-N4A Safety Nozzle 1 B-D B3.110 1PZR-01-N4B Safety Nozzle 1 B-D B3.110 1PZR-01-N4C Safety Nozzle 2 B-D B3.110 2PZR-01-N1 Surge Nozzle 2 B-D B3.110 2PZR-01-N2 Spray Nozzle 2 B-D B3.110 2PZR-01-N3 Relief Nozzle 2 B-D B3.110 2PZR-01-N4A Safety Nozzle 2 B-D B3.110 2PZR-01-N4B Safety Nozzle 2 B-D B3.110 2PZR-01-N4C Safety Nozzle

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 2 of 53)

Byron Component IDs:

Unit ASME ASME Category Item Component ID Component Description 1 B-B B2.11 1RY-01-S/PC-01 Shell - Bottom Head 1 B-B B2.11 1RY-01-S/PC-05 Shell - Upper Head 2 B-B B2.11 2RY-01-S/PC-01 Shell - Bottom Head 2 B-B B2.11 2RY-01-S/PC-05 Shell - Upper Head 1 B-B B2.12 1RY-01-S/PL-01 Lower Longitudinal Weld 1 B-B B2.12 1RY-01-S/PL-04 Upper Longitudinal Weld 2 B-B B2.12 2RY-01-S/PL-01 Lower Longitudinal Weld 2 B-B B2.12 2RY-01-S/PL-04 Upper Longitudinal Weld 1 B-D B3.110 1RY-01-S/PN-01 Surge Nozzle 1 B-D B3.110 1RY-01-S/PN-02 Spray Nozzle 1 B-D B3.110 1RY-01-S/PN-03 Relief Nozzle 1 B-D B3.110 1RY-01-S/PN-04 Safety Nozzle 1 B-D B3.110 1RY-01-S/PN-05 Safety Nozzle 1 B-D B3.110 1RY-01-S/PN-06 Safety Nozzle 2 B-D B3.110 2RY-01-S/PN-01 Surge Nozzle 2 B-D B3.110 2RY-01-S/PN-02 Spray Nozzle 2 B-D B3.110 2RY-01-S/PN-03 Relief Nozzle 2 B-D B3.110 2RY-01-S/PN-04 Safety Nozzle 2 B-D B3.110 2RY-01-S/PN-05 Safety Nozzle 2 B-D B3.110 2RY-01-S/PN-06 Safety Nozzle

2 Applicable Code Edition and Addenda

The following table identifies the American Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel (B&PV)Section XI Code of Record for performing Inservice Inspection (ISI) activities at Braidwood and Byron:

PLANT INTERVAL EDITION START1 END1

Braidwood Station, Edition August 29, 2018 (Unit1) July 28, 2028 (Unit 1)

Units 1 and 2 Fourth 2013 November 5, 2018 (unit 2) October 16, 2028 (Unit 2)

Established 18 months Braidwood Station, prior to 5th interval start July 29, 2028 (Unit 1) July 28, 2038 (Unit 1)

Units 1 and 2 Fifth date in accordance October 17, 2028 (Unit 2) October 16, 2038 (Unit 2) with 10 CFR 50.55a(g)(4)(ii)

Byron Station, Units 1 and 2 Fourth 2007 Edition, through 2008 Addenda July 16, 2016 July 15, 2025

Byron Station, Units 1 and 2 Fifth 2019 Edition July 16, 2025 July 15, 2035 Note 1: The fifth interval start and end dates are estimates based on the current interval.

Interval start and end dates and may be adjusted as allowed by ASME Section XI. Start and end dates for the sixth interval, as well as the ASME XI code of record, will be determined in accordance with the requirements of 10 CFR 50.55a(g)(4)(ii).

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 3 of 53)

The 2019 Edition of ASME Section XI, Table G-2110-1 will be utilized to extend the use of Figure G-2110-1, Reference Critical Stress Intensity Factor for Material, to material SA-533 Grade A, Class 2. (Note: The 2019 Edition of ASME Section XI is approved in 10 CFR 50.55a.)

3 Applicable Code Requirement

ASME Code,Section XI, IWB-2500(a), Table IWB-2500-1, Examination Category B-B, requires examination of the applicable Item Numbers as follows:

Item No. B2.11 - Volumetric examination of 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 foot of one Longitudinal Shell-to-Head weld intersecting the circumferential weld per head each inspection interval. The examination volume is shown in Figure IWB-2500-2.

ASME Code,Section XI, IWB-2500(a), Table IWB-2500-1, Examination Category B-D, requires examination of the applicable Item Number as follows:

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), (c), and (d).

4 Reason for Request

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

[1] of the basis for the ASME Section XI examination requirements specified for the above-listed ASME Code,Section XI, Division 1 (ASME Section XI) Examination Categories and Item Numbers for pressurizer welds. The assessments include a survey of inspection results from 74 units as well as flaw tolerance evaluations using probabilistic fracture mechanics (PFM) and deterministic fracture mechanics (DFM). The Reference [1] report results indicate that the current ASME Section XI inspection frequency of ten years for these welds can be increased with no impact to plant safety. It is upon the basis of those results that an alternate inspection frequency is requested.

5 Proposed Alternative and Basis for Use

Constellation requests an inspection alternative to the examination requirements of ASME Section XI, Table IWB-2500-1, for Examination Categories B-B and B-D, Item Numbers B2.11, B2.12, and B3.110. The proposed alternative is to defer inspection of these Item Numbers through the end of the currently approved Period of Extended Operation (PEO) for Braidwood Station (Braidwood), Units 1 and 2, and Byron Station (Byron), Units 1 and 2 as summarized in Table 1.

The NRC has stated during several public meetings that PFM is inherently risk-10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 4 of 53)

informed and consequently, any submittal that uses PFM technology as its basis will be reviewed as a risk informed submittal. This proposed alternative relies on PFM technology as part of the technical basis. Therefore Revision 1 of the proposed alternative includes a performance monitoring plan which will provide direct evidence to the presence and extent of any degradation, confirm the continued adequacy of the analysis, and provide a timely method to detect any novel or unexpected degradation.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 5 of 53)

Table 1. Summary of Inspection Deferrals in this Proposed Alternative

End of Current Length of Item Operating Period Requested Station Unit ASME Category No. Description Date of Last Inspection (60 Years) Deferral for This Request (Years)

Braidwood 1 B-B B2.11 Pressurizer, Shell-to-Head 10/12/2019 10/17/2046 27.01 Welds, Circumferential B-B B2.12 Pressurizer, shell-to-Head 10/12/2019 27.01 Welds, Longitudinal B-D B3.110 Pressurizer, Nozzle-to-Vessel 4/8/2021 25.01 Welds Braidwood 2 B-B B2.11 Pressurizer, Shell-to-Head 10/11/2018 12/18/2047 29.21 Welds, Circumferential B-B B2.12 Pressurizer, shell-to-Head 10/11/2018 29.21 Welds, Longitudinal B-D B3.110 Pressurizer, Nozzle-to-Vessel 10/11/2018 29.21 Welds Byron 1 B-B B2.11 Pressurizer, Shell-to-Head 9/18/2015 10/31/2044 29.1 Welds, Circumferential B-B B2.12 Pressurizer, shell-to-Head 9/18/2015 29.1 Welds, Longitudinal B-D B3.110 Pressurizer, Nozzle-to-Vessel 9/14/2018 26.1 Welds Byron 2 B-B B2.11 Pressurizer, Shell-to-Head 10/6/2014 11/6/2046 32.12 Welds, Circumferential B-B B2.12 Pressurizer, shell-to-Head 10/6/2014 32.12 Welds, Longitudinal B-D B3.110 Pressurizer, Nozzle-to-Vessel 4/13/2019 27.62 Welds Notes:

1 - Performance monitoring examinations will be performed during the 6th interval at Braidwood Unit 1 or Unit 2 with this proposed alternative.

2 - Performance monitoring examinations will be performed during the 5th interval at Byron Unit 2 with this proposed alternative.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 6 of 53)

The key aspects of the technical basis for this request are summarized below. The applicability of the technical basis to Braidwood and Byron is shown in Appendix A.

It is noted that the inputs for transients and cycle counting remain valid for this Revision 1 submittal.

Degradation Mechanism Evaluation

An evaluation of degradation mechanisms that could potentially impact the reliability of the pressurizer welds was performed in Reference [1]. Evaluated mechanisms included stress corrosion cracking (SCC), environmental assisted fatigue (EAF),

microbiologically influenced corrosion (MIC), pitting, crevice corrosion, erosion-cavitation, erosion, flow accelerated corrosion (FAC), general corrosion, galvanic corrosion, and mechanical/ thermal fatigue. Other than the potential for EAF and mechanical/thermal fatigue, there are no active degradation mechanisms identified that significantly affect the long-term structural integrity of the pressurizer welds.

Stress Analysis

Finite element analyses (FEA) were performed in Reference [1] to determine the stresses in the subject pressurizer welds. The analyses were performed using representative pressurized water reactor (PWR) geometries, representative transients, and typical material properties. The results of the stress analyses were used to produce flaw tolerance evaluations. The applicability of the FEA to Braidwood and Byron in accordance with Section 9 of Reference [1] is shown in Appendix A and confirms that all plant-specific applicability requirements are satisfied. Therefore, the evaluation results and conclusions of Reference [1] are applicable to Braidwood and Byron.

Flaw Tolerance Evaluation

Flaw tolerance evaluations were performed in Reference [1] consisting of PFM and DFM evaluations. The results of the PFM analyses indicate that, after a preservice inspection (PSI), no other inspections are required for up to 60 years of plant operation to meet the U.S. Nuclear Regulatory Commissions (NRCs) safety goal of 10-6 failures per year. For the case of Braidwood and Byron, PSI was followed by three full 10-year interval inspections, which have been performed on the subject pressurizer welds. Table 8-12 of Reference [1] indicates that if PSI are followed by at least three full 10-year interval inspections subsequent examinations do not need to be performed for up to 80 years of plant operation, and they will still meet the NRC safety goal (with considerable margin). The DFM evaluations confirm the PFM results by demonstrating that it takes approximately 400 years for a postulated flaw with an initial depth equal to the ASME Section XI acceptance standards to grow to 80% of the wall thickness without exceeding the ASME Section XI allowable fracture toughness.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 7 of 53)

Inspection History

Plant operating experience (including examinations performed to-date, examination findings, inspection coverage, and previously submitted Proposed Alternatives) is summarized in Tables 2 and 3. As shown in these tables, some previous examinations for the subject welds had limited coverage because of limited volumetric examination scan access due to existing plant obstructions.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 8 of 53)

Table 2. Braidwood Station, Units 1 and 2, Pressurizer Welds Inspection History

Weld ID PSI1 Interval 1 Interval 2 Interval 3 Interval 4 Unit Summary Number Ultrasonic Test (UT) (Date/Outage, (Date/Outage, (Date/Outage, (Date/Outage, Exam Cat, Item Number (Date, Coverage/Results) Coverage/Results) Coverage/Results) Coverage/Results) Coverage/Results) 1PZR-01-08E 8-1985 10-1995/A1R05 4-2006/A1R12 9-2016/A1R19 Not required per 1 BRW-1-B02.11.0005 Coverage was not documented; but 100%/NRI 100%/NRI 98.5%/NRI ML22307A246 B-B/B2.11 obstructions noted/NRI 1PZR-01-08A 8-1985 10-1989/A1R01 9-1998/A1R07 4-2009/A1R14 10-2019/A1R21 1 BRW-1-B02.11.0006 100%/Laminar type flaws noted 100%/NRI 95.3%/NRI 95.3%/NRI 95.3%/NRI B-B/B2.11 1PZR-01-09A 8-1985 10-1989/A1R01 10-1998/A1R07 4-2009/A1R14 10-2019/A1R21 1 BRW-1-B02.12.0005 100%/Laminations/slag 100%/NRI 100%/NRI 100%/NRI 100%/NRI B-B/B2.12 1PZR-01-09D 8-1985 10-1995/A1R05 4-2006/A1R12 9-2016/A1R19 Not required per 1 BRW-1-B02.12.0006 100%/NRI 100%/NRI 100%/NRI 100%/NRI ML22307A246 B-B/B2.12 1PZR-01-N4A 8-1985 10-1995/A1R053 4-2006/A1R12 9-2016/A1R19 Not required per 1 BRW-1-B03.110.0013 Coverage was not documented; Coverage was not 100% Shell Side only due 68.7%/NRI ML22307A246 B-D/B3.110 limitations due to Geometry/NRI documented/NRI to config/NRI 1PZR-01-N4B 8-1985 1 BRW-1-B03.110.0014 Coverage was not documented; 10-1995/A1R05 10-2004/A1R11 4-2015/A1R18 Not required per B-D/B3.110 limitations due to Geometry/Laminar 100%/NRI 92.66%/NRI 90.9%/NRI ML22307A246 flaw 1PZR-01-N1 8-1985 A1R06 10-2007/A1R132 9-2016/A1R192 Not required per 1 BRW-1-B03.110.0015 Coverage was not documented; (VT-2 per Relief Request 59.2%/NRI 74.7%/NRI ML22307A246 B-D/B3.110 limitations due to Geometry/NRI per NR-24) 1PZR-01-N2 8-1985 10-1989/A1R013 9-1998/A1R073 1 BRW-1-B03.110.0016 Coverage was not documented; Coverage was not Coverage was not 10-2010/A1R15 4-2021/A1R22 B-D/B3.110 limitations due to Geometry/NRI documented/NRI documented; geometry 56.56%/NRI 56.43%/NRI limitations/NRI 1PZR-01-N3 8-1985 10-1989/A1R013 9-1998/A1R073 1 BRW-1-B03.110.0017 Coverage was not documented; Coverage was not Coverage was not 10-2010/A1R15 4-2021/A1R22 B-D/B3.110 limitations due to Geometry/NRI documented/NRI documented; geometry 60.92%/NRI 60%/NRI limitations/NRI 1PZR-01-N4C 8-1985 10-1995/A1R053 1 BRW-1-B03.110.0018 Coverage was not documented; Coverage was not 10-2004/A1R11 4-2015/A1R18 Not required per B-D/B3.110 limitations due to Geometry/NRI documented; geometry 92.66%/NRI 90.9%/NRI ML22307A246 limitations/NRI 2PZR-01-08A 1-1987 4-1990/A2R013 5-1999/A2R07 2 BRW-2-B02.11.0005 Coverage was not documented; Coverage was not 99.85%/Laminations 10-2009/A2R14 Not required per B-B/B2.11 limitations due to obstructions/ documented/NRI noted 95.57%/NRI ML22307A246 Laminar and Planar flaws noted 2PZR-01-08E 12-1986 2 BRW-2-B02.11.0006 Coverage was not documented; 4-1996/A2R05 10-2006/A2R12 10-2018/A2R20 Not required per B-B/B2.11 obstructions noted/Laminar and 100%/NRI 100%/NRI 100%/NRI ML22307A246 Planar flaws noted 10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 9 of 53)

Table 2. Braidwood Station, Units 1 and 2, Pressurizer Welds Inspection History

Weld ID PSI1 Interval 1 Interval 2 Interval 3 Interval 4 Unit Summary Number Ultrasonic Test (UT) (Date/Outage, (Date/Outage, (Date/Outage, (Date/Outage, Exam Cat, Item Number (Date, Coverage/Results) Coverage/Results) Coverage/Results) Coverage/Results) Coverage/Results) 2PZR-01-09A 1-1987 4-1990/A2R01 5-1999/A2R07 10-2009/A2R14 Not required per 2 BRW-2-B02.12.0005 100%/NRI 100%/NRI 100%/NRI 100%/NRI ML22307A246 B-B/B2.12 2PZR-01-09D 12-1986 4-1996/A2R05 10-2006/A2R12 10-2018/A2R20 Not required per 2 BRW-2-B02.12.0006 100%/Laminar type flaws noted 100%/NRI 100%/NRI 100%/NRI ML22307A246 B-B/B2.12 2PZR-01-N1 1-1987 A2R06 5-2008/A2R13 4-2017/A2R192 Not required per 2 BRW-2-B03.110.0013 100%/NRI (VT-2 per Relief Request 59.2%/NRI 59.2%/NRI ML22307A246 B-D/B3.110 NR-24) 2PZR-01-N2 12-1986 4-1996/A2R05 4-1999/A2R07 4-2011/A2R15 Not required per 2 BRW-2-B03.110.0014 100%/NRI 100%/NRI 88.5%/NRI 88.5%/NRI ML22307A246 B-D/B3.110 2PZR-01-N3 12-1986 4-1990/A2R013 2 BRW-2-B03.110.0015 Coverage was not documented; Coverage was not 4-1999/A2R07 4-2011/A2R15 10-2021/A2R22 B-D/B3.110 geometry limitations/NRI documented; geometry 88.5%/NRI 88.5%/NRI 71.6%/NRI limitations /NRI 2PZR-01-N4A 12-1986 4-1990/A2R013 2 BRW-2-B03.110.0016 Coverage was not documented; Coverage was not 10-2006/A2R12 10-2018/A2R20 Not required per B-D/B3.110 nozzle geometry limitations/NRI documented; geometry 91.5%/NRI 91.5%/NRI ML22307A246 limitations /NRI 2PZR-01-N4B 12-1986 3-1996/A2R053 2 BRW-2-B03.110.0017 Coverage was not documented; Coverage was not 11-2003/A2R10 5-2014/A2R17 Not required per B-D/B3.110 nozzle geometry limitations /Laminar documented; geometry 88.5%/NRI 88.5%/NRI ML22307A246 type flaw limitations /NRI 2PZR-01-N4C 12-1986 3-1996/A2R053 11-2003/A2R103 2 BRW-2-B03.110.0018 Coverage was not documented; Coverage was not Coverage was not 5-2014/A2R17 Not required per B-D/B3.110 nozzle geometry limitations /NRI documented; geometry documented; geometry 88.5%/NRI ML22307A246 limitations /NRI limitations/NRI Notes:

1. PSI included radiographic (RT) examinations per Section III and ultrasonic (UT) examinations per Section XI. Together, these examinations constituted PSI examination in Reference [1], and 100% coverage was assumed because such coverage was required by Section III for the RT exams, and the RT exams were successfully completed for the subject pressurizer welds for both Braidwood, Units 1 and 2.

2. The increase in examination coverage from the Second Interval to the Third Interval for weld 1PZR-01-N1 was not a result of a change in physical limitations, but was due to a change in the methods of calculating coverage. For Unit 1 the actual coverage reported in 2007 (A1R13) utilized the most conservative angle for the total coverage (60° axial = 59.2%). In 2016 (A1R19) the coverage utilized an aggregate of all angles 0°, 45° axial and circumferential and 60° axial and circumferential to achieve 74.7% coverage. For Unit 2 during the exam in 2017 (A2R19) for the similar weld, 2PZR-01-N1, coverage was not recalculated, and the same coverage was recorded as the prior interval examination. The limitations are the same between Unit 1 and Unit 2 and the reported Third Interval coverage for weld 2PZR-01-N1 should be similar to the Unit 1 Third Interval coverage at 74.7%.

3. Constellation has reviewed the missing coverages for the examinations in the First and Second Intervals and has concluded that the NDE code requirements and equipment for early interval examinations would not have resulted in an examination coverage lower than the minimum achieved coverage of 56.43%. In addition, there have been no design configuration changes that would alter examination access for any of the welds.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 10 of 53)

During the Fourth Interval, the Braidwood, Unit 1, pressurizer spray nozzle weld 1PZR-01-N2 had the minimum coverage of 56.43%. Section 8.3.5 and Table 8-33 of Reference [1] discuss the limited coverage and show that the conclusions of the report are applicable to components with limited coverage as low as 50%. The minimum coverage of 56.43% for this weld is higher than the 50% minimum coverage assumed in the sensitivity study of the base case in the EPRI report; therefore, the sensitivity results from the EPRI report are bounding for application to Braidwood, Units 1 and 2.

No flaws that exceeded the ASME Section XI acceptance standards were identified during any prior examinations for Braidwood.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 11 of 53)

Table 3. Byron Station, Units 1 and 2, Pressurizer Welds Inspection History Weld ID PSI1 Interval 1 Interval 2 Interval 3 Interval 4 Unit Summary Number Ultrasonic Test (UT) (Date/Outage, (Date/Outage, (Date/Outage, (Date/Outage, Exam Cat, Item (Date, Coverage/Results) Coverage/Results) Coverage/Results) Coverage/Results) Coverage/Results)

Number 4-1987 & 4-1987/B1R014 Coverage was not documented; scan limitations 8-1983 due to sampling nozzles and 3-2002/B1R11 Coverage was not documented; X axis plates/ 92%/5 embedded scan limitations (welded pads) Resizing of 5 subsurface indications identified (1 1RY-01-S/PC-012 were identified/ indications identified in PSI with 45° and 4 with 60°) 3-2011/B1R17 Not required per 1 BYR-1-B2.11.0001 2 rejectable 45° indications and 92%/NRI ML22307A246 B-B/B2.11 3 rejectable 60° indications 1-1990/B1R03 Fourth exam showed (total of 5 indications that did Resizing of 5 subsurface no change in indication not meet the requirements of indications identified in B1R01 sizes.

ASME Section XI, IWB-3511 4-1996/B1R07 Resizing of 5 subsurface indications identified in B1R03 1RY-01-S/PC-05 8-1982 1 BYR-1-B2.11.0002 Coverage was not documented; 4-1996/B1R07 3-2005/B1R13 9-2015/B1R20 Not required per B-B/B2.11 scan limitations (welded pads) 96%/NRI 96%/NRI 95.47%/NRI ML22307A246 were identified/NRI 1RY-01-S/PL-01 8-1982 4-1987/B1R01 1 BYR-1-B2.12.0001 Coverage was not documented; 90% - limited by sample 3-2002/B1R11 3-2011/B1R17 Not required per B-B/B2.12 no scan limitations were nozzles and axis plates/NRI 100%/NRI 100%/NRI ML22307A246 identified/NRI 1RY-01-S/PL-04 8-1982 1 BYR-1-B2.12.0002 Coverage was not documented; 4-1996/B1R07 3-2005/B1R13 9-2015/B1R20 Not required per B-B/B2.12 no scan limitations were 100%/NRI 100%/NRI 100%/NRI ML22307A246 identified/NRI 7-1982 9-2006/B1R143 9-2018/B1R223 1RY-01-S/PN-01 Coverage was not documented; B1R07 (VT-2 per Relief Request B1R11 (VT-2 per Relief 40% - limited by nozzle 72.03% - limited by 1 BYR-1-B3.110.0001 no scan limitations were NR-19) Request I2R-03) configuration and heater nozzle configuration B-D/B3.110 identified/NRI penetrations/NRI and heater penetrations/NRI 10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 12 of 53)

Table 3. Byron Station, Units 1 and 2, Pressurizer Welds Inspection History Weld ID PSI1 Interval 1 Interval 2 Interval 3 Interval 4 Unit Summary Number Ultrasonic Test (UT) (Date/Outage, (Date/Outage, (Date/Outage, (Date/Outage, Exam Cat, Item (Date, Coverage/Results) Coverage/Results) Coverage/Results) Coverage/Results) Coverage/Results)

Number 7-1982 1RY-01-S/PN-02 Coverage was not documented; 1-1990/B1R03 11-1997/B1R08 9-2006/B1R14 1 BYR-1-B3.110.0002 no scan limitations were 75% - limited by nozzle 73% - limited by nozzle 77% - limited by nozzle Not required per B-D/B3.110 identified/ configuration/NRI with 0° and configuration/NRI configuration/NRI ML22307A246 Indications - appears to be ID 45°. ID Geometry with 60° geometry 1RY-01-S/PN-03 7-1982 11-1997/B1R08 9-2006/B1R14 1 BYR-1-B3.110.0003 Coverage was not documented; 3-1987/B1R01 69% - limited by nozzle 77% - limited by nozzle Not required per B-D/B3.110 no scan limitations were 100%/NRI configuration/NRI configuration/NRI ML22307A246 identified/NRI 7-1982 1RY-01-S/PN-04 Coverage was not documented; 1-1990/B1R03 3-2002/B1R11 9-2015/B1R20 1 BYR-1-B3.110.0004 no scan limitations were 75% - limited by nozzle 66% - limited by nozzle 64.93% - limited by nozzle Not required per B-D/B3.110 identified configuration/NRI with 0° and configuration/NRI configuration/NRI ML22307A246 Indications - appears to be ID 45°. ID Geometry with 60° geometry 9-2006/B1R14 7-1982 68% - limited by nozzle 1RY-01-S/PN-05 Coverage was not documented; 4-1996/B1R07 11-1997/B1R08 configuration/NRI Not required per 1 BYR-1-B3.110.0005 no scan limitations were 100%/NRI 73% - limited by nozzle ML22307A246 B-D/B3.110 identified/NRI configuration/NRI 9-2015/B1R20 64.93% - limited by nozzle configuration/NRI 7-1982 Coverage was not documented; 9-1988/B1R02 1RY-01-S/PN-06 no scan limitations were 76% (0°) 75% (45°), 75% (60°) 11-1997/B1R08 9-2006/B1R14 Not required per 1 BYR-1-B3.110.0006 identified/ limited due to 73% - limited by nozzle 77% - limited by nozzle ML22307A246 B-D/B3.110 Indications - appears to be ID configuration/Spot indication configuration/NRI configuration/NRI geometry and one spot indication similar to PSI (seen with 45°)

10-1985 1-1989/B2R014 3-2004/B2R11 2RY-01-S/PC-01 Coverage was not documented; Coverage was not 92% - scan limitations 2 BYR-2-B2.11.0001 scan limitations (welded pads, documented; scan limitations (welded pads, 10-2014/B2R18 Not required per B-B/B2.11 sampling nozzle and skirt) were (welded pads, sampling sampling nozzle and 94%/NRI ML22307A246 identified/NRI nozzle and skirt) were skirt)/NRI identified/NRI 10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 13 of 53)

Table 3. Byron Station, Units 1 and 2, Pressurizer Welds Inspection History Weld ID PSI1 Interval 1 Interval 2 Interval 3 Interval 4 Unit Summary Number Ultrasonic Test (UT) (Date/Outage, (Date/Outage, (Date/Outage, (Date/Outage, Exam Cat, Item (Date, Coverage/Results) Coverage/Results) Coverage/Results) Coverage/Results) Coverage/Results)

Number 10-1985 3-1995/B2R054 2RY-01-S/PC-05 Coverage was not documented; Coverage was not 9-2005/B2R12 4-2010/B2R15 Not required per 2 BYR-2-B2.11.0002 scan limitations (welded pads) documented; scan limitations 96%/NRI 96%/NRI ML22307A246 B-B/B2.11 were identified/NRI (welded pads) were identified/NRI 10-1985 Coverage was not documented; 1-1989/B2R014 2RY-01-S/PL-01 no scan limitations were 2 BYR-2-B2.12.0001 identified/ Coverage was not 3-2004/B2R11 10-2014/B2R18 Not required per B-B/B2.12 Indications - 0° identified 4 documented; no scan 100%/NRI 100%/NRI ML22307A246 embedded indications, 45° was NRI, limitations were identified/NRI 60° identified 2 embedded indications.

2RY-01-S/PL-04 10-1985 3-1995/B2R054 2 BYR-2-B2.12.0002 Coverage was not documented; no Coverage was not 9-2005/B2R12 4-2010/B2R15 Not required per B-B/B2.12 scan limitations were documented; no scan 100%/NRI 100%/NRI ML22307A246 identified/NRI limitations were identified/NRI 10-1985 4-2019/B2R213 Coverage was not documented; no 4-2007/B2R133 2RY-01-S/PN-01 scan limitations were identified/ 2-1989/B2R01 11-1999/B2R08 40% - limited by nozzle 40% - limited by 2 BYR-2-B3.110.0001 Indications - 0° was NRI, 45° (VT-2 per Relief Request NR-19) (VT-2 per Relief configuration and heater nozzle configuration B-D/B3.110 identified 1 embedded indication, Request I2R-03) penetrations/NRI and heater 60° identified 1 embedded penetrations/NRI indication.

2RY-01-S/PN-02 10-1985 9-1993/B2R044 3-2004/B2R11 4-2010/B2R15 2 BYR-2-B3.110.0002 Coverage was not documented; no Coverage was not 62.3% - limited by 62.3% - limited by nozzle Not required per B-D/B3.110 scan limitations were identified/NRI documented; no scan nozzle configuration/NRI ML22307A246 limitations were identified/NRI configuration/NRI 2RY-01-S/PN-03 10-1985 10-1990/B2R02 4-2001/B2R09 4-2010/B2R15 Not required per 2 BYR-2-B3.110.0003 Coverage was not documented; no 75% - limited by nozzle 66% - limited by nozzle 66.3% - limited by nozzle ML22307A246 B-D/B3.110 scan limitations were identified/NRI configuration/NRI configuration/NRI configuration/NRI 2RY-01-S/PN-04 10-1985 9-1993/B2R044 3-2004/B2R11 10-2014/B2R18 2 BYR-2-B3.110.0004 Coverage was not documented; no Coverage was not 62% - limited by nozzle 62.5% - limited by nozzle Not required per B-D/B3.110 scan limitations were identified/NRI documented; no scan configuration/NRI configuration/NRI ML22307A246 limitations were identified/NRI 10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 14 of 53)

Table 3. Byron Station, Units 1 and 2, Pressurizer Welds Inspection History Weld ID PSI1 Interval 1 Interval 2 Interval 3 Interval 4 Unit Summary Number Ultrasonic Test (UT) (Date/Outage, (Date/Outage, (Date/Outage, (Date/Outage, Exam Cat, Item (Date, Coverage/Results) Coverage/Results) Coverage/Results) Coverage/Results) Coverage/Results)

Number 2RY-01-S/PN-05 10-1985 3-1995/B2R054 4-2001/B2R09 10-2014/B2R18 4-2019/B2R21 2 BYR-2-B3.110.0005 Coverage was not documented; no Coverage was not 66.3% - limited by nozzle 62.5% - limited by nozzle 66% - limited by B-D/B3.110 scan limitations were identified/NRI documented; no scan configuration/NRI configuration/NRI nozzle limitations were identified/NRI configuration/NRI 2RY-01-S/PN-06 10-1985 3-1995/B2R054 4-2001/B2R09 4-2010/B2R15 4-2019/B2R21 2 BYR-2-B3.110.0006 Coverage was not documented; no Coverage was not 66% - limited by nozzle 66.3% - limited by nozzle 62% - limited by B-D/B3.110 scan limitations were identified/NRI documented; no scan configuration/NRI configuration/NRI nozzle limitations were identified/NRI configuration/NRI Notes:

1. PSI included radiographic (RT) examinations per Section III and ultrasonic (UT) examinations per Section XI. Together, these examinations constituted PSI examination in Reference [1], and 100% coverage was assumed because such coverage was required by Section III for the RT exams, and the RT exams were successfully completed for the subject pressurizer welds for both Byron units.

2. Indications were found in weld 1RY-01-S/PC-01 that did not meet the acceptance standards of IWB-3511 during the initial PSI examination, but were accepted by engineering evaluation per IWB-3600 and three successive examinations were performed in B1R01, B1R03, and B1R07, which showed no changes in the flaw sizes.

3. The increase in examination coverage from the Third Interval to the Fourth Interval for weld 1RY-01-S/PN-01 was not a result of a change in physical limitations, but was due to a change in the methods of calculating coverage. For Unit 1 in 2018 (B1R22) the coverage calculated included the axial coverage of the examination which increased the coverage to 72.03%. For Unit 2 the exam in 2019 (B2R21) did not recalculate the coverage and conservatively assumed the same coverages as the 2007 (B2R13) exam, which did not include the axial coverage. The limitations are the same between Unit 1 and Unit 2 and the reported Fourth Interval coverage for weld 2RY-01-S/PN-01 should be similar to the Unit 1 Fourth Interval coverage at 72.03%.

4. Constellation has reviewed the missing coverages for the examinations in the First Interval and has concluded that the Code NDE requirements and equipment for early interval examinations would not have resulted in an examination coverage lower than the minimum documented coverage of 40%. In addition, there have been no design configuration changes that would alter examination access for any of the welds.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 15 of 53)

During Interval 3, the Byron, Units 1 and 2 pressurizer surge nozzle welds 1RY S/PN-01 and 2RY-01-S/PN-01 had the minimum documented coverage of 40%.

During Interval 4, the Byron, Unit 2 pressurizer surge nozzle weld 2RY-01-S/PN-01 again had a minimum documented coverage of 40%. (Note: The subsequent B1R22 examination for Weld ID 1RY-01-S/PN-01 recorded significantly more coverage at 72.03%, which was calculated by including the axial examination. The change in coverage was not a result of changes to physical limitations. The limitations are the same between Unit 1 and Unit 2 and the reported Fourth Interval coverage for weld 2RY-01-S/PN-01 should be similar to the Unit 1 Fourth Interval coverage at 72.03%). During the First and Second Intervals Relief Requests NR-19 and I2R-03 were approved by the NRC to perform VT-2 examination in lieu of UT for welds 1RY-01-S/PN-01 (Outage B1R07 and B1R11) and 2RY-01-S/PN-01 (Outages B2R01 and B2R08). Since the minimum coverage was identified for the surge line nozzle welds, a review of the probability of leakage and rupture for various sensitivity studies from Reference [1] was done for leak path PRNV-BW-1C, which is applicable to welds 1RY-01-S/PN-01 and 2RY-01-S/PN-01 as follows:

x Section 8.3.4.1.1 and Table 8-11 of Reference [1] discuss the probability of rupture and leakage for the case study if only PSI exams are performed.

The most limiting crack path for the surge line nozzle welds (PRNV-BW-1C),

yielded a probability of leakage of 6.25x10 -9 after 80 years and a probability of rupture of 1.25x10-9 after 80 years. The probability of leakage and the probability of rupture for this Case ID listed in Table 8-11 are approximately three orders of magnitude below the acceptance criteria utilized in Reference [1], Section 8.3.2.9, of 10 -6 failures per year. Additionally, the impact of performing more examinations than the PSI-only case study, even with limited coverage, which was the case for welds 1RY-01-S/PN-01 and 2RY-01-S/PN-01, would reduce the probability of rupture and leakage.

Furthermore, as discussed in Section 8.3.4.1.1 of Reference [1], leakage of the pressure boundary is detectable by plant operators and plant procedures allow for safe plant shutdown under leaking conditions.

Probabilities of rupture values are maintained well below the acceptance criterion for 80 years of operation even considering only PSI inspection. The examination results with coverage as low as 40% for welds 1RY-01-S/PN-01 and 2RY-01-S/PN-01 at Byron are bounded by the PSI-only case study in Reference [1] and the failure frequency for these welds would be below the NRC acceptance criteria of 10-6 failures per year.

x Section 8.3.5 and Table 8-33 of Reference [1] discuss the probability of leakage for the case study where only 50% coverage was achieved. For the leak path most applicable to welds 1RY-01-S/PN-01 and 2RY-01-S/PN-01, PRNV-BW-1C was evaluated to have a probability of leakage of 3.75x10 -9 after 80 years. This was an increase of 2.5x10 -9 compared to the base case where 100% coverage was achieved. This increase in the probability of leakage is small in comparison to the acceptance criteria of 10 -6 failures per year. The sensitivity studies performed in Reference [1], including the Inspection Coverage case study, show that probabilities of rupture and leakage are not significantly affected by using a range of values for most of the input variables.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 16 of 53)

Considering all of the findings from Reference [1] it can be concluded that the probability of failure is below 10 -6 failures per year for welds 1RY-01-S/PN-01 and 2RY-01-S/PN-01.

At Byron, weld 1RY-01-S/PC-01 had five indications that did not meet acceptance standards of IWB-3511 during the PSI examination, but all five indications were accepted through engineering evaluation in accordance with IWB-3600 of Section XI. Successive examinations were performed in B1R01 (1987), B1R03 (1990), and B1R07 (1996) which showed no change in the five indications. The most recent examination performed on weld 1RY-01-S/PC-01 during B1R17 (2011) did not identify any recordable indications. The absence of indications in this latest examination is due to improved NDE techniques and changes to sizing requirements. Therefore, the B1R17 examination determined that the previously reported indications meet the acceptance standards of ASME Section XI.

Based on the discussions above, the results of the EPRI report are applicable to Byron, Units 1 and 2.

Performance Monitoring Requirements

In a Request for Additional Information (RAI) dated February 22, 2024 (ML24053A338) the NRC requested Constellation to demonstrate how the previously proposed performance monitoring plan ensures that a sample of at least 25% of the ASME Code required examinations are conducted. This RAI is consistent with an April 27, 2023, public meeting (ML23114A034) and the June 15, 2023, industry technical exchange meetings (ML23158A180) during which the NRC stated that a performance monitoring sample of 25% will yield acceptable results, under certain assumptions. To address this RAI, the proposed performance monitoring plan included in the initial submittal dated November 1, 2023 (ML23305A069) is being replaced with the revised performance monitoring plan provided below.

Given the similarities in design, materials, construction methods, service conditions, and operating strategies between Braidwood and Byron, a common performance monitoring plan is being proposed. The proposed performance monitoring plan is for one (1) out of the four (4) units to complete all required ASME Section XI pressurizer examinations each ISI interval. This represents a 25% performance monitoring sample since one (of four required) pressurizer worth of examinations will be performed during each ISI interval. The units selected for performance monitoring will distribute the examinations across the periods as required by Table IWB-2411-1, as applicable. This will ensure a continuous stream of data and allow for timely identification of any service induced degradation or the emergence of any novel degradation mechanisms.

As previously committed (ML22307A246) Byron Unit 2 will be selected for performance monitoring examinations in the 5 th ISI interval during which all ASME Section XI, Category B-B and Category B-D examinations applicable to the pressurizer will be performed to the maximum extent possible. The components to be examined are shown in Section 1 of the proposed alternative and summarized below:

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 17 of 53)

o One (1) Category B-B, Item B2.11 top head circumferential weld o One (1) Category B-B, Item B2.12 top head longitudinal weld o One (1) Category B-B, Item B2.11 bottom head circumferential weld o One (1) Category B-B, Item B2.12 bottom head longitudinal weld o Six (6) Category B-D, Item B3.110 nozzle to vessel welds consisting of:

One (1) spray nozzle One (1) relief nozzle Three (3) safety nozzles One (1) surge nozzle

The performance monitoring examinations will be distributed among the periods as required by Table IWB-2411-1. The scheduling of the required examinations will be performed during the Byron 5th ISI interval code of record update, which is currently underway. The 5th ISI interval is currently scheduled to begin in July 2025 for Byron Unit 2.

Braidwood Unit 1 or Braidwood Unit 2 will be selected for performance monitoring examinations during the 6 th ISI interval. The specific unit and components selected for examination will be determined during the 6 th ISI interval code of record update and scheduled in the ISI database. The components eligible for selection as part of the performance monitoring plan are the same as described above for Byron Unit 2 and are shown in Section 1 of the proposed alternative.

A graphical representation of the performance monitoring plan is included in the Figure below. Distributing the examinations as required by IWB-2411-1 will ensure that examinations are performed in each inspection period shown in the Figure below for the selected Units at Byron and Braidwood. Distributing the performance monitoring examinations will ensure a continuous stream of examination data for the duration of the proposed alternative.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 18 of 53)

Graphical Representation of Proposed Performance Monitoring Plan

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 19 of 53)

In the unlikely event that any new unacceptable indications (i.e., new indications exceeding the acceptance standards of IWB-3500 that are accepted by Repair/Replacement Activity or analytical evaluation) are identified during the performance monitoring examinations, the indications will be evaluated as required by ASME Section XI, and the Constellation corrective action program. The additional examination and successive inspection requirements of ASME Code Section XI, also apply. The number of additional examinations, to be performed during the current refueling outage, shall be the number of performance monitoring examinations included in the inspection item number that were scheduled to be performed during the current inspection period. If additional unacceptable indications are identified, the examinations shall be further expanded to include all remaining welds in the inspection item number. Any new unacceptable indications identified as part of the performance monitoring examinations at Byron (5 th interval) or Braidwood (6th interval) will result in all four Units reverting to ASME Section XI examination requirements for the remainder of the interval during which the unacceptable indications were identified. ASME Section XI examinations shall resume no later than the first or second refueling outage following discovery of the initial unacceptable indication. The same weld with the initial unacceptable indication shall be included for examination at the remaining Byron and Braidwood Units within the first or second refueling outage following discovery. If the unacceptable indication is identified during the 5 th ISI interval, and no further unacceptable indications are identified during the resumption of ASME XI examinations, then the proposed alternative may continue to be applied for the 6 th ISI interval for all Byron and Braidwood Units.

If Braidwood or Byron pursue subsequent license renewal to operate beyond the end of the current operating licenses, the ASME Section XI examination requirements would again become applicable at that time. If Braidwood or Byron desire to use a similar proposed alternative in the second period of extended operation, it is required to be addressed in the license renewal application or a separate proposed alternative will be submitted in accordance with 10 CFR 50.55a(z), either of which will require separate NRC review and approval.

In addition to the direct evidence provided by the performance monitoring examinations at Byron (5th interval) and Braidwood (6 th interval), examination of pressurizer welds is expected to continue to be performed by other units across the domestic and international PWR fleet. Continued examination of these pressurizer welds across the industry will provide additional opportunities to detect known degradation mechanisms and will also provide the opportunity to detect any new or unexpected degradation mechanism that may occur in the future in the subject components. If a new degradation mechanism is identified during continued industry examinations, Constellation will follow the industry guidance to address the new degradation mechanism.

The absence of any new unacceptable indications in the Byron (5 th interval) and Braidwood (6th interval) performance monitoring examinations and the absence of any unexpected degradation across the operating fleet provides validation that the assumptions and methods of the PFM Model used in the EPRI Report are adequate to predict the future behavior of the subject welds. The strong technical basis provided by the results of the PFM Model and EPRI Report, along with the 10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 20 of 53)

implementation of the proposed performance monitoring plan, including scope expansion criteria, will provide additional assurance that the pressurizer components at Braidwood and Byron can operate safely for the remainder of the current operating licenses and will continue to provide an acceptable level of quality and safety.

Conclusion

Based on the results of Reference [1] and its demonstrated applicability to Braidwood and Byron, the subject pressurizer welds contained in this proposed alternative are very flaw tolerant. PFM and DFM evaluations performed as part of the technical basis in Reference [1] demonstrate that, after PSI, no other inspections are required for up to 60 years of operation to meet the NRC safety goal of 10-6 failures per reactor year. Plant-specific applicability of the technical basis to Braidwood and Byron is demonstrated in Appendix A. While the technical bases demonstrate longer inspection intervals are possible, Constellation considers that deferral of these inspections until the end of the currently approved period of extended operation, with the proposed performance monitoring plan, is justified and provides an acceptable level of quality and safety. The proposed performance monitoring plan will gather continuous data for Braidwood and Byron for the duration of the proposed alternative which will provide direct evidence of the presence and/or extent of degradation, validation of the continued adequacy of the analysis, and a timely method to detect novel or unexpected degradation.

The PWR fleet inspection history for the applicable components (obtained from an EPRI industry survey) is presented in Appendix B. The results of the survey indicate that these components are very flaw tolerant and prior instances of flaw detection are minimal.

Constellation operating and examination experience demonstrates that these welds have performed with very high reliability, mainly due to their robust designs and structural margins. As shown in the Tables 2 and 3, to-date, Constellation has performed over 150 inspections (PSI and ISI) of the subject pressurizer welds at Braidwood and Byron, with only one of these welds found to have recordable indications. The sizes of these indications have not increased since PSI. As indicated in the inspection history Tables 2 and 3, some of the examinations involved limited coverage of as low as 40%. However, due to the robust fabrication of these components and the intensive pre-service examinations performed, Reference [1] was able to conclude that performing only the PSI examination without any other follow-on ISI examinations is acceptable for up to 80 years of operation while still maintaining plant safety. In addition, it is important to note that all other inspection activities, including the ASME Section XI, Examination Category B-P system leakage test conducted during each refueling outage, will continue to be performed, providing further assurance of safety.

Finally, as discussed in Reference [2], for situations where no active degradation mechanism is present, it was concluded that subsequent inservice inspections do not provide additional value after the PSI has been performed. Braidwood and Byron pressurizer welds have received the required PSI examinations along with nearly 120 subsequent inservice inspections with no service-induced indications 10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 21 of 53)

documented.

Therefore, Constellation requests that the NRC authorize this proposed alternative for the Braidwood and Byron in accordance with 10 CFR 50.55a(z)(1).

6 Duration of Proposed Alternative

The proposed alternative is requested for Braidwood and Byron for the remainder of the currently approved operating licenses which are scheduled to end on October 17, 2046 (Braidwood Unit 1), December 18, 2047 (Braidwood, Unit 2),

October 31, 2044 (Byron, Unit 1), and November 6, 2046 (Byron, Unit 2).

7 Precedent

x Letter from N. Salgado (U.S. Nuclear Regulatory Commission) to D. Rhoades (Constellation Energy Generation), Braidwood Station, Units 1 and 2 and Bryon 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 10, 2022, ADAMS Accession No. ML22307A246.

x NRC Safety Evaluation Report, 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.

x NRC Safety Evaluation Report, Catawba Nuclear Station, Unit Nos. 1 and 2; Shearon Harris Nuclear Power Plant, Unit No. 1; McGuire Nuclear Station, Unit Nos. 1 and 2; Oconee Nuclear Station, Unit Nos. 1, 2, and 3; and H. B.

Robinson Steam Electric Plant, Unit No. 2 - Issuance of Alternative to Pressurizer Welds (EPID L-2023-LLR-0020), dated October 19, 2023, ADAMS Accession No. ML23264A853.

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.

x Based on studies presented in Reference [11], the NRC approved extending the SG vessel and nozzle welds from 10 to 30 years for Vogtle in Reference

[12].

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

[4].

x Based on work performed in BWRVIP-108, Reference [5], and BWRVIP-241, Reference [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 [6] and [8]. The work performed in BWRVIP-108 and BWRVIP-241 provided the technical basis for ASME Code Case N-702, Reference [9], which has been conditionally approved by the NRC in Revision 18 of Regulatory Guide 1.147, Reference [10].

8 Acronyms 10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 22 of 53)

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 ISI Inservice Inspection MIC Microbiologically influenced corrosion NPS Nominal pipe size NRC Nuclear Regulatory Commission NSSS Nuclear steam supply system PFM Probabilistic fracture mechanics PWR Pressurized Water Reactor SCC Stress corrosion cracking

9

REFERENCES:

1. Technical Bases for Inspection Requirements for PWR Pressurizer Head, Shell-to-Head and Nozzle-to-Vessel Welds. EPRI, Palo Alto, CA: 2019. 3002015905, ADAMS Accession No. ML21021A271.

2. American Society of Mechanical Engineers, Risk-Based Inspection: Development of Guidelines, Volume 2-Part 1 and Volume 2-Part 2, Light Water Reactor (LWR) Nuclear Power Plant Components. CRTD-Vols. 20-2 and 20-4, ASME Research Task Force on Risk-Based Inspection Guidelines, Washington, D.C., 1992 and 1998.

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

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

5. BWRVIP-108: BWR Vessels and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA 2002. 1003557.

6. U.S. NRC, Safety Evaluation of Proprietary EPRI Report, BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108),

December 19, 2007, ADAMS Accession No. ML073600374.

7. BWRVIP-241: BWR Vessels and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA 2010. 1021005.

8. U.S. NRC, Safety Evaluation of Proprietary EPRI Report, BWR Vessel and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii (BWRVIP-241), April 19, 2013, ADAMS Accession Nos. ML13071A240 and ML13071A233.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 23 of 53)

9. Code Case N-702, Alternate Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds, ASME Code Section XI, Division 1, Approval Date: February 20, 2004.

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

11. Technical Bases for Inspection Requirements for PWR Steam Generator Class 1 Nozzle-to-Vessel Welds and Class 1 and Class 2 Vessel Head, Shell, Tubesheet-to-Head and Tubesheet-to-Shell Welds. EPRI, Palo Alto, CA: 2019. 3002015906.

12. U.S. NRC, Vogtle Electric Generating Plant, Units 1 and 2 - Relief Request for Proposed Inservice Inspection Alternative VEGP-ISI-ALT-04-04 to the Requirements of the ASME Code (EPID L-2020-LLR-0109), January 11, 2021, ADAMS Accession No. ML20352A155.

13. American Society of Mechanical Engineers,Section XI Rules for Inservice Inspection of Nuclear Power Plant Components, 2007 Edition with 2008 Addenda, 2013 Edition, and 2019 Edition.

14. U.S. NRC, Vogtle Electric Generating Plant, Units 1 and 2 - Audit Report for the Promise Version 1.0 Probabilistic Fracture Mechanics Software Used in Relief Request VEGP-ISI-ALT-04-04 (EPID L-2019-LLR-0109), December 10, 2020, ADAMS Accession No. ML20258A002.

15. U.S. NRC Regulatory Federal Register Volume 86, Issue 57, March 26, 2021.

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.

17. Letter from D. Gudger (Constellation Energy Generation, LLC) to U.S. Nuclear Regulatory Commission, Supplemental Information - Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds, dated April 8, 2022, ADAMS Accession No. ML22098A179.

18. Letter from N. Salgado (U.S. Nuclear Regulatory Commission) to D. Rhoades (Constellation Energy Generation), Braidwood Station, Units 1 and 2 and Byron Station, Units 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 10, 2022, ADAMS Accession No. ML22307A246.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 24 of 53)

APPENDIX A

Plant-Specific Applicability 10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 25 of 53)

Plant-Specific Applicability for Braidwood Station

Section 9 of Reference [A1] provides applicability 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 Braidwood is provided in Table A-1.

Table A-1 indicates that all plant-specific requirements are met for Braidwood. Therefore, the results and conclusions of the EPRI report are applicable to Braidwood.

Table A-1 Plant-Specific Applicability of Reference [A1] Representative Analyses to Braidwood Station, Units 1 and 2 Pressurizer Shell-to-Head Welds (Circumferential and Longitudinal) and Nozzle-to-Shell Welds (Item Numbers B2.11, B2.12, and B3.110)

Category Requirement from Reference [A1] Applicability to Braidwood Station, Units 1 and 2

General The plant-specific In Appendix C of this proposed alternative, the number Requirements pressurizer general and type of the Braidwood, Units 1 and 2 general transients and cycles transients are compared to the transients listed in must be bounded by Table 5-6 of Reference [A1]. As shown in Table C-2, the those shown in Table 5-Braidwood, Units 1 and 2 transients are bounded by the 6 for a 60- year transients listed in Table 5-6 of Reference [A1].

operating life. It should be noted that the number of cycles were extrapolated to 80 years in the evaluations.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 26 of 53)

Category Requirement from Reference [A1] Applicability to Braidwood Station, Units 1 and 2

The materials of the The Braidwood, Units 1 and 2, pressurizer upper and pressurizer shell and lower heads are fabricated of SA-533 Grade A, Class 2, nozzles must be low to meet ASME Nuclear Vessels Code Section III.

alloy ferritic steels which SA-533 Grade A, Class 2 material is not specifically conform to the listed in ASME Code,Section XI, Appendix G. However, requirements of ASME SA-533 Grade A, Class 2 material has similar toughness Code,Section XI, and chemical composition to SA-533, Grade B, Class 1, Appendix G, Paragraph SA-508-1, SA-508-2, and SA-508-3. SA-533 Grade A, G-2110. Class 2 has a specified minimum yield strength at room temperature of 70 ksi (which is greater than 50 ksi), and the maximum RTNDT values for the Braidwood pressurizer bottom head materials are 60°F or less (so the RTNDT of 60°F used in the EPRI report is bounding).

Appendix G, Table G-2110-1, of the 2019 Edition of ASME Section XI acknowledges that Figure G-2210-1 is applicable to material SA-533 Grade A, Class 2.

Therefore, it can be concluded that the EPRI report is applicable to material SA-533 Grade A, Class 2.

Also, material properties in SA-533 Grade A, Class 2 material are identical compared to the SA-533 Grade B Class 1 material used in the FEA in the EPRI report as is shown Table 5-2 of Reference [A1]. Therefore, by comparison, SA-533 Grade A, Class 2 material is consistent with the requirements of ASME Code,Section XI, Appendix G and satisfies the requirements for application of the EPRI report.

The Braidwood, Units 1 and 2, pressurizer shells are fabricated from SA-533 Grade A, Class 2, material. The Braidwood, Units 1 and 2, pressurizer Surge, Spray, and Safety-Relief valve nozzles are all fabricated from SA-508, Class 2, material.

The materials for the pressurizer shells conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

Specific The plant-specific The Braidwood, Units 1 and 2, pressurizer shell-to-head Requirements pressurizer surge nozzle and nozzle-to-vessel weld configurations included in this and bottom head weld request are shown in Figures A-1, A-2, A-3, and A-4 and configurations must show conformance with the figures shown in Reference conform to those shown [A1].

in Figure 1-1 (Item No.

B2.11), Figure 1-2 (Item No. B2.12) and Figures 1-4 and 1-5 (Item No.

B3.110) of Reference

[A1].

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 27 of 53)

Category Requirement from Reference [A1] Applicability to Braidwood Station, Units 1 and 2

The plant-specific The comparison of the Braidwood, Units 1 and 2, dimensions of the pressurizer shell dimensions with those in Table 9-1 of pressurizer shell and the Reference [A1] is provided in Table A-2. The surge nozzle must be comparison shows that the Braidwood, Units 1 and 2 within the range of configurations are within the range of values shown in values listed in Table 9-1 Table 9-1 of Reference [A1].

of Reference [A1].

The plant-specific In Appendix D of this proposed alternative, the Insurge/Outsurge Braidwood, Units 1 and 2, Insurge/Outsurge transients transient definitions are compared to the number and type of transients (temperature difference listed in Table 5-10 of Reference [A1]. As can be seen between the pressurizer from Table D-2, the Braidwood, Units 1 and 2, transients shell and the pressurizer are bounded by those transients listed in Table 5-10 of surge nozzle fluid Reference [A1].

temperature and associated number of cycles) must be bounded by those shown in Table 5-10 for a Westinghouse/CE plant of Reference [A1].

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 28 of 53)

Table A-2 Range of Geometric Parameters for which the Evaluation is Applicable in Comparison with Braidwood Station, Units 1 and 2

Component Geometric Parameter For a Westinghouse Plant Braidwood Station, Units 1 and 2 Dimensions Pressurizer Shell Inside Diameter (in) Must be between 80 and 88 84 [A3]

NPS of piping or Surge Nozzle component (e.g., reducer) Must be between 12 and 18 14 [A2]

attached to nozzle safe-end (in)

NPS of piping or Safety/Relief component (e.g., reducer) Must be between 4 and 8 6 [A2]

Nozzle attached to nozzle safe-end (in)

NPS of piping or Spray Nozzle component (e.g., reducer) Must be between 4 and 6 4 [A2]

attached to nozzle safe-end (in)

REFERENCES

A1. 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.

A2. Westinghouse Electric Corporation Drawing No. 1100J48, Outline: Pressurizer 1800 Cu.

Ft. [50.96].

A3. Westinghouse Electric Corporation Drawing No. 1101J22, General Arrangement:

Pressurizer 1800 Cu. Ft. [50.96].

A4. Braidwood Station, Drawings 1PZR-01 and 2PZR-01, Inspection Identification Drawing Inservice Inspection for Pressurizer NC. 1RY01S and 2RY01S.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 29 of 53)

Figure A-1: Braidwood Station, Unit 1 Pressurizer Vessel Weld Locations [A4]

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 30 of 53)

Figure A-2: Braidwood Station, Unit 2 Pressurizer Vessel Weld Locations [A4]

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 31 of 53)

Figure A-3: Braidwood Station, Units 1 and 2 Typical Pressurizer Nozzle-to-Shell Weld Details 10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 32 of 53)

Figure A-4: Braidwood Station, Units 1 and 2 Typical Shell and Head Weld Details [A3]

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2 Revision 1 (Page 33 of 53)

Plant-Specific Applicability for Byron Station

Section 9 of Reference [A5] provides applicability 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 Byron is provided in Table A-3.

Table A-3 indicates that all plant-specific requirements are met for Byron. Therefore, the results and conclusions of the EPRI report are applicable Byron.

Table A-3 Plant-Specific Applicability of Reference [A5] Representative Analyses to Byron Station, Units 1 and 2 Pressurizer Shell-to-Head Welds (Circumferential and Longitudinal) and Nozzle-to-Shell Welds (Item Numbers B2.11, B2.12, and B3.110)

Category Requirement from Reference [A5] Applicability to Byron Station, Units 1 and 2 General The plant-specific In Appendix C of this proposed alternative, the number Requirements pressurizer general and type of the Byron, Units 1 and 2 general transients transients and cycles are compared to the transients listed in Table 5-6 of must be bounded by Reference [A5]. As shown in Table C-4, the Byron, Units those shown in Table 5-1 and 2 transients are bounded by the transients listed 6 for a 60- year in Table 5-6 of Reference [A5].

operating life. It should be noted that the number of cycles were extrapolated to 80 years in the evaluations.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2 Revision 1 (Page 34 of 53)

Category Requirement from Reference [A5] Applicability to Byron Station, Units 1 and 2 The materials of the The Byron, Units 1 and 2, pressurizer upper and lower pressurizer shell and heads are fabricated of carbon steel casting SA-533 nozzles must be low Grade A, Class 2, to meet ASME Nuclear Vessels Code alloy ferritic steels which Section III.

conform to the SA-533 Grade A, Class 2 material is not specifically requirements of ASME listed in ASME Code,Section XI, Appendix G. However, Code,Section XI, SA-533 Grade A, Class 2 material has similar toughness Appendix G, Paragraph and chemical composition to SA-533, Grade B, Class 1, G-2110. SA-508-1, SA-508-2, and SA-508-3. SA-533 Grade A, Class 2 has a specified minimum yield strength at room temperature of 70 ksi (which is greater than 50 ksi), and the maximum RTNDT values for the Byron pressurizer bottom head materials are 60°F or less (so the RTNDT of 60°F used in the EPRI report is bounding). Appendix G, Table G-2110-1, of the 2019 Edition of ASME Section XI acknowledges that Figure G-2210-1 is applicable to material SA-533 Grade A, Class 2. Therefore, it can be concluded that the EPRI report is applicable to material SA-533 Grade A, Class 2.

Also, material properties in SA-533 Grade A, Class 2 material is identical compared to the SA-533 Grade B Class 1 material used in the FEA in the EPRI report as is shown Table 5-2 of Reference [A5]. Therefore, by comparison, SA-533 Grade A, Class 2 material is consistent with the requirements of ASME Code,Section XI, Appendix G and satisfies the requirements for application of the EPRI report.

The Byron, Units 1 and 2, pressurizer shells are fabricated from SA-533 Grade A, Class 2, material. The Byron, Units 1 and 2 pressurizer Surge, Spray, and Safety-Relief valve nozzles are all fabricated from SA-508, Class 2, material.

The materials for the pressurizer shells conform to the requirements of ASME Code,Section XI, Appendix G, Paragraph G-2110.

Specific The plant-specific The Byron, Units 1 and 2, pressurizer shell-to-head and Requirements pressurizer surge nozzle nozzle-to-vessel weld configurations included in this and bottom head weld request are shown in Figures A-5, A-6, A-7, and A-8 and configurations must show conformance with the figures shown in Reference conform to those shown [A5].

in Figure 1-1 (Item No.

B2.11), Figure 1-2 (Item No. B2.12) and Figures 1-4 and 1-5 (Item No.

B3.110) of Reference

[A5].

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2 Revision 1 (Page 35 of 53)

Category Requirement from Reference [A5] Applicability to Byron Station, Units 1 and 2 The plant-specific The comparison of the Byron, Units 1 and 2, pressurizer dimensions of the shell dimensions with those in Table 9-1 of Reference pressurizer shell and the [A5] is provided in Table A-4. The comparison shows surge nozzle must be that the Byron, Units 1 and 2, configurations are within within the range of the range of values shown in Table 9-1 of Reference values listed in Table 9-1 [A5].

of Reference [A5].

The plant-specific In Appendix D of this proposed alternative, the Byron, Insurge/Outsurge Units 1 and 2, Insurge/Outsurge transients are transient definitions compared to the number and type of transients listed in (temperature difference Table 5-10 of Reference [A5]. As can be seen from between the pressurizer Table D-4, the Byron, Units 1 and 2, transients are shell and the pressurizer bounded by those transients listed in Table 5-10 of surge nozzle fluid Reference [A5].

temperature and associated number of cycles) must be bounded by those shown in Table 5-10 for a Westinghouse/CE plant of Reference [A5].

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2 Revision 1 (Page 36 of 53)

Table A-4 Range of Geometric Parameters for which the Evaluation is Applicable in Comparison with Byron Station, Units 1 and 2

Component Geometric Parameter For a Westinghouse Plant Byron Station, Units 1 and 2 Dimensions

Pressurizer Shell Inside Diameter (in) Must be between 80 and 88 84 [A7]

NPS of piping or Surge Nozzle component (e.g., reducer) Must be between 12 and 18 14 [A6]

attached to nozzle safe-end (in)

NPS of piping or Safety/Relief component (e.g., reducer) Must be between 4 and 8 6 [A6]

Nozzle attached to nozzle safe-end (in)

NPS of piping or Spray Nozzle component (e.g., reducer) Must be between 4 and 6 4 [A6]

attached to nozzle safe-end (in)

REFERENCES

A5. 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.

A6. Westinghouse Electric Corporation Drawing No. 1100J48, Outline: Pressurizer 1800 Cu.

Ft. [50.96].

A7. Westinghouse Electric Corporation Drawing No. 1101J22, General Arrangement:

Pressurizer 1800 Cu. Ft. [50.96].

A8. Byron Station, Drawings 1PZR-1-ISI and 2PZR-1-ISI, Inspection Identification Drawing Inservice Inspection for Pressurizer No. 1RY01S and 2RY01S.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2 Revision 1 (Page 37 of 53)

Figure A-5: Byron Station, Unit 1 Pressurizer Vessel Weld Locations [A8]

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2 Revision 1 (Page 38 of 53)

Figure A-6: Byron Station, Unit 2 Pressurizer Vessel Weld Locations [A8]

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 39 of 53)

Figure A-7: Byron Station, Units 1 and 2 Typical Pressurizer Nozzle-to-Shell Weld Details 10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 40 of 53)

Figure A-8: Byron Station, Units 1 and 2 Typical Shell and Head Weld Details [A5]

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 41 of 53)

APPENDIX B

Results of Industry Survey 10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 42 of 53)

Overall Industry Inspection Summary

The results of an industry survey of past inspections of pressurizer welds are summarized in Reference

[B1]. Table B-1 provides a summary of the combined survey results for Item Numbers B2.11, B2.12, 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 47 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 1,128 examinations of components for the three affected Item Numbers included in this proposed alternative 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 1,128 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 on examinations showed no change in flaw sizes since the original inspections. No other indications were identified in any in-scope components.

Table B-1 Summary of Survey Results

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

B2.12 269 0 B3.110 590 0

Note:

1. Flaw evaluations were performed to show acceptability of these indications and follow on examinations showed no change in flaw sizes since the original inspections.

REFERENCE

B1. 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.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 43 of 53)

APPENDIX C

Comparison of Braidwood Station and Byron Station General Transients to the Transients Evaluated in EPRI Report 10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 44 of 53)

Comparison of Braidwood Station, Units 1 and 2 General Transients to the Transients Evaluated in Reference [C1]

Braidwood general transients are tracked by Constellation and the number of cycles encountered as of 2020 for the transients relevant to this request are provided in Table C-1

[C2]. As indicated in Reference [C1], not all the transients tracked by Constellation for Braidwood are required for this request. Table 5-5 of Reference [C1] identified the significant cycles to be used in the evaluation based on expected cycles from a fleet fatigue monitoring review. The Reference [C1] report considered the heatup/coodown and loss of load. Leakage tests are conducted as an integral part of the plant heatup process; therefore, no additional cycles were included solely for leakage testing. Braidwood would also expect to perform operating leakage testing, instead of hydrostatic testing, following any potential Braidwood Units 1 and 2 pressurizer repairs required by Paragraph IWA-4540(a) of ASME Section XI, in the future. The report considered 300 heatup and cooldown transients for 60 years of operation. The loss of load condition was the most limiting transient, but the cycles were increased to account for other similar events (reactor trip, loss of flow, and loss of power) and thus increased the number of cycles to 360.

For comparison with Table 5-6 of Reference [C1], the actual number of cycles in Table C-1 were projected to 60 years. The comparison of Braidwood general transients to the requirements in Reference [C1] is shown in Table C-2.

Table C-1 Braidwood Station, Units 1 and 2, General Transients Applicable to This Request [C2-C7]

Braidwood Station, Unit 1 Braidwood Station, Unit 2 Transient Name Maximum Maximum Up to 60-Year Cycles Up to 60-Year Cycles 2020 Projected (Controlling 2020 Projected (Controlling Limit) Limit)

1 RCS Heat Up 40 72 200 43 79 200 2 RCS Cool Down 39 70 200 51 94 200 3 Pressurizer Cooldown 40 72 200 41 75 200 4 Reactor Trips 6 11 230 38 69 230 Reactor Trips 5 w/Subsequent 11 20 160 3 5 160 Cooldown

6 Reactor Trip w/ Subsequent SI 3 5 10 0 0 10

50% Step Load See See Steps See Steps See See Steps See Steps 7a, 7 Decrease with Steam Steps 7a, 7a, 7b, 7c, 7d 7a, 7b, 7c, 7d Steps 7a, 7a, 7b, 7c, 7b, 7c, 7d Dump (See Below) 7b, 7c, 7d 7b, 7c, 7d 7d 7a Unload @5%/min 104 187 12240 86 158 13200 10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 45 of 53)

Braidwood Station, Unit 1 Braidwood Station, Unit 2 Transient Name Maximum Maximum Up to 60-Year Cycles Up to 60-Year Cycles 2020 Projected (Controlling 2020 Projected (Controlling Limit) Limit)

7b Step Load 10% Decrease 3 5 2000 5 9 2000

7c Unloading, 15%-0% Power 17 31 500 15 28 500

7d Large Step Load Decrease 2 4 200 1 2 200

8 Loss of Load 1 2 80 1 2 80

9 Loss of Offsite AC Power 1 2 40 2 4 40

See See Steps See Steps See See Steps See Steps 10a, 10 Loss of Flow in One RC Loop Only (BRW below) Steps 10a, 10b 10a, 10b Steps 10a, 10b 10b 10a, 10b 10a, 10b 10a Partial Loss of Flow 1 2 80 1 2 80 10b Complete Loss of Flow 0 0 5 1 2 5

Note:

1. Allowable limits and totals are from the Braidwood, Unit 1 and 2, Fatigue Monitoring Reports (December 2020) Work Orders 05107062 and 05107063 (Procedure BwVP 850-7).

2. 60-year projection is based on Braidwood, Unit 1 and 2, Semi-Annual Fatigue Monitoring Report (January 2021) under Work Orders 05064062 and 05064063.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 46 of 53)

Table C-2 Comparison of Braidwood Station, Units 1 and 2, General Transients to the Transients Evaluated in Reference [C1]

Number of Braidwood Station, Braidwood Station, Cycles for 60 Unit 1 60-Year Unit 2 60-Year Transient Years from Projections from Projections from Reference [C1] Table C-1 Table C-1

Heatup / Cooldown 300 72 94 Loss of Load (Sum of Reactor Trips, 50%

Step Load Decrease with 360 269 281 Steam Dump, Loss of Load, Loss of Flow in One RC Loop Only, and Loss of Offsite AC Power Events)

REFERENCES

C1. 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.

C2. Braidwood Station, Unit 1, December Fatigue Monitoring Report (Monthly), Work Order 05107062.

C3. Braidwood Station, Unit 2, December Fatigue Monitoring Report (Monthly), Work Order 05107063.

C4. Braidwood Station, Unit 1, Semi-Annual Report (January), Work Order 05064063.

C5. Braidwood Station, Unit 2, Semi-Annual Report (January), Work Order 05064062.

C6. Procedure BwVP 850-7, Operational Transient Cycle Counting.

C7. WCAP-15966, Evaluation of Pressurizer Insurge / Outsurge Transients for Byron and Braidwood, 2002.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 47 of 53)

Comparison of Byron Station, Units 1 and 2, General Transients to the Transients Evaluated in Reference [C8]

Byron general transients are tracked by Constellation and the number of cycles encountered as of 2020 for the transients relevant to this request are provided in Table C-3 [C9]. As indicated in Reference [C8], not all the transients tracked by Constellation for Byron are required for this request. Table 5-5 of Reference [C8] identified the significant cycles to be used in the evaluation based on expected cycles from a fleet fatigue monitoring review. The Reference [C8]

report considered the heatup/coodown and loss of load. Leakage tests are conducted as an integral part of the plant heatup process; therefore, no additional cycles were included solely for leakage testing. Byron would also expect to perform operating leakage testing, instead of hydrostatic testing, following any potential Byron Units 1 and 2 pressurizer repairs, required by Paragraph IWA-4540(a) of ASME Section XI, in the future. The report considered 300 heatup and cooldown transients for 60 years of operation. The loss of load condition was the most limiting transient, but the cycles were increased to account for other similar events (reactor trip, loss of flow, and loss of power) and thus increased the number of cycles to 360.

For comparison with Table 5-6 of Reference [C8], the actual number of cycles in Table C-3 were projected to 60 years. The comparison of Byron general transients to the requirements in Reference [C8] is shown in Table C-4.

Table C-3 Byron Station, Units 1 and 2, General Transients Applicable to This Request [C9]

Byron Station, Unit 1 Byron Station, Unit 2 Maximum Maximum Transient Name Up to 60-Year Cycles Up to 60-Year Cycles 2020 Projected (Controlling 2020 Projected (Controlling Limit) Limit)

Heat Up@ <100°F/hr 38 66 200 34 62 200 Cooldown@ <100°F/hr 38 66 200 34 62 200 Reactor Trips 5 9 230 7 13 230 50% Step Load Decrease with Steam Dump 2 4 200 3 6 200 Loss of Load 3 6 80 1 2 80 Loss of Flow in One RC Loop Only 0 0 80 0 0 80 Loss of Offsite AC Power 1 2 40 3 6 40

Note:

1. Allowable limits and totals are from the Byron, Unit 1 and 2, Annual Fatigue Monitoring Report (2020),

EC 633359, Revision 0.

2. 60-year projection is based on Byron, Unit 1 and 2, Annual Fatigue Monitoring Report (2020), EC 633359, Revision 0.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 48 of 53)

Table C-4 Comparison of Byron Station, Units 1 and 2, General Transients to the Transients Evaluated in Reference [C8]

Number of Byron Station, Unit Byron Station, Unit Cycles for 60 1 60-Year 2 60-Year Transient Years from Projections from Projections from Reference [C8] Table C-3 Table C-3

Heatup / Cooldown 300 66 62 Loss of Load (Sum of Reactor Trips, 50%

Step Load Decrease with 360 21 27 Steam Dump, Loss of Load, Loss of Flow in One RC Loop Only, and Loss of Offsite AC Power Events)

REFERENCES

C8. 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.

C9. EC 633359, Revision 0, Annual Fatigue Monitoring Report 2020 Unit 1 & Unit 2.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 49 of 53)

APPENDIX D

Comparison of Braidwood Station and Byron Station Insurge/Outsurge Transients to the Insurge/Outsurge Transients Evaluated EPRI Report 10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 50 of 53)

Comparison of Braidwood Station, Units 1 and 2, Insurge/Outsurge Transients to the Insurge/Outsurge Transients Evaluated in Reference [D1]

Braidwood Insurge/Outsurge transients are provided in Table D-1 [D2-D7]. The temperature differences

( Ts) identified in Table D-1 are combined conservatively by summing all the events into the 320°F T bin. It should be noted that the transients in Table D-1 reflect 40 years of operation, so for comparison with Table 5-10 of Reference [D1], they are extrapolated to 60 years by multiplying by 1.5. With this conservative treatment of the Insurge/Outsurge transients, the comparison of Braidwood Insurge/Outsurge transients to the requirements in Reference [D1] is shown in Table D-2. The results of Table D-2 indicate that the Braidwood Insurge/Outsurge transients are bounded by those in Reference

[D1].

Table D-1 40-Year Insurge/Outsurge Transients for Braidwood Station, Units 1 and 2 [D2-D7]

No. Transient Name (1,2,3) Unit 1 Unit 2 Up to 60-Year Allowable Up to 60-Year Allowable

2020 Projected Limit 2020 Projected Limit

[D2] [D4](5) [D7](4) [D3] [D5](5) [D7](4) 1 PZR I/O SURGE MOPHU320 4 10 56 4 10 56 2 PZR I/O SURGE MOPHU300 2 5 20 0 0 20 3 PZR I/O SURGE MOPHU280 3 8 21 0 0 21 4 PZR I/O SURGE MOPHU270 9 23 70 6 16 70 5 PZR I/O SURGE MOPCD320 0 0 41 0 0 41 6 PZR I/O SURGE MOPCD310 0 0 15 0 0 15 7 PZR I/O SURGE MOPCD300 1 3 20 2 5 20 8 PZR I/O SURGE MOPCD270 1 3 76 0 0 76 9 PZR I/O SURGE MOPCD250 4 10 15 8 21(6) 15

Notes:

1. The Transient Name is MOPXXnnn, where MOP = post Modified Operating Procedures, XX = HU for insurge/outsurge transients that occur during Heatup events, or CD for insurge/outsurge transients that occur during Cooldown events, and nnn = temperature difference, delta T, between the RCS piping at the beginning of the transient and pressurizer temperature at the end of the transient

2. Insurge/Outsurge determination is based on WCAP-15966

3. Braidwood does not explicitly monitor the breakdown of the transients shown in this table. Rather, the number of cycles shown in this table were used in the governing fatigue evaluations for the pressurizer surge nozzles.

Braidwood does NOT currently monitor the environmental fatigue usage factor for the surge nozzle but is expected to do so with implementation of WESTEMs as part of license renewal commitments for the station.

4. Per Note 2 (WCAP-15966) the allowable insurge/outsurge numbers were used to evaluate the stresses for Braidwood. As such, these numbers are the same for 40 years and 60 years.

5. 60-year projection is based on Braidwood, Unit 1 and 2, Semi-Annual Fatigue Monitoring Report (January 2021) under Work Orders 05064062 and 05064063.

6. In Reference [D5], PZR I/O SURGE MOPCD250 was identified as reaching 53.33% of the allowable limit for Unit 2, and was projected to exceed the allowable limit in 2033. This projection is skewed based on the occurrences in 2008 and 2009, when three of the events occurred during a refueling outage cool down. Except for one event in 2014 during the A2R17 refueling outage, there have been zero occurrences of the PZR I/O SURGE MOPCD250 events between 2010 and 2020. Based on the latest trend, the 60-Year Projected count should decrease as the plant reaches its PEO and the projected year to reach the limit will continue to increase beyond 2033.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 51 of 53)

Table D-2 Comparison of Braidwood Station, Units 1 and 2, Insurge/Outsurge Transient Temperature Differences and Numbers of Cycles with the Insurge/Outsurge Transient Date from Reference [D1]

60-Year No. of Braidwood Station, Unit Braidwood Station, Cycles 1 Cycles Projected to 60 Unit 2 Cycles T (oF)(1) From Reference Years of Operation Projected to 60 Years

[D1] of Operation 330 600 0 0 320 3,000 62 52 103 1,500 0 0 Notes:

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

REFERENCES

D1. 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.

D2. Braidwood Station, Unit 1, December Fatigue Monitoring Report (Monthly), Work Order 05107062.

D3. Braidwood Station, Unit 2, December Fatigue Monitoring Report (Monthly), Work Order 05107063.

D4. Braidwood Station, Unit 1, Semi-Annual Report (January), Work Order 05064063.

D5. Braidwood Station, Unit 2, Semi-Annual Report (January), Work Order 05064062.

D6. Procedure BwVP 850-7, Operational Transient Cycle Counting.

D7. WCAP-15966, Evaluation of Pressurizer Insurge / Outsurge Transients for Byron and Braidwood, 2002.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 52 of 53)

Comparison of Byron Station, Units 1 and 2, Insurge/Outsurge Transients to the Insurge/Outsurge Transients Evaluated in Reference [D8]

Byron Insurge/Outsurge transients are provided in Table D-3 [D9-D14]. The temperature differences

( Ts) identified in Table D-3 are combined conservatively by summing all the events into the 320°F T bin. It should be noted that the transients in Table D-3 reflect 40 years of operation, so for comparison with Table 5-10 of Reference [D8], they are extrapolated to 60 years by multiplying by 1.5. With this conservative treatment of the Insurge/Outsurge transients, the comparison of Byron Insurge/Outsurge transients to the requirements in Reference [D8] is shown in Table D-4. The results of Table D-4 indicate that the Byron Insurge/Outsurge transients are bounded by those in Reference [D8].

Table D-3 40-Year Insurge/Outsurge Transients for Byron Station, Units 1 and 2 [D9-D14]

No. Transient Name (1,2) WCAP-Unit 1 Unit 2 15966 Past Up to 60-Year Allowable Up to 60-Year Allowable Total 2020 Projected(3) Limit 2020 Projected(3) Limit Count [D11](2) [D10](4) [D11](2) [D10](4)

[D10]

1 PZR I/O Surge-MOPHU320 10 3 20 35 3 20 35 2 PZR I/O Surge-MOPHU310 5 0 5 17 0 5 17 3 PZR I/O Surge-MOPHU300 5 1 9 17 1 9 17 4 PZR I/O Surge-MOPHU280 15 0 15 52 1 19 52 5 PZR I/O Surge-MOPHU270 12 2 19 42 0 12 42 6 PZR I/O Surge-MOPHU250 11 8 36 37 10 44 37 7 PZR I/O Surge-MOPCD320 7 0 7 24 1 11 24 8 PZR I/O Surge-MOPCD310 9 0 9 31 0 9 31 9 PZR I/O Surge-MOPCD300 9 0 9 31 0 9 31 10 PZR I/O Surge-MOPCD290 11 0 11 38 0 11 38 11 PZR I/O Surge-MOPCD280 4 0 4 14 0 4 14 12 PZR I/O Surge-MOPCD270 11 0 11 38 1 15 38 13 PZR I/O Surge-MOPCD250 7 3 17 24 6 27 24

Notes:

1. The Transient Name is PZR I/O Surge-MOP XX nnn, where XX = HU for insurge/outsurge transients that occur during Heatup events, or CD for insurge/outsurge transients that occur during Cooldown events, and nnn = the temperature difference, T, between the pressurizer fluid temperature and the fluid temperature in the surge nozzle.

2. Byron has explicitly monitored the breakdown of the transients shown in this table since May 2008, when BVP 900-3, Revision 6 [D11] was issued. This monitoring was implemented to confirm the MOP strategies recommended in WCAP-14950 [D14] are effective and to ensure that the projections in WCAP-15966 [D10] remain bounding.

3. The projection for the number of Insurge/Outsurge Transients over 60 years of operation is equal to the sum of the conservatively estimated past events as documented in WCAP-15966 [D10] and the number of events recorded between 2008 and 2020 [D9] increased to assume the same rate of occurrence until the end of the 60 year operating period.

4. The Allowable Limit is the sum of the conservatively estimated past events as documented in WCAP-15966 [D10] and the projected number of future MOP Heatup/Cooldown events assumed in WCAP-15966 [D10].

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, and Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, Revision 1 (Page 53 of 53)

Table D-4 Comparison of Byron Station, Units 1 and 2, Insurge/Outsurge Transient Temperature Differences and Numbers of Cycles with the Insurge/Outsurge Transient Date from Reference

[D8]

Byron Unit 1 Byron Unit 2 T (oF)(1) 60-Year No. of Cycles from Cycles Projected Cycles Projected Reference [D8] to 60 Years of to 60 Years of Operation Operation 330 600 27(2) 31(2) 320 3,000 172(3) 195(3) 103 1,500 53(4) 71(4)

Notes:

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

2. Byron has an administrative limit of 320°F on system T. WCAP-15966 [D10] did identify a small number of events that exceeded 320°F on system T and assumed those events to be 320°F for the purposes of developing the system T distribution in the analysis. Based on this precedence, the number of cycles at T = 330°F is conservatively considered to be equal to sum of all heatup and cooldown events in Table D-3 for T = 320°F, since Byron does not specifically monitor for transients with a temperature difference of 330°F.

3. The number of cycles at T = 320°F is conservatively considered to be equal to sum of all events in Table D-3.

4. The number of cycles at T = 103°F is conservatively considered to be equal to sum of the heatup and cooldown events in Table D-3 for T = 250°F, since Byron does not specifically monitor for transients with a temperature difference of 103°F. Byron does count any event with T >80°F and 250°F in as applicable 250°F surge condition.

Therefore, the number of events in Table D-3 for T = 250°F is conservatively bounding.

REFERENCES

D8. 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.

D9. EC 633359, Revision 0, Annual Fatigue Monitoring Report 2020 Unit 1 & Unit 2.

D10. WCAP-15966, Evaluation of Pressurizer Insurge/Outsurge Transients for Byron and Braidwood, Rev. 0, dated December 2002.

D11. BVP 900-3, Documentation of Operating Plant/Component Cyclic or Transient Events, Rev.

6, issued May 9, 2008.

D12. BVP 900-3, Documentation of Operating Plant/Component Cyclic or Transient Events, Rev.

8 (current revision), issued February 27, 2012.

D13. Exelon Procedure ER-AA-470, Fatigue and Transient Monitoring Program, Rev. 8, issued January 30, 2019.

D14. WCAP-14950, Mitigation and Evaluation of Pressurizer Insurge/Outsurge Transients,

February 1998.