RS-21-056, Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Weld

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Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Weld
ML21133A297
Person / Time
Site: Calvert Cliffs, Byron, Braidwood  Constellation icon.png
Issue date: 05/12/2021
From: David Gudger
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-21-056
Download: ML21133A297 (71)
v • d • e


Text

200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com 10 CFR 50.55a RS-21-056 May 12, 2021 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455 Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Renewed Facility Operating License Nos. DPR-53 and DPR-69 NRC Docket Nos. 50-317 and 50-318

Subject:

Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds In accordance with 10 CFR 50.55a(z)(1), Exelon Generation Company, LLC (Exelon) hereby requests NRC approval of a proposed alternative to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," on the basis that the proposed alternative provides an acceptable level of quality and safety. Specifically, Exelon is requesting an alternative to volumetric examination of pressurizer circumferential and longitudinal shell-to-head welds and nozzle-to-shell welds to extend the inspection frequency from 10 years to the remainder of the currently licensed operating periods for Braidwood Generating Station (Braidwood), Units 1 and 2, Byron Generating Station (Byron), Units 1 and 2, and Calvert Cliffs Nuclear Power Plant (Calvert Cliffs), Units 1 and 2.

Proposed Alternative I4R-15 for Braidwood, I4R-21 for Byron, and ISI-05-016 for Calvert Cliffs is provided in Attachment 1. Exelon requests approval of the proposed alternative by March 1, 2022 to support the Spring 2022 outage season when some of the subject examinations are currently scheduled.

There are no regulatory commitments contained in this letter.

Proposed Alternative for Examination of Pressurizer Circumferential and Longitudinal Shell-to-Head Welds and Nozzle-to-Vessel Welds May 12, 2021 Page 2 Should you have any questions concerning this matter, please contact Tom Loomis (610) 765-5510.

Respectfully, David T. Gudger Senior Manager - Licensing Exelon Generation Company, LLC : 10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Revision 0 cc: Regional Administrator - NRC Region I Regional Administrator - NRC Region III NRC Senior Resident Inspector - Braidwood Station NRC Senior Resident Inspector - Byron Station NRC Senior Resident Inspector - Calvert Cliffs Nuclear Power Plant NRC Project Manager - Braidwood Station NRC Project Manager - Byron Station NRC Project Manager - Calvert Cliffs Nuclear Power Plant Illinois Emergency Management Agency - Division of Nuclear Safety S. Seaman, State of Maryland

Attachment 1 10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Revision 0

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-16 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Revision 0 (Page 1 of 68)

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:

ASME ASME Unit Component ID Component Description Category Item 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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-16 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Revision 0 (Page 2 of 68)

Byron Component IDs:

ASME ASME Unit Component ID Component Description Category Item 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

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-16 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Revision 0 (Page 3 of 68)

Calvert Cliffs Component IDs:

ASME ASME Unit Component ID Component Description Category Item 1 B-B B2.11 3-401 Shell - Lower Head 1 B-B B2.11 8-411 Shell - Upper Head 2 B-B B2.11 3-401 Shell - Lower Head 2 B-B B2.11 8-411 Shell - Upper Head 1 B-B B2.12 2-401A Upper Shell At 270 Deg.

1 B-B B2.12 2-401B Upper Shell At 90 Deg.

1 B-B B2.12 2-401C Lower Shell At 180 Deg.

1 B-B B2.12 2-401D Lower Shell At 0 Deg.

2 B-B B2.12 2-401A Upper Shell At 270 Deg.

2 B-B B2.12 2-401B Upper Shell At 90 Deg.

2 B-B B2.12 2-401C Lower Shell At 180 Deg.

2 B-B B2.12 2-401D Lower Shell At 0 Deg.

1 B-D B3.110 4-404 Surge Nozzle 1 B-D B3.110 4-405 Spray Nozzle 1 B-D B3.110 16-405A Safety & Relief Nozzle 1 B-D B3.110 16-405B Safety & Relief Nozzle 2 B-D B3.110 4-404 Surge Nozzle 2 B-D B3.110 4-405 Spray Nozzle 2 B-D B3.110 16-405A Safety & Relief Nozzle 2 B-D B3.110 16-405B Safety & Relief 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, Byron, and Calvert Cliffs:

PLANT INTERVAL EDITION START END Braidwood Station, August 29, 2018 (Unit 1) July 28, 2028 (Unit 1)

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

Byron Station, 2007 Edition, through Fourth July 16, 2016 July 15, 2025 Units 1 and 2 2008 Addenda Calvert Cliffs Nuclear Power Fifth 2013 Edition July 1, 2019 June 30, 2029 Plant, Units 1 and 2 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 published in the proposed rules of the Federal Register, Vol. 86, No. 57. Currently there are conditions in the proposed NRC Rulemaking (86 FR 16087) regarding Table G-2110-1, but these conditions do not pertain to the use of Table G-2110-1 for material SA-533 Grade A, Class 2 as used in this proposed alternative).

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-16 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Revision 0 (Page 4 of 68) 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 interval 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 interval is requested.

5 Proposed Alternative and Basis for Use Exelon 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 from the current ASME Section XI 10-year requirement to the end of the currently approved Period of Extended Operation (PEO) for Braidwood Station (Braidwood), Units 1 and 2, Byron Station (Byron), Units 1 and 2, and Calvert Cliffs Nuclear Power Plant (Calvert Cliffs), Units 1 and 2, as summarized in the following table.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-16 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Revision 0 (Page 5 of 68)

Table 1. Summary of Inspection Deferrals in this Proposed Alternative End of Length of Current Time Until ASME Item Date of Last Licensed Next Station Unit Description Category No. Inspection Operating Inspection for Period This Request (60 Years) (Years)

Braidwood 1 B-B B2.11 Pressurizer, Shell-to-Head 10/12/2019 10/17/2046 27.0 Welds, Circumferential B-B B2.12 Pressurizer, shell-to-Head 10/12/2019 27.0 Welds, Longitudinal B-D B3.110 Pressurizer, Nozzle-to- 4/8/2021 25.0 Vessel Welds Braidwood 2 B-B B2.11 Pressurizer, Shell-to-Head 10/11/2018 12/18/2047 29.2 Welds, Circumferential B-B B2.12 Pressurizer, shell-to-Head 10/11/2018 29.2 Welds, Longitudinal B-D B3.110 Pressurizer, Nozzle-to- 10/11/2018 29.2 Vessel 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- 9/14/2018 26.1 Vessel Welds Byron 2 B-B B2.11 Pressurizer, Shell-to-Head 10/6/2014 11/6/2046 32.1 Welds, Circumferential B-B B2.12 Pressurizer, shell-to-Head 10/6/2014 32.1 Welds, Longitudinal B-D B3.110 Pressurizer, Nozzle-to- 4/13/2019 27.6 Vessel Welds Calvert 1 B-B B2.11 Pressurizer, Shell-to-Head 2/25/2020 7/31/2034 14.4 Cliffs Welds, Circumferential B-B B2.12 Pressurizer, shell-to-Head 2/25/2020 14.4 Welds, Longitudinal B-D B3.110 Pressurizer, Nozzle-to- 2/21/2016 18.5 Vessel Welds Calvert 2 B-B B2.11 Pressurizer, Shell-to-Head 3/13/2021 8/13/2036 15.4 Cliffs Welds, Circumferential B-B B2.12 Pressurizer, shell-to-Head 3/12/2021 15.4 Welds, Longitudinal B-D B3.110 Pressurizer, Nozzle-to- 3/10/2021 15.4 Vessel Welds As indicated in Table 1, the proposed alternative results in a maximum effective operating period of 32.1 years from the last inspection for Item Numbers B2.11 and B2.12 for Byron Unit 2 included in this proposed alternative. As summarized in Reference [1], the EPRI report demonstrates that for time intervals longer than 30 years (up to 80 years) no other inspections

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-16 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Revision 0 (Page 6 of 68) are required to maintain plant safety and relevant acceptance criteria.

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

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, Byron, and Calvert Cliffs 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, Byron, and Calvert Cliffs.

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. For Calvert Cliffs, PSI was followed by four 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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-16 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Revision 0 (Page 7 of 68)

Inspection History Plant operating experience (including examinations performed to-date, examination findings, inspection coverage, and previously submitted Proposed Alternatives) is summarized in Tables 2 through 4. 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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 8 of 68)

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 Examination currently 10-1995/A1R05 4-2006/A1R12 9-2016/A1R19 1 BRW-1-B02.11.0005 Coverage was not documented; but 100%/NRI 100%/NRI 98.5%/NRI scheduled for A2R25 B-B/B2.11 obstructions noted/NRI (2025) 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 Examination 8-1985 10-1995/A1R05 4-2006/A1R12 9-2016/A1R19 1 BRW-1-B02.12.0006 100%/NRI 100%/NRI 100%/NRI 100%/NRI currently scheduled B-B/B2.12 for A2R25 (2025) 1PZR-01-N4A 8-1985 10-1995/A1R053 4-2006/A1R12 Examination 9-2016/A1R19 1 BRW-1-B03.110.0013 Coverage was not documented; Coverage was not 100% Shell Side only due 68.7%/NRI currently scheduled B-D/B3.110 limitations due to Geometry/NRI documented/NRI to config/NRI for A2R25 (2025) 8-1985 1PZR-01-N4B Examination Coverage was not documented; 10-1995/A1R05 10-2004/A1R11 4-2015/A1R18 1 BRW-1-B03.110.0014 limitations due to Geometry/Laminar 100%/NRI 92.66%/NRI 90.9%/NRI currently scheduled B-D/B3.110 for A2R24 (2024) flaw 1PZR-01-N1 8-1985 A1R06 Examination 10-2007/A1R132 9-2016/A1R192 1 BRW-1-B03.110.0015 Coverage was not documented; (VT-2 per Relief Request 59.2%/NRI 74.7%/NRI currently scheduled B-D/B3.110 limitations due to Geometry/NRI per NR-24) for A2R25 (2025) 9-1998/A1R073 1PZR-01-N2 8-1985 10-1989/A1R013 Coverage was not 10-2010/A1R15 4-2021/A1R22 1 BRW-1-B03.110.0016 Coverage was not documented; Coverage was not documented; geometry 56.56%/NRI 56.43%/NRI B-D/B3.110 limitations due to Geometry/NRI documented/NRI limitations/NRI 9-1998/A1R073 1PZR-01-N3 8-1985 10-1989/A1R01 3 Coverage was not 10-2010/A1R15 4-2021/A1R22 1 BRW-1-B03.110.0017 Coverage was not documented; Coverage was not documented; geometry 60.92%/NRI 60%/NRI B-D/B3.110 limitations due to Geometry/NRI documented/NRI limitations/NRI 10-1995/A1R053 1PZR-01-N4C 8-1985 Examination Coverage was not 10-2004/A1R11 4-2015/A1R18 1 BRW-1-B03.110.0018 Coverage was not documented; documented; geometry 92.66%/NRI 90.9%/NRI currently scheduled B-D/B3.110 limitations due to Geometry/NRI for A2R24 (2024) limitations/NRI 1-1987 2PZR-01-08A 4-1990/A2R013 5-1999/A2R07 Examination Coverage was not documented; 10-2009/A2R14 2 BRW-2-B02.11.0005 limitations due to obstructions/

Coverage was not 99.85%/Laminations 95.57%/NRI currently scheduled B-B/B2.11 documented/NRI noted for A2R23 (2023)

Laminar and Planar flaws noted

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 9 of 68)

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) 12-1986 2PZR-01-08E Examination Coverage was not documented; 4-1996/A2R05 10-2006/A2R12 10-2018/A2R20 2 BRW-2-B02.11.0006 currently scheduled obstructions noted/Laminar and 100%/NRI 100%/NRI 100%/NRI B-B/B2.11 for A2R26 (2027)

Planar flaws noted 2PZR-01-09A Examination 1-1987 4-1990/A2R01 5-1999/A2R07 10-2009/A2R14 2 BRW-2-B02.12.0005 100%/NRI 100%/NRI 100%/NRI 100%/NRI currently scheduled B-B/B2.12 for A2R22 (2021) 2PZR-01-09D Examination 12-1986 4-1996/A2R05 10-2006/A2R12 10-2018/A2R20 2 BRW-2-B02.12.0006 currently scheduled 100%/Laminar type flaws noted 100%/NRI 100%/NRI 100%/NRI B-B/B2.12 for A2R26 (2027) 2PZR-01-N1 A2R06 Examination 1-1987 5-2008/A2R13 4-2017/A2R192 2 BRW-2-B03.110.0013 (VT-2 per Relief Request currently scheduled 100%/NRI 59.2%/NRI 59.2%/NRI B-D/B3.110 NR-24) for A2R25 (2026) 2PZR-01-N2 Examination 12-1986 4-1996/A2R05 4-1999/A2R07 4-2011/A2R15 2 BRW-2-B03.110.0014 currently scheduled 100%/NRI 100%/NRI 88.5%/NRI 88.5%/NRI B-D/B3.110 for A2R23 (2023) 4-1990/A2R013 2PZR-01-N3 12-1986 Examination Coverage was not 4-1999/A2R07 4-2011/A2R15 2 BRW-2-B03.110.0015 Coverage was not documented; currently scheduled documented; geometry 88.5%/NRI 88.5%/NRI B-D/B3.110 geometry limitations/NRI for A2R22 (2021) limitations /NRI 4-1990/A2R013 2PZR-01-N4A 12-1986 Examination Coverage was not 10-2006/A2R12 10-2018/A2R20 2 BRW-2-B03.110.0016 Coverage was not documented; currently scheduled documented; geometry 91.5%/NRI 91.5%/NRI B-D/B3.110 nozzle geometry limitations/NRI for A2R26 (2027) limitations /NRI 12-1986 3-1996/A2R053 2PZR-01-N4B Examination Coverage was not documented; Coverage was not 11-2003/A2R10 5-2014/A2R17 2 BRW-2-B03.110.0017 currently scheduled nozzle geometry limitations /Laminar documented; geometry 88.5%/NRI 88.5%/NRI B-D/B3.110 for A2R24 (2024) type flaw limitations /NRI 3-1996/A2R053 11-2003/A2R103 2PZR-01-N4C 12-1986 Examination Coverage was not Coverage was not 5-2014/A2R17 2 BRW-2-B03.110.0018 Coverage was not documented; currently scheduled documented; geometry documented; geometry 88.5%/NRI B-D/B3.110 nozzle geometry limitations /NRI for A2R24 (2024) 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

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 10 of 68) 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. Exelon 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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 11 of 68)

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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 12 of 68)

Table 3. Byron Station, Units 1 and 2, Pressurizer Welds Inspection History Weld ID PSI1 Interval 1 Interval 2 Interval 3 Interval 4 Summary Number Unit Exam Cat, Item Ultrasonic Test (UT) (Date/Outage, (Date/Outage, (Date/Outage, (Date/Outage, (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 Examination with 45° and 4 with 60°) 3-2011/B1R17 1 BYR-1-B2.11.0001 2 rejectable 45° indications and currently scheduled 92%/NRI B-B/B2.11 3 rejectable 60° indications 1-1990/B1R03 for B1R25 (2023)

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 8-1982 1RY-01-S/PC-05 Examination Coverage was not documented; 4-1996/B1R07 3-2005/B1R13 9-2015/B1R20 1 BYR-1-B2.11.0002 currently scheduled scan limitations (welded pads) 96%/NRI 96%/NRI 95.47%/NRI B-B/B2.11 for B1R25 (2023) were identified/NRI 8-1982 1RY-01-S/PL-01 4-1987/B1R01 Examination Coverage was not documented; 3-2002/B1R11 3-2011/B1R17 1 BYR-1-B2.12.0001 90% - limited by sample currently scheduled no scan limitations were 100%/NRI 100%/NRI B-B/B2.12 nozzles and axis plates/NRI for B1R25 (2023) identified/NRI 8-1982 1RY-01-S/PL-04 Examination Coverage was not documented; 4-1996/B1R07 3-2005/B1R13 9-2015/B1R20 1 BYR-1-B2.12.0002 currently scheduled no scan limitations were 100%/NRI 100%/NRI 100%/NRI B-B/B2.12 for B1R25 (2023) identified/NRI 9-2018/B1R223 7-1982 9-2006/B1R143 1RY-01-S/PN-01 72.03% - limited by Coverage was not documented; B1R07 (VT-2 per Relief Request B1R11 (VT-2 per Relief 40% - limited by nozzle 1 BYR-1-B3.110.0001 NR-19) Request I2R-03) nozzle configuration no scan limitations were configuration and heater B-D/B3.110 and heater identified/NRI penetrations/NRI penetrations/NRI

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 13 of 68)

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

Number 7-1982 Coverage was not documented; 1-1990/B1R03 1RY-01-S/PN-02 11-1997/B1R08 9-2006/B1R14 Examination no scan limitations were 75% - limited by nozzle 1 BYR-1-B3.110.0002 73% - limited by nozzle 77% - limited by nozzle currently scheduled identified/ configuration/NRI with 0° and B-D/B3.110 configuration/NRI configuration/NRI for B1R26 (2024)

Indications - appears to be ID 45°. ID Geometry with 60° geometry 7-1982 1RY-01-S/PN-03 11-1997/B1R08 9-2006/B1R14 Examination Coverage was not documented; 3-1987/B1R01 1 BYR-1-B3.110.0003 69% - limited by nozzle 77% - limited by nozzle currently scheduled no scan limitations were 100%/NRI B-D/B3.110 configuration/NRI configuration/NRI for B1R26 (2024) identified/NRI 7-1982 Coverage was not documented; 1-1990/B1R03 1RY-01-S/PN-04 3-2002/B1R11 9-2015/B1R20 Examination no scan limitations were 75% - limited by nozzle 1 BYR-1-B3.110.0004 66% - limited by nozzle 64.93% - limited by nozzle currently scheduled identified configuration/NRI with 0° and B-D/B3.110 configuration/NRI configuration/NRI for B1R25 (2023)

Indications - appears to be ID 45°. ID Geometry with 60° geometry 9-2006/B1R14 68% - limited by nozzle 7-1982 1RY-01-S/PN-05 11-1997/B1R08 configuration/NRI Examination Coverage was not documented; 4-1996/B1R07 1 BYR-1-B3.110.0005 73% - limited by nozzle currently scheduled no scan limitations were 100%/NRI B-D/B3.110 configuration/NRI 9-2015/B1R20 for B1R25 (2023) identified/NRI 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 Examination 1 BYR-1-B3.110.0006 identified/ limited due to 73% - limited by nozzle 77% - limited by nozzle currently scheduled B-D/B3.110 Indications - appears to be ID configuration/Spot indication configuration/NRI configuration/NRI for B1R26 (2024) geometry and one spot indication similar to PSI (seen with 45°)

1-1989/B2R014 10-1985 3-2004/B2R11 Coverage was not 2RY-01-S/PC-01 Coverage was not documented; 92% - scan limitations Examination documented; scan limitations 10-2014/B2R18 2 BYR-2-B2.11.0001 scan limitations (welded pads, (welded pads, currently scheduled (welded pads, sampling 94%/NRI B-B/B2.11 sampling nozzle and skirt) were sampling nozzle and for B2R23 (2022) nozzle and skirt) were identified/NRI skirt)/NRI identified/NRI

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 14 of 68)

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

Number 3-1995/B2R054 10-1985 2RY-01-S/PC-05 Coverage was not Examination Coverage was not documented; 9-2005/B2R12 4-2010/B2R15 2 BYR-2-B2.11.0002 documented; scan limitations currently scheduled scan limitations (welded pads) 96%/NRI 96%/NRI B-B/B2.11 (welded pads) were for B2R24 (2023) were identified/NRI identified/NRI 10-1985 Coverage was not documented; no scan limitations were 1-1989/B2R014 2RY-01-S/PL-01 Examination identified/ Coverage was not 3-2004/B2R11 10-2014/B2R18 2 BYR-2-B2.12.0001 currently scheduled Indications - 0° identified 4 documented; no scan 100%/NRI 100%/NRI B-B/B2.12 for B2R23 (2022) embedded indications, 45° was NRI, limitations were identified/NRI 60° identified 2 embedded indications.

10-1985 3-1995/B2R054 2RY-01-S/PL-04 Examination Coverage was not documented; no Coverage was not 9-2005/B2R12 4-2010/B2R15 2 BYR-2-B2.12.0002 currently scheduled scan limitations were documented; no scan 100%/NRI 100%/NRI B-B/B2.12 for B2R24 (2023) identified/NRI limitations were identified/NRI 10-1985 Coverage was not documented; no 4-2019/B2R213 4-2007/B2R133 2RY-01-S/PN-01 scan limitations were identified/ 11-1999/B2R08 40% - limited by 2-1989/B2R01 40% - limited by nozzle 2 BYR-2-B3.110.0001 Indications - 0° was NRI, 45° (VT-2 per Relief Request NR-19)

(VT-2 per Relief nozzle configuration B-D/B3.110 Request I2R-03) configuration and heater identified 1 embedded indication, and heater 60° identified 1 embedded penetrations/NRI penetrations/NRI indication.

9-1993/B2R044 3-2004/B2R11 2RY-01-S/PN-02 10-1985 4-2010/B2R15 Examination Coverage was not 62.3% - limited by 2 BYR-2-B3.110.0002 Coverage was not documented; no 62.3% - limited by nozzle currently scheduled documented; no scan nozzle B-D/B3.110 scan limitations were identified/NRI configuration/NRI for B2R24 (2023) limitations were identified/NRI configuration/NRI 2RY-01-S/PN-03 10-1985 10-1990/B2R02 4-2001/B2R09 4-2010/B2R15 Examination 2 BYR-2-B3.110.0003 Coverage was not documented; no 75% - limited by nozzle 66% - limited by nozzle 66.3% - limited by nozzle currently scheduled B-D/B3.110 scan limitations were identified/NRI configuration/NRI configuration/NRI configuration/NRI for B2R24 (2023) 9-1993/B2R044 2RY-01-S/PN-04 10-1985 3-2004/B2R11 10-2014/B2R18 Examination Coverage was not 2 BYR-2-B3.110.0004 Coverage was not documented; no 62% - limited by nozzle 62.5% - limited by nozzle currently scheduled documented; no scan B-D/B3.110 scan limitations were identified/NRI configuration/NRI configuration/NRI for B2R24 (2023) limitations were identified/NRI

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 15 of 68)

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

Number 3-1995/B2R054 4-2019/B2R21 2RY-01-S/PN-05 10-1985 4-2001/B2R09 10-2014/B2R18 Coverage was not 66% - limited by 2 BYR-2-B3.110.0005 Coverage was not documented; no 66.3% - limited by nozzle 62.5% - limited by nozzle documented; no scan nozzle B-D/B3.110 scan limitations were identified/NRI configuration/NRI configuration/NRI limitations were identified/NRI configuration/NRI 3-1995/B2R05 4 4-2019/B2R21 2RY-01-S/PN-06 10-1985 4-2001/B2R09 4-2010/B2R15 Coverage was not 62% - limited by 2 BYR-2-B3.110.0006 Coverage was not documented; no 66% - limited by nozzle 66.3% - limited by nozzle documented; no scan nozzle B-D/B3.110 scan limitations were identified/NRI configuration/NRI configuration/NRI 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. Exelon 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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 16 of 68)

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:

  • 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.

  • 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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 17 of 68)

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.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 18 of 68)

Table 4. Calvert Cliffs, Units 1 and 2, Pressurizer Welds Inspection History PSI1 Weld ID Ultrasonic Test Interval 1 Interval 2 Interval 3 Interval 4 Interval 5 Summary Number Unit (UT) (Date/Outage, (Date/Outage, (Date/Outage, (Date/Outage, (Date/Outage, Exam Cat, Item (Date, Coverage/Results) Coverage/Results) Coverage/Results) Coverage/Results) Coverage/Results)

Number Coverage/Results) 1 8-411 8-1974 DATA RECORD NOT DATA RECORD NOT 3-2000/1RFO14 2-2010/1RFO19 2-2020/CC1R25 CCNP-1-003900 Coverage was not FOUND5 FOUND5 100%/NRI 100%/NRI 99.7%/NRI B-B/B2.11 documented/NRI 1 3-401 2-1978/1RFO26 8-1974 Examination currently CCNP-1-004000 Coverage was not 3-1994 /1RFO11 4-2004/1RFO16 2-2018/1RFO23 Coverage was not scheduled for CC1R29 B-B/B2.11 documented 99.8%/NRI 90.7%/NRI 99.1%/NRI documented/ NRI (2028)

/NRI 1 2-401A 8-1974 2-1978/1RFO26 DATA RECORD NOT 3-2000/1RFO14 2-2010/1RFO19 2-2020/CC1R25 CCNP-1-003700 Coverage was not Coverage was not FOUND5 96%/NRI 100%/NRI 97.5%/NRI B-B/B2.12 documented/ NRI documented /NRI 1 2-401B 8-1974 11-1980/1RFO46 CCNP-1-003750 Coverage was not Coverage was not NA4 NA4 NA4 NA4 B-B/B2.12 documented/ NRI documented /NRI 1 2-401C 71974 DATA RECORD NOT CCNP-1-003800 Coverage was not NA4 NA4 NA4 NA4 FOUND5 B-B/B2.12 documented/ NRI 1 2-401D 9-1974 Examination currently DATA RECORD NOT 3-1994/1RFO11 4-2004/1RFO16 2-2018/1RFO23 CCNP-1-003850 Coverage was not scheduled for CC1R29 FOUND5 100%/NRI 100%/NRI 94%/NRI B-B/B2.12 documented/ NRI (2028) 1 4-404 7-1974 11-1980/1RFO46 Examination currently 2-1994/1RFO11 4-2004/1RFO163 2-2014/1RFO213 CCNP-1-004050 Coverage was not Coverage was not scheduled for CC1R28 71%/NRI 65.9%/NRI 31.8%/NRI B-D/B3.110 documented/ NRI documented /NRI (2026) 1 4-405 8-1974 2-1978/1RFO26 Examination currently DATA RECORD NOT 3-2000/1RFO14 2-2010/1RFO19 CCNP-1-004100 Coverage was not Coverage was not scheduled for CC1R28 FOUND5 66.4%/NRI 66.4%/NRI B-D/B3.110 documented/ NRI documented /NRI (2026) 1 16-405A 7-1974 Examination currently DATA RECORD NOT DATA RECORD NOT 2-2006/1RFO172 2-2016/1RFO222 CCNP-1-004150 Coverage was not scheduled for CC1R28 FOUND5 FOUND5 36%/NRI 60.5%/NRI B-D/B3.110 documented/ NRI (2026) 1 16-405B 7-1974 Examination currently DATA RECORD NOT 3-1994/1RFO112 2-2006/1RFO172 2-2016/1RFO222 CCNP-1-004200 Coverage was not scheduled for CC1R28 FOUND5 79.7%/NRI 36%/NRI 60.5%/NRI B-D/B3.110 documented/ NRI (2026) 2 8-411 9-1972 2RFO13 DATA RECORD NOT DATA RECORD NOT 2-2011 /2RFO18 3-2021/CC2R24 CCNP-2-103050 Coverage was not DATA RECORD NOT FOUND5 FOUND5 100%/NRI 99.7%/NRI B-B/B2.11 documented/ NRI FOUND5 2 3-401 9-1972 6-1984/2RFO66 Examination currently 4-1993/2RFO9 3-2003/2RFO14 2-2017/2RFO21 CCNP-2-103070 Coverage was not Coverage was not scheduled for CC2R27 100%/NRI 100%/NRI 100%/NRI B-B/B2.11 documented/ NRI documented /NRI (2027)

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 19 of 68)

Table 4. Calvert Cliffs, Units 1 and 2, Pressurizer Welds Inspection History PSI1 Weld ID Ultrasonic Test Interval 1 Interval 2 Interval 3 Interval 4 Interval 5 Summary Number Unit (UT) (Date/Outage, (Date/Outage, (Date/Outage, (Date/Outage, (Date/Outage, Exam Cat, Item (Date, Coverage/Results) Coverage/Results) Coverage/Results) Coverage/Results) Coverage/Results)

Number Coverage/Results) 2 2-401A 9-1972 2RFO13 DATA RECORD NOT DATA RECORD NOT 2-2011/2RFO18 3-2021/CC2R24 CCNP-2-103010 Coverage was not DATA RECORD NOT FOUND5 FOUND5 96%/NRI 96.4%/NRI B-B/B2.12 documented/ NRI FOUND5 2 2-401B 9-1972 DATA RECORD NOT CCNP-2-103020 Coverage was not NA4 NA4 NA4 NA4 FOUND5 B-B/B2.12 documented/ NRI 2 2-401C 9-1972 6-1984/2RFO66 CCNP-2-103030 Coverage was not Coverage was not NA4 NA4 NA4 NA4 B-B/B2.12 documented/ NRI documented /NRI 2 2-401D 9-1972 6-1984/2RFO66 Examination currently DATA RECORD NOT 3-2003/2RFO14 2-2017/2RFO21 CCNP-2-103040 Coverage was not Coverage was not scheduled for CC2R27 FOUND5 100%/NRI 95.3%/NRI B-B/B2.12 documented/ NRI documented /NRI (2027) 2 4-404 9-1972 DATA RECORD NOT 4-1989/2RFO8 4-2001/2RFO13 2-2011/2RFO183 3-2021/CC2R243 CCNP-2-103080 Coverage was not FOUND5 63%/NRI 69.5%/NRI 56%/NRI 28.2%/NRI B-D/B3.110 documented/ NRI 2 4-405 9-1972 6-1984/2RFO66 2RFO13 4-1989/2RFO8 2-2011/2RFO18 3-2021/CC2R24 CCNP-2-103090 Coverage was not Coverage was not DATA RECORD NOT 53%/NRI 65%/NRI 62.1%/NRI B-D/B3.110 documented/ NRI documented /NRI FOUND5 2 16-405A 9-1972 6-1984/2RFO66 4-1993/2RFO92 3-2003/2RFO142 2-2013/2RFO192 3-2021/CC2R242 CCNP-2-103100 Coverage was not Coverage was not 98%/NRI 41%/NRI 58%/NRI 55.3%/NRI B-D/B3.110 documented/ NRI documented /NRI 2 16-405B 9-1972 6-1984/2RFO66 DATA RECORD NOT 3-2007/2RFO162 2-2017/2RFO212 3-2021/CC2R242 CCNP-2-103110 Coverage was not Coverage was not FOUND5 41.9%/NRI 60.5%/NRI 55.3%/NRI B-D/B3.110 documented/ NRI documented /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 Calvert Cliffs units.
2. The variations in examination coverage over the Intervals for welds 16-405A and 16-405B (Units 1 and 2) was not a result of a change in physical limitations, but was due to a change in the methods of calculating coverage.
3. The decrease in examination coverage from the Third Interval to the Fourth Interval (Unit 1) and Fourth Interval to the Fifth Interval (Unit 2) for weld 4-404 was not a result of a change in physical limitations, but was due to a change in the examination boundaries. The weld locations utilized in 2004 (1RFO16) and 2011 (2RFO18) were adjusted in 2014 (1RFO21) and 2021 (2RFO24) to be closer to the nozzle, based on the nozzle design drawing, which increased the limitation.

The adjustment resulted in a drop in coverage.

4. The pressurizers at Calvert Cliffs have two longitudinal welds per shell. For Successive Inspection Intervals, ASME Section XI requires only 1ft of one weld per head to be volumetrically examined.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 20 of 68)

5. Examination records for earlier Intervals could not be located in plant records during the preparation of this proposed alternative; however, recent examination data shows these welds have had acceptable results in accordance with Section XI.
6. Exelon has reviewed the missing coverages for the examinations in the First Interval 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 28.2%. 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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 21 of 68)

During Interval 3, the Calvert Cliffs, Units 1 and 2, pressurizer safety/relief nozzle welds 16-405A and 16-405B (for each unit) had a coverage of 36% (Unit 1) and 41% (Unit 2) (Note: The examinations performed during other Intervals for these welds have resulted in a coverage greater than 60%). Since the limited coverage was identified for the safety/relief nozzle welds, a review of the probability of leakage and rupture for various sensitivity studies from Reference [1] was done for leak paths PRNV-CE-4A/C, which is applicable to welds 16-405A and 16-405B as follows:

  • Section 8.3.4.1.1 and Table 8-11 of Reference [1] discuss the probability of rupture and leakage for the case study where PSI-only was performed. The most limiting crack path for safety/relief nozzle welds (PRNV-CE-4A &

PRNV-CE-4C), yielded a probability of leakage of 1.25x10-9 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 16-405A and 16-405B, 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 36% for welds 16-405A and 16-405B at Calvert Cliffs are bounded by the PSI-only sensitivity study in Reference [1] and the failure frequency for these welds would be below the NRC acceptance criteria of 10-6 failures per year.

  • 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 16-405A and 16-405B, PRNV-CE-4A/C were evaluated to have a probability of leakage of 1.25x10-9 after 80 years.

There is a negligible change in the probability of leakage compared to the base case where 100% coverage was achieved. 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.

Also, during Interval 4 for Calvert Cliffs, Unit 1, and Interval 5 for Calvert Cliffs, Unit 2, pressurizer surge nozzle weld 4-404 had the minimum coverage of 31.8% and 28.2%, respectively. It should be noted that the prior examinations performed for these welds reported a coverage greater than 55%. 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 weld 4-404.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 22 of 68)

  • Section 8.3.4.1.1 and Table 8-11 of Reference [1] discuss the probability of rupture and leakage for the case study where PSI-only was 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 weld 4-404, would reduce the probability of rupture and leakage; therefore, the examination results with coverage as low as 31.8% for weld 4-404 at Calvert Cliffs 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.

  • 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 weld 4-404, 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.

Considering all the findings from Reference [1] it can be concluded that the probability of failure is below 10-6 failures per year for welds 4-404, 16-405A, and 16-405B.

No flaws exceeding the ASME Section XI acceptance standards were identified during any prior examinations for Calvert Cliffs.

Conclusion Based on the results of Reference [1] and its demonstrated applicability to Braidwood, Byron, and Calvert Cliffs, 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, Byron, and Calvert Cliffs is demonstrated in Appendix A. While the technical bases demonstrate longer inspection intervals are possible, Exelon considers that deferral of these inspections until the end of the currently approved Period of Extended Operation (PEO), as shown in Table 1, is justified and provides an acceptable level of quality and safety in lieu of the current ASME Section XI 10-year inspection frequency.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 23 of 68)

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.

Exelon 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 through 4, to-date, Exelon has performed nearly 200 inspections of the subject pressurizer welds at Braidwood, Byron, and Calvert Cliffs, 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 through 4, some of the examinations involved limited coverage of as low as 28.2%. 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, Byron, and Calvert Cliffs pressurizer welds have received the required PSI examinations along with nearly 200 subsequent inservice inspections with no service-induced indications documented.

Therefore, Exelon requests that the NRC authorize this proposed alternative in accordance with 10 CFR 50.55a(z)(1).

6 Duration of Proposed Alternative The proposed alternative is requested for Braidwood, Byron, and Calvert Cliffs for the remainder of their currently approved operating license, currently scheduled to end on October 17, 2046 (Braidwood, Unit 1), December 18, 2047 (Braidwood, Unit 2), October 31, 2044 (Byron, Unit 1), November 6, 2046 (Byron, Unit 2), July 31, 2034 (Calvert Cliffs, Unit 1), and August 13, 2036 (Calvert Cliffs, Unit 2), as summarized in Table 1.

7 Precedent To-date, one previous submittal has been made requesting relief from the ASME Code, Section XI, Examination Category B-B (Item Numbers B2.11 and B2.12) volumetric examinations on the basis of the Reference [1] Technical Report, as follows:

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 24 of 68)

1. Letter from Paul R. Duke, Jr., Manager - Licensing, PSEG Nuclear LLC, to U.S. Nuclear Regulatory Commission Document Control Desk, Proposed Alternative for Examination of ASME Section XI, Examination Category B-B, Item Number B2.11 and B2.12, August 5, 2020, ADAMS Accession No. ML20218A587.
2. Letter from Paul R. Duke, Jr., Manager - Licensing, PSEG Nuclear LLC, to U.S. Nuclear Regulatory Commission Document Control Desk, Response to Request for Additional Information for Proposed Alternative for Examination of ASME Section XI, Examination Category B-B, Item Number B2.11 and B2.12, April 12, 2021, ADAMS Accession No. ML21102A024.

In addition, the following is a list of Proposed Alternatives and other precedents related to inspections of pressurizer welds and components:

  • Letter from M. Lesniak (ComEd) to NRC, Relief from Inservice Inspection Program Requirements for Pressurizer Surge Nozzle-to-Vessel Weld and Pressurizer Surge Nozzle Inner Radius Section, March 28, 1996, ADAMS Accession No. ML20101L096.
  • Letter from R. Capra (NRC) to D. Farrar (Commonwealth Edison Company),

Safety Evaluation of Inservice Inspection Program Relief Request for Inspection of the Pressure Surge Nozzle-to-Vessel Weld; Byron and Braidwood Stations (TAC Nos. M95088, M95089, M95172, and M95173),

May 3, 1996, ADAMS Accession Nos. ML20108D006 and ML20108D016.

  • Letter from M. G. Kowal (NRC) to M. A. Balduzi (Entergy Nuclear Operations, Inc.),

Indian Point Nuclear Generating Unit No. 2 - Relief Request No. RR-01 (TAC No.

MD4695), September 5, 2007, ADAMS Accession No. ML072130487.

  • Letter from T. L. Tate (NRC) to Vice President, Operations (Entergy Nuclear Operations, Inc.), Indian Point Nuclear Generating Unit No. 2 - Safety Evaluation for Relief Request No. IP2-ISI-RR-01, Examination of Upper Pressurizer Welds (CAC No. MF7082), September 14, 2016, ADAMS Accession No. ML16179A178.
  • Letter from N. DiFrancesco (NRC) to M. J. Pacilio (Exelon Generation Company, LLC), Braidwood Station, Units 1 and 2 - Relief from the Requirements of the ASME Code for the Third 10-Year Interval of Inservice Inspection (TAC NOS. ME9748 AND ME9749), January 30, 2013, ADAMS Accession No. ML13016A515.
  • Letter from N. L. Salgado (NRC) to B. C. Hanson (Exelon Generation Company, LLC), Braidwood Station, Units 1 and 2 - Relief from the Requirements of the ASME Code (EPID L-2019-LLR-0081), May 14, 2020, ADAMS Accession No. ML20133K093.
  • Letter from D. J. Wrona (NRC) to B. C. Hanson (Exelon Generation Company, LLC), Byron Station, Units 1 and 2 - Request for Relief Nos.

13R-12 and 13R-15 from the Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (CAC NOS.

MF9884, MF9885, and MF9886; EPID NOS. 000976/05000454/L-2017-LLR-0055, 000976/05000455/L-2017-LLR-0055, AND 000976/05000455/L-2017-LLR-0056), January 25, 2018, ADAMS Accession No. ML17349A960.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 25 of 68)

  • Letter from J. G. Danna (NRC) to D. P. Rhoades (Exelon Generation Company, LLC), Calvert Cliffs Nuclear Power Plant, Units 1 and 2 - Relief from the Requirements of the ASME Code Concerning Volumetric or Surface Examination Coverage for the Subject Welds (EPID L-2020-LLR-0089), January 15, 2021, ADAMS Accession No. ML20356A121.

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

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

[12].

  • 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].

  • 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 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

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 26 of 68) 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.
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.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 27 of 68)

APPENDIX A Plant-Specific Applicability

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 28 of 68)

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)

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

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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 29 of 68)

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

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.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 30 of 68)

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

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].

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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 31 of 68)

Table A-2 Range of Geometric Parameters for which the Evaluation is Applicable in Comparison with Braidwood Station, Units 1 and 2 Braidwood Station, Component Geometric Parameter For a Westinghouse Plant Units 1 and 2 Dimensions Pressurizer Inside Diameter (in) Must be between 80 and 88 84 [A3]

Shell NPS of piping or component (e.g., reducer)

Surge Nozzle 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 component (e.g., reducer)

Spray Nozzle 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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 32 of 68)

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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 33 of 68)

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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 34 of 68)

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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 35 of 68)

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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 36 of 68)

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)

Requirement from Category Applicability to Byron Station, Units 1 and 2 Reference [A5]

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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 37 of 68)

Requirement from Category Applicability to Byron Station, Units 1 and 2 Reference [A5]

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.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 38 of 68)

Requirement from Category Applicability to Byron Station, Units 1 and 2 Reference [A5]

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].

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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 39 of 68)

Table A-4 Range of Geometric Parameters for which the Evaluation is Applicable in Comparison with Byron Station, Units 1 and 2 Byron Station, Units 1 Component Geometric Parameter For a Westinghouse Plant and 2 Dimensions Pressurizer Inside Diameter (in) Must be between 80 and 88 84 [A7]

Shell NPS of piping or component (e.g., reducer)

Surge Nozzle 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 component (e.g., reducer)

Spray Nozzle 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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 40 of 68)

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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 41 of 68)

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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 42 of 68)

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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 43 of 68)

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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 44 of 68)

Plant-Specific Applicability for Calvert Cliffs Section 9 of Reference [A9] 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 Calvert Cliffs, Units 1 and 2 is provided in Table A-5.

Table A-5 indicates that all plant-specific requirements are met for Calvert Cliffs, Units 1 and 2.

Therefore, the results and conclusions of the EPRI report are applicable to Calvert Cliffs, Units 1 and 2.

Table A-5 Plant-Specific Applicability of Reference [A9] Representative Analyses to Calvert Cliffs Nuclear Power Plant, 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)

Requirement from Category Applicability to Calvert Cliffs, Units 1 and 2 Reference [A9]

General The plant-specific In Appendix C of this proposed alternative, the number Requirements pressurizer general and type of the Calvert Cliffs, 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 [A9]. As shown in Table C-6, the those shown in Table 5- Calvert Cliffs, Units 1 and 2, transients are bounded by 6 for a 60- year the transients listed in Table 5-6 of Reference [A9].

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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 45 of 68)

Requirement from Category Applicability to Calvert Cliffs, Units 1 and 2 Reference [A9]

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

alloy ferritic steels which SA-533, Grade B, Class 1 material is specifically listed conform to the in ASME Code, Section XI, Appendix G. SA-533, Grade requirements of ASME B, Class 1 has a specified minimum yield strength at Code, Section XI, room temperature of 50 ksi, and the RTNDT values for Appendix G, Paragraph the Calvert Cliffs, Units 1 and 2, pressurizer top head G-2110.

materials are not greater than 60°F (so the RTNDT of 60°F used in the EPRI report is bounding).

Also, material properties in SA-533, Grade B, Class 1 material are identical to the material used in the FEA in the EPRI report. Therefore, by comparison, SA-533, Grade B, Class 1 material is consistent with the requirements of ASME Code, Section XI, Appendix G and satisfies the requirements for application of the EPRI report.

The Calvert Cliffs, Units 1 and 2, pressurizer shells are fabricated from SA-533, Grade B, Class 1, material. The Calvert Cliffs, Units 1 and 2, pressurizer Surge, Spray, and Safety-Relief valve nozzles are all fabricated from A-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 Calvert Cliffs Units 1 and 2 pressurizer shell-to-Requirements pressurizer surge nozzle head and nozzle-to-vessel weld configurations included and bottom head weld in this request are shown in Figures A-9, A-10, A-11, configurations must and A-12 and show conformance with the figures shown conform to those shown in Reference [A9].

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

[A9].

The plant-specific The comparison of the Calvert Cliffs Units 1 and 2 dimensions of the pressurizer shell dimensions with those in Table 9-1 of pressurizer shell and the Reference [A9] is provided in Table A-6. The surge nozzle must be comparison shows that the Calvert Cliffs 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 [A9].

of Reference [A9].

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 46 of 68)

Requirement from Category Applicability to Calvert Cliffs, Units 1 and 2 Reference [A9]

The plant-specific Calvert Cliffs does not track individual thermal transients Insurge/Outsurge as it is presented in the EPRI Report. Therefore, in transient definitions Appendix D of this proposed alternative, the Calvert (temperature difference Cliffs Units 1 and 2 stress-based fatigue monitoring between the pressurizer values (U and UEN) were projected out to the current shell and the pressurizer end-of-license operating period (60 years) and are surge nozzle fluid compared to the design limit value of 1.0 as shown in temperature and Tables D-5 and D-6. As can be seen from Table D-6, associated number of the Calvert Cliffs Units 1 and 2 fatigue values are well cycles) must be below the design limits and are satisfactory through 60 bounded by those years of operation.

shown in Table 5-10 for a Westinghouse/CE plant of Reference [A9].

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 47 of 68)

Table A-6 Range of Geometric Parameters for which the Evaluation is Applicable in Comparison with Calvert Cliffs, Units 1 and 2 Calvert Cliffs Units 1 Component Geometric Parameter For a CE Plant and 2 Dimensions Pressurizer Inside Diameter (in) Must be between 90 and 102 96 [A10]

Shell NPS of piping or component (e.g., reducer)

Surge Nozzle Must be between 10 and 14 12 [A11]

attached to nozzle safe-end (in)

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

Must be between 4 and 6 4 [A12]

Nozzle attached to nozzle safe-end (in)

NPS of piping or component (e.g., reducer)

Spray Nozzle Must be between 4 and 6 4 [A12]

attached to nozzle safe-end (in)

REFERENCES A9. 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.

A10. Calvert Cliffs, Units 1 and 2, Drawing 12019-0005, 96 I.D. Pressurizer.

A11. Calvert Cliffs, Units 1 and 2, Drawing 12019-0010, Nozzle Details 96 I.D. Pressurizer.

A12. Calvert Cliffs, Units 1 and 2, Drawing 12019-0012, Nozzle Details 96 I.D. Pressurizer.

A13. Calvert Cliffs, Units 1 and 2, ISI Figure A-3.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 48 of 68)

Figure A-9: Calvert Cliffs, Unit 1 Pressurizer Vessel Weld Locations [A13]

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 49 of 68)

Figure A-10: Calvert Cliffs, Unit 2 Pressurizer Vessel Weld Locations [A13]

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 50 of 68)

Figure A-11: Calvert Cliffs, Units 1 and 2 Typical Pressurizer Nozzle-to-Shell Weld Details [A11 & A12]

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 51 of 68)

Figure A-12: Calvert Cliffs, Units 1 and 2 Typical Pressurizer Vessel Shell Weld Details [A10]

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 52 of 68)

APPENDIX B Results of Industry Survey

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 53 of 68)

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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 54 of 68)

APPENDIX C Comparison of Braidwood Station, Byron Station, and Calvert Cliffs General Transients to the Transients Evaluated in EPRI Report

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 55 of 68)

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

Braidwood general transients are tracked by Exelon 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 Exelon 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 Maximum Maximum Transient Name 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 Reactor Trip w/

6 3 5 10 0 0 10 Subsequent SI 50% Step Load See See See Steps See Steps See Steps See Steps 7a, 7 Decrease with Steam Steps 7a, Steps 7a, 7a, 7b, 7c, 7a, 7b, 7c, 7d 7a, 7b, 7c, 7d 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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 56 of 68)

Braidwood Station, Unit 1 Braidwood 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)

Step Load 10%

7b 3 5 2000 5 9 2000 Decrease Unloading, 15%-0%

7c 17 31 500 15 28 500 Power Large Step Load 7d 2 4 200 1 2 200 Decrease 8 Loss of Load 1 2 80 1 2 80 Loss of Offsite AC 9 1 2 40 2 4 40 Power See See Loss of Flow in One RC See Steps See Steps See Steps See Steps 10a, 10 Steps Steps Loop Only (BRW below) 10a, 10b 10a, 10b 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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 57 of 68)

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 Steam Dump, Loss of Load, 360 269 281 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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 58 of 68)

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

Byron general transients are tracked by Exelon 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 Exelon 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 Cycles Cycles Transient Name Up to 60-Year Up to 60-Year (Controlling (Controlling 2020 Projected 2020 Projected 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 2 4 200 3 6 200 with Steam Dump Loss of Load 3 6 80 1 2 80 Loss of Flow in One RC 0 0 80 0 0 80 Loop Only 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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 59 of 68)

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 Steam Dump, Loss of Load, 360 21 27 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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 60 of 68)

Comparison of Calvert Cliffs, Units 1 and 2 General Transients to the Transients Evaluated in Reference [C10]

Calvert Cliffs general transients are tracked by Exelon and the number of cycles encountered as of 2018 for the transients relevant to this request are provided in Table C-5 [C11]. As indicated in Reference [C10], not all the transients tracked by Exelon for Calvert Cliffs are required for this request. Table 5-5 of Reference [C10] identified the significant cycles to be used in the evaluation based on expected cycles from a fleet fatigue monitoring review. The Reference [C10] 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.

Calvert Cliffs would also expect to perform operating leakage testing, instead of hydrostatic testing, following any potential Calvert Cliffs 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 [C10], the actual number of cycles in Table C-5 were projected to 60 years. The comparison of Calvert Cliffs general transients to the requirements in Reference [C10] is shown in Table C-6.

Table C-5 Calvert Cliffs, Units 1 and 2, General Transients Applicable to This Request [C11]

Calvert Cliffs, Unit 1 Calvert Cliffs, Unit 2 Maximum Maximum Cycles Cycles Transient Name Up to 60-Year Up to 60-Year (Controlling (Controlling 2018 Projected 2018 Projected Limit) Limit)

Pressurizer Heat Up 121 152 500 93 120 500 Pressurizer Cooldown 120 149 500 91 117 500 Reactor Trips 134 171(2) 164 112 147 164 Plant Loading 15-100% 107 146 6150 90 123 6150 Power Plant Unloading 100%-15% 101 137 6150 84 114 6150 Power Loss of Load 2 6 40 2 6 40 Partial Loss of RCS Flow 10 12 40 9 12 40 Loss of Offsite AC Power(1) NA NA NA NA NA NA Step Load Decrease 10% 378 457 820 289 360 820 Step Load Increase 10% 382 468 820 279 344 820 Note:

1. Loss of Offsite AC power is not a transient that is required to be analyzed as part of the Calvert Cliffs design basis transients.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 61 of 68)

2. The Reactor Trips Transient for Unit 1 is projected to exceed the design allowable number of cycles for 60-Years.

While the number of this one transient exceeds the allowable number of transient cycles, the 2018 fatigue analysis determined the fatigue usage still remains below the allowable value of 1.0. All locations monitored for fatigue are currently below the design allowable limit of 1.0 and are projected to remain below the design allowable limit of 1.0 through 60 years of operation.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 62 of 68)

Table C-6 Comparison of Calvert Cliffs, Units 1 and 2, General Transients to the Transients Evaluated in Reference [C10]

Number of Calvert Cliffs, Unit 1 Calvert Cliffs, Unit 2 Cycles for 60 60-Year Projections 60-Year Projections Transient Years from from Table C-5 from Table C-5 Reference [C10]

Heatup / Cooldown 300 152/149 120/117 Loss of Load (Sum of Reactor Trips, 50%

Step Load Decrease with Steam Dump, Loss of Load, 360 189 165 Loss of Flow in One RC Loop Only, and Loss of Offsite AC Power Events)

REFERENCES C10. 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.

C11. ECP-18-000545, Fatigue Plant Transient Review.

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 63 of 68)

APPENDIX D Comparison of Braidwood Station, Byron Station, and Calvert Cliffs 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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 64 of 68)

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]

Unit 1 Unit 2 No. Transient Name (1,2,3)

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 (6) 9 PZR I/O SURGE MOPCD250 4 10 15 8 21 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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 65 of 68)

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, Braidwood Station, Unit Cycles Unit 2 Cycles 1 Cycles Projected to 60 T (oF)(1) From Reference Projected to 60 Years Years of Operation

[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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 66 of 68)

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]

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

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, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 67 of 68)

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

10 CFR 50.55a Proposed Alternative I4R-15 for Braidwood Station, Units 1 and 2, Proposed Alternative I4R-21 for Byron Station, Units 1 and 2, and Proposed Alternative ISI-05-016 for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Revision 0 (Page 68 of 68)

Comparison of Calvert Cliffs, Units 1 and 2 Insurge/Outsurge Transients to the Insurge/Outsurge Transients Evaluated in Reference [D15]

Calvert Cliffs does not track individual thermal transients as it is presented in the EPRI Report. Instead, the plant performs stress-based fatigue monitoring for the relevant surge nozzle locations that tracks Fatigue Usage (U) and Environmental Assisted Fatigue Usage (UEN) values to ensure they remain below the allowable value of 1.0 as shown in Tables D-5 and D-6 [D16]. As a result, instead of tabulated transients, the current U and UEN estimates are provided for the subject welds for Calvert Cliffs and their projections out to the current end-of-license operating period (60 years). The results of Table D-5 and D-6 indicate that the actual Calvert Cliffs fatigue values are well below the design limits and are satisfactory through 60 years of operation.

Table D-5 General Transients Fatigue Usage (U) Summary Report for Calvert Cliffs, Units 1 and 2 [D16]

Unit 1 Unit 2 U U U U Pressurizer U U 60-Year Maximum 60-Year Maximum Locations Up to Up to Projected Cycles Projected Cycles 12/2018 12/2018 (7/21/2034) (Design Limit) (7/21/2034) (Design Limit)

Lower Head 0.1853 0.221 1.0 0.1168 0.151 1.0 Surge Nozzle 0.0173 0.019 1.0 0.0746 0.093 1.0 Notes:

1. The U values projected out to 60 years is calculated with the Fatigue Pro Software utilized by Calvert Cliffs.

Table D-6 General Transients Environmental Assisted Fatigue Usage (UEN) Summary Report for Calvert Cliffs, Units 1 and 2 [D15]

Unit 1 Unit 2 Pressurizer UEN UEN UEN UEN UEN UEN 60-Year Maximum 60-Year Maximum Locations Up to Up to Projected Cycles Projected Cycles 12/2018 12/2018 (7/21/2034) (Design Limit) (7/21/2034) (Design Limit)

Lower Head 0.4548 0.543 1.0 0.5287 0.685 1.0 Surge Nozzle 0.0876 0.096 1.0 0.3806 0.495 1.0 Notes:

1. The UEN values projected out to 60 years is calculated with the Fatigue Pro Software utilized by Calvert Cliffs.

REFERENCES D15. 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.

D16. ECP-18-000545, Fatigue Plant Transient Review.

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