RA-21-0144, Proposed Alternative to Use Reactor Vessel Head Penetration Embedded Flaw Repair for Life of Plant

Proposed Alternative to Use Reactor Vessel Head Penetration Embedded Flaw Repair for Life of Plant

ML22020A283
Person / Time
Site:	Catawba
Issue date:	01/20/2022
From:	Simril T Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML22020A282	List: RA-21-0144, Proposed Alternative to Use Reactor Vessel Head Penetration Embedded Flaw Repair for Life of Plant
References
RA-21-0144 WCAP-18708-NP, Rev 0
Download: ML22020A283 (59)

Text

Tom Simril Vice President Catawba Nuclear Station Duke Energy CN01VP / 4800 Concord Road York, SC 29745 o: 803.701.3340 f: 803.701.3221 Tom.Simril@duke-energy.com PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 1 THIS LETTER IS UNCONTROLLED 10 CFR 50.55a Serial: RA-21-0144 January 20, 2022 U.S. Nuclear Regulatory Commission (NRC)

ATTN: Document Control Desk Washington, DC 20555-0001 Catawba Nuclear Station, Unit No. 2 Renewed Facility Operating License No. NPF-52 Docket No. 50-414

SUBJECT:

Proposed Alternative to Use Reactor Vessel Head Penetration Embedded Flaw Repair for Life of Plant

References:

1. Safety Evaluation Report, Catawba Nuclear Station, Unit 2 - Proposed Alternative Request RA-21-0145 to Use Reactor Vessel Head Penetration Embedded Flaw Repair Method (EPID L-2021-LLR-0028), dated September 20, 2021 [ADAMS Accession No. ML21253A082]

Ladies and Gentlemen:

Pursuant to 10 CFR 50.55a(z)(1), Duke Energy Carolinas, LLC (Duke Energy), requests NRC approval for Catawba Nuclear Station (CNS), Unit 2, of a proposed alternative to ASME Code repair and replace requirements, on the basis that the proposed alternative provides an acceptable level of quality and safety. Specifically, Duke Energy is requesting to utilize an alternative to the requirements of American Society of Mechanical Engineers (ASME) Code Section XI, IWA-4000 and ASME Code Section III, NB-4450 for an embedded flaw repair method implemented on CNS Unit 2 Reactor Vessel Closure Head (RVCH) Penetration #74 during the Spring, 2021 outage.

In Reference 1 the NRC staff authorized for CNS Unit 2 the use of an embedded flaw repair method to repair a relevant indication that was identified turing a Non-Destructive Examination (NDE) of the J-groove weld surface for RVCH penetration #74. NRC staff authorized use of the embedded flaw repair method for one cycle of operation at CNS Unit 2, until the end of Cycle 25 that is scheduled to end in Fall, 2022. Subsequently, Duke Energy hereby requests NRC approval to utilize the alternative repair beyond the current cycle for CNS Unit 2, RVCH Penetration #74.

The proposed alternative is provided in the Enclosure to this letter. The proposed alternative is based on a fracture mechanics analysis which provides a basis for extended use of the PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 1 THIS LETTER IS UNCONTROLLED

U.S. Nuclear Regulatory Commission RA-21-0144 Page 2 embedded flaw repair for RVCH penetration #74. The fracture mechanics analysis is provided as Attachment 1. Attachment 1 contains information proprietary to Westinghouse and is requested to be withheld from public acces under 10 CFR 2.390. A non-proprietary version of the fracture mechanics analysis is provided as Attachment 2. An affidavit attesting to the proprietary nature of Attachment 1 is included as Attachment 3.

A presubmittal meeting with the NRC staff was held on December 7, 2021. Because Cycle 25 for CNS Unit 2 is scheduled to end on September 10, 2022, Duke Energy is requesting NRC review and approval of this alternative by September 10, 2022.

Should you have any questions concerning this letter and its enclosure, please contact Mr. Lee Grzeck, Fleet Licensing Manager (Acting), at (980) 373-1530.

Sincerely,

--..::> */'\

Tom Simril Vice President, Catawba Nuclear Station

Enclosure:

Proposed Alternative to Use Reactor Vessel Head Penetration Embedded Flaw Repair for Life of Plant

Attachment:

1. WCAP-18708-P, "Technical Basis for Westinghouse Embedded Flaw Repair of Catawba Unit 2 Reactor Vessel Head Penetration Nozzles and Attachment Welds" (Proprietary)

2. WCAP-18708-NP, "Technical Basis for Westinghouse Embedded Flaw Repair of Catawba Unit 2 Reactor Vessel Head Penetration Nozzles and Attachment Welds" (Non-Proprietary)

3. Affidavit Attesting to Proprietary Nature of Information in Attachment 1 cc:

L. Dudes, USNRC, Region II Regional Administrator Z. Stone, USNRC NRR Project Manager for CNS J. Austin, USNRC Senior Resident Inspector for CNS PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 1 THIS LETTER IS UNCONTROLLED

Enclosure RA-21-0144 Enclosure Proposed Alternative to Use Reactor Vessel Head Penetration Embedded Flaw Repair for Life of Plant

Relief Request RA-21-0144 Enclosure Page 1 of 8 1.0 ASME CODE COMPONENT(S) AFFECTED:

Component: Reactor Pressure Vessel (RPV) Head Penetration #74 Code Class: 1 Examination Category: ASME Code Case N-729-6 (Reference 8.1)

Item Number: B4.20 2.0 APPLICABLE CODE EDITION AND ADDENDA:

The applicable Edition and Addenda of the ASME Code,Section XI (Reference 8.2) is identified in Table 1.

Table 1 Current Current Interval Current Current Current License Plant/Unit(s) ISI ASME Section XI Interval Interval End Date Interval Code Edition/Addenda Start Date End Date1 Catawba 2007 Edition, Through Nuclear Station Fourth 8/19/2015 2/24/2026 12/5/2043 2008 Addenda Unit 2 (CNS2)

Notes:

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

Examinations of the Reactor Vessel Closure Head (RVCH) penetrations are performed in accordance with 10 CFR 50.55a(g)(6)(ii)(D), which currently specifies the use of ASME Code Case N-729-6, with conditions.

The RPV Construction Code is ASME Section III, 1971 Edition through Winter 1972 Addenda (Reference 8.3).

The Construction Code applicable to the repair of weld defects on the RVCH at CNS2 is the 1989 Edition of ASME Section III (Reference 8.4).

3.0 APPLICABLE CODE REQUIREMENT:

IWA-4000 of ASME Section XI contains requirements for the removal of defects from and welded repairs performed on ASME components. The specific Code requirements for which use of the proposed alternative is being requested are as follows:

ASME Section XI, IWA-4421 states:

Defects shall be removed or mitigated in accordance with the following requirements:

Relief Request RA-21-0144 Enclosure Page 2 of 8 (a) Defect removal by mechanical processing shall be in accordance with IWA-4462.

(b) Defect removal by thermal methods shall be in accordance with IWA-4461.

(c) Defect removal or mitigation by welding or brazing shall be in accordance with IWA-4411.

(d) Defect removal or mitigation by modification shall be in accordance with IWA-4340.

• Note that the use of the Mitigation of Defects by Modification provisions of IWA-4340 is prohibited per 10 CFR 50.55a(b)(2)(xxv)(A).

For the removal or mitigation of defects by welding, ASME Section XI, IWA-4411 states, in part:

Welding, brazing, fabrication, and installation shall be performed in accordance with the Owners Requirements and, except as modified below, in accordance with the Construction Code of the item.

(a) Later editions and addenda of the Construction Code, or a later different Construction Code, either in its entirety or portions thereof, and Code Cases may be used provided the substitution is as listed in IWA-4221(c). Filler metal requirements shall be reconciled, as required, in accordance with IWA-4224.

The Construction Code used for performing repair of welds on the RVCH at CNS2 is the 1989 Edition of ASME Section III. Requirements applicable to repair of weld defects are specified in NB-4450 as follows:

• NB-4451, which provides general requirements for repair of weld metal defects, states:

Defects in weld metal detected by the examinations required by NB-5000, or by the tests of NB-6000, shall be eliminated and repaired when necessary.

• NB-4452, which provides requirements for eliminating weld surface defects without welding states:

Weld metal surface defects may be removed by grinding or machining, and need not be repaired by welding, provided that the requirements of (a) through (c) below are met.

(a) The remaining thickness of the section is not reduced below that required by NB-3000.

(b) The depression, after defect elimination, is blended uniformly into the surrounding surface.

(c) The area is examined by a magnetic particle or liquid penetrant method in accordance with NB-5110 after blending and meets the acceptance standards of NB-5300 to ensure that the defect has been removed or reduced to an imperfection of acceptable limit. Defects detected by visual or volumetric method and located on the interior surface need only be reexamined by the method which initially detected the defect when the interior surface is inaccessible for surface examination.

• NB-4453.1, which provides requirements for repairing weld defects with welding, states:

Relief Request RA-21-0144 Enclosure Page 3 of 8 Defects may be removed by mechanical means or by thermal gouging processes. The area prepared for repair shall be examined by a liquid penetrant or magnetic particle method in accordance with NB-5110, and meet the acceptance standards of NB-5340 or NB-5350. This examination is not required where defect elimination removes the full thickness of the weld and where the backside of the weld joint is not accessible for removal of examination materials.

4.0 REASON FOR REQUEST:

During refueling outage C2R24, Catawba performed a volumetric examination of the RVCH penetrations in accordance with ASME Code Case N-729-6 (Item No. B4.20) as conditioned by 10CFR50.55a(g)(6)(ii)(D). Due to the discovery of a relevant indication in the J-groove weld at RVCH penetration #74 (References 8.5, 8.6, and 8.7), a weld repair was necessary. As an alternative to the defect removal and weld repair provisions of ASME Section XI, IWA-4000 and ASME Section III, NB-4450, Duke Energy submitted Proposed Alternative RA-21-0145 (Reference 8.12) on April 24, 2021 which proposed repair of the subject J-groove weld using the embedded flaw repair (EFR) process as described in WCAP-15987-P, Revision 2-P-A (Reference 8.8). The duration of Proposed Alternative RA-21-0145 was one cycle of operation until the completion of Cycle 25 (Fall 2022). NRC verbal authorization of Proposed Alternative RA 0145 was provided on April 24, 2021 (Reference 8.13) and the EFR was implemented on RVCH penetration #74. The NRC Safety Evaluation was issued on September 20, 2021 (Reference 8.14).

Relief is requested beyond Cycle 25 from the requirements of ASME Section XI, IWA-4000 and ASME Section III, NB-4450 for the EFR implemented on CNS2 RVCH penetration #74 for the duration specified in Section 6.0.

5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE:

Pursuant to 10 CFR 50.55a(z)(1), Duke Energy proposes continued use of the EFR for CNS2 RVCH penetration #74 as an alternative to the defect removal and weld repair provisions of ASME Section XI, IWA-4000 and ASME Section III, NB-4450. The EFR process is described in WCAP-15987-P and was approved by the NRC in Reference 8.9. This proposed alternative is requested for use beyond Cycle 25 for the duration specified in Section 6.0 herein.

Basis for Use:

Duke Energys review concludes that the EFR process based on WCAP-15987-P provides an acceptable level of quality and safety. In the Safety Evaluation Report for WCAP-15987-P, the NRC documented the same conclusion subject to their specified conditions and limitations (Reference 8.9). The EFR and post-repair nondestructive examinations (NDE) of RVCH penetration #74 were implemented during refueling outage C2R24 to the requirements of Proposed Alternative RA-21-0145 consistent with the WCAP-15987-P methodology. Below are the NDE requirements and technical basis for continued use of the RVCH penetration #74 EFR beyond Cycle 25.

5.1 Nondestructive Examinations Requirements In lieu of the examination requirements identified for J-groove welds in the NRC staff safety evaluation of WCAP-15987-P, inservice non-destructive examination (NDE) of the completed EFR will be performed as specified below.

Relief Request RA-21-0144 Enclosure Page 4 of 8 Table 2 Repair Location Flaw Orientation Repair Method ISI NDE UT and Surface J-groove weld Circumferential Seal Weld (Notes 1 and 2)

Notes:

(1) Ultrasonic Testing (UT) to be consistent with 10 CFR 50.55a(g)(6)(ii)(D), which requires implementation of Code Case N-729-6 with conditions, or NRC-approved alternatives.

a. Ultrasonic (UT) personnel and procedures are qualified in accordance with 10 CFR 50.55a(g)(6)(ii)(D), which requires implementation of Code Case N-729-6 with conditions, or NRC-approved alternatives.

b. UT inspection of the nozzle tube shall be performed during the refueling outage following implementation of the repair. Subsequent UT inspection frequency shall be consistent with 10 CFR 50.55a(g)(6)(ii)(D), which requires implementation of Code Case N-729-6 with conditions, or NRC-approved alternatives.

(2) Surface examination of the EFR shall be performed to ensure the repair satisfies ASME Section III, NB-5350 acceptance standards. The frequency of surface examination shall be as follows:

a. Surface examination shall be performed during the first and second refueling outage after installation or repair of the EFR.

b. When the examination results in 2.a above verify acceptable results then re-inspection of the EFR will be continued at a frequency of every other refueling outage. If these examinations identify unacceptable results that require flaw removal, flaw reduction to acceptable dimensions, or welded repair, the requirements of 2.a above shall be applied during the next refueling outage.

Note 2 above permits a reinspection frequency of every other cycle when the surface examination results of the EFR are verified to be acceptable for two consecutive cycles after the original installation or repair of the EFR. Westinghouse Report LTR-PSDR-TAM-14-005, Revision 3 (Reference 8.10) provides the technical bases for reducing surface examination requirements for J-groove weld repairs. This technical justification includes a detailed review of PT examination history, review of potential causes of PT indications in EFRs, and the use of crack resistant alloys in the EFR. The EFR is a robust design that is resistant to primary water stress corrosion cracking (PWSCC). EFR installation, examination, and operational history indicate that the EFR performs acceptably.

The proposed inservice examination requirements assure that the EFR repaired nozzle is adequately monitored through a combination of volumetric and surface examinations throughout the life of the installation at a frequency approved by the NRC, thus ensuring the EFR repaired nozzle will continue to perform its required function.

5.2 Technical Basis for Proposed Alternative The purpose of the proposed alternative is to embed and isolate the Alloy 600 RVCH penetration J-groove weld and identified flaw with a non-structural, corrosion resistant Alloy 52/52M seal

Relief Request RA-21-0144 Enclosure Page 5 of 8 weld (overlay). As discussed in WCAP-15987-P, the EFR technique is considered a permanent repair. As long as a PWSCC flaw remains isolated from the primary system water environment, it cannot propagate. Alloy 52/52M is considered highly resistant to PWSCC and is intended to preclude new PWSCC flaw initiation and growth through the overlay to prevent the primary coolant environment from interacting with the susceptible material. Structural integrity of the affected J-groove weld will be maintained by the remaining unflawed portion of the weld. The resistance of Alloy 52/52M weld metal to PWSCC has been demonstrated by laboratory tests and operating experience in replacement steam generators.

As stated in WCAP-15987-P, the residual stresses produced by the embedded flaw technique have been measured and found to be relatively low because of the small seal weld thickness.

This implies that no new flaws will initiate and grow in the area adjacent to the repair weld.

There are no other known mechanisms for significant flaw propagation in the reactor vessel closure head and penetration tube region since cyclic loading is negligible. Section 3.2 of the Safety Evaluation for WCAP-15987-P highlights this conclusion and its basis which is provided in WCAP-13998, Revision 1 (Reference 8.11).

The thermal expansion properties of Alloy 52 or 52M weld metal are not specified in the ASME Code. In this case, the properties of the equivalent base metal (Alloy 690) should be used. For Alloy 690, the thermal expansion coefficient at 600 degrees Fahrenheit (F) is 8.2E-6 in/in/degree F as found in Section II Part D. A five percent difference in coefficient of thermal expansion exists between Alloy 690 and the Alloy 600 base material which has a coefficient of thermal expansion of 7.8 E-6 in/in/degree F. This small difference in thermal expansion will cause the weld metal to contract more than the base metal when it cools, thus producing a compressive stress on the Alloy 600 J-groove weld. The benefit of the compressive stress effect has been accounted for in the residual stress measurements reported in the technical basis for the EFR, as noted in the WCAP-15987-P.

Technical basis document WCAP-18708-P, Revision 0 (Reference 8.15) provides the plant-specific technical basis for the CNS2 EFR consistent with Appendix C of WCAP-15987-P.

Section 3 of WCAP-18708-P provides details of the evaluation for an EFR applied to a flawed head attachment j-groove weld, including the weld repair thickness considered in the analysis.

This evaluation demonstrated that a flaw postulated in the RVCH penetration J-groove weld is not expected to grow through the EFR seal weld thickness in less than 47 years of plant life.

Since the CNS2 RVCH penetration #74 EFR was installed in April of 2021, the service life of the repair bounds the duration of the remaining licensed plant life of Catawba Unit 2 for the period of extended operation (PEO) through December 5, 2043.

Duke Energys review concludes that the EFR implemented on RVCH penetration #74 is a technically sound alternative to performing an ASME Section XI Code repair in accordance with IWA-4000 and ASME Section III, NB-4450. The EFR isolates the PWSCC susceptible Alloy 82/182 J-groove weld from the primary water environment by deposition of a 360-degree overlay consisting of Alloy 52 or 52M weld metal. Because Alloy 52/52M welds are considered highly resistant to PWSCC, a new PWSCC flaw should not initiate and grow through the Alloy 52/52M seal weld to reconnect the primary water environment to the embedded flaw.

Furthermore, the existing flaw in the J-groove weld can no longer grow by PWSCC since the weld is isolated from the reactor coolant environment. As described in Section 5.1, inservice inspection will be performed using both surface and UT examination methods ensuring the continued structural integrity of the EFR.

Relief Request RA-21-0144 Enclosure Page 6 of 8 The above proposed alternative, as supported by the plant-specific technical basis within WCAP-18708-P, provides an acceptable level of quality and safety as required by 10 CFR 50.55a(z)(1).

6.0 DURATION OF PROPOSED ALTERNATIVE:

Typically, the duration for requests approved under the provisions of 10 CFR 50.55a(z) are limited to the active inservice inspection ten-year interval for which they are submitted. However, there are instances where approval has been granted for requests under the provisions of 10 CFR 50.55a(z) for durations longer than the 10 years associated with an inservice inspection interval (for example, Reference 8.16 & 8.17). As stated herein, the plant-specific technical basis within WCAP-18708-P determined the CNS2 RVCH penetration #74 EFR has a service life of at least 47 years, which bounds the duration of the remaining licensed plant life for Catawba Unit 2 for the period of extended operation (PEO) through December 5, 2043. Therefore, Duke Energy requests approval of this proposed alternative for the remainder of licensed plant life through the PEO which ends on December 5, 2043.

7.0 PRECEDENTS

7.1 ADAMS Accession Number ML18227A733. NRC approval dated August 27, 2018.

Beaver Valley, Unit No. 2 - Request for Relief from the Requirements of the ASME Code (EPID L-2018-LLR-0026).

7.2 ADAMS Accession Number ML18142A431. NRC approval dated May 31, 2018. Indian Point Nuclear Generating Unit No. 2 - Safety Evaluation for Relief Request IP2-ISI-RR-06 Regarding Approval of Alternative to use Embedded Weld Repair (EPID L-2018-LLR-0050).

7.3 ADAMS Accession Number ML17062A428. NRC approval dated March 6, 2017. Byron Station, Unit Nos. 1 and 2 - Request for Relief from the Requirements of the ASME Code (CAC NOS. MF8282 and MF8283).

7.4 ADAMS Accession Number ML14107A332. NRC approval dated April 30, 2014. Virgil C. Summer Nuclear Station, Unit 1 - Alternative Request Weld Repair for Reactor Vessel Upper Head Penetrations (TAC NO. MF3546).

7.5 ADAMS Accession Number ML080280033. NRC approval dated February 14, 2008.

Indian Point Nuclear Generating Unit No. 2 - Relief Request (RR) No. RR-07 on Embedded Flaw Weld Repair (TAC NO. MD4702).

8.0 REFERENCES

8.1 ASME Code Case N-729-6, Alternative Examination Requirements for PWR Reactor Vessel Upper Heads with Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1, American Society of Mechanical Engineers, New York, Approved March 3, 2016.

8.2 ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2007 Edition with the 2008 Addenda.

8.3 ASME Boiler and Pressure Vessel Code,Section III, Rules for Construction of Nuclear Power Plant Components, 1971 Edition through Winter 1972 Addenda.

Relief Request RA-21-0144 Enclosure Page 7 of 8 8.4 ASME Boiler and Pressure Vessel Code,Section III, Rules for Construction of Nuclear Power Plant Components, 1989 Edition.

8.5 RVCH ISI INR C2R24-001 Rev 4, Catawba Unit 2 (C2R24) Reactor Vessel Closure Head Penetration In-Service Inspection Indication Notification Report, Reactor Vessel Closure Head, Penetration ID: 74, Change in back-wall reflectivity (UTLP), dated April 22, 2021.

8.6 RVCH ECT INR C2R24-001 Rev 2, Catawba Unit 2 (C2R24) Reactor Vessel Closure Head Penetration In-Service Inspection Indication Notification Report, Reactor Vessel Closure Head, Penetration ID: 74, CRDM Nozzle to RVCH Shell Weld Surface, dated April 21, 2021.

8.7 RVCH PT INR C2R24-001 Rev 0, Catawba Unit 2 (C2R24) Reactor Vessel Closure Head Penetration In-Service Inspection PT Indication Notification Report, Reactor Vessel Closure Head, Penetration ID: 74, dated April 22, 2021.

8.8 Westinghouse Topical Report, WCAP-15987-P, Revision 2-P-A, Technical Basis for the Embedded Flaw Process for Repair of Reactor Vessel Head Penetrations, dated December 2003.

8.9 Letter from H. N. Berkow (U.S. NRC) to H. A. Sepp (Westinghouse Electric Company),

Acceptance for Referencing -Topical Report WCAP-15987-P, Revision 2, Technical Basis for the Embedded Flaw Process for Repair of Reactor Vessel Head Penetrations, (TAC No.

MB8997), dated July 3, 2003, ADAMS Accession Number ML031840237.

8.10 LTR-PSDR- TAM-14-005-NP, Revision 3, Technical Basis for Optimization or Elimination of Liquid Penetrant Exams for the Embedded Flaw Repair, May 2015.

8.11 WCAP-13998, Revision 1, RV Closure Head Penetration Tube ID Weld Overlay Repair, November 1995.

8.12 Duke Energy Letter RA-21-0145, Revision to Proposed Alternative to Use Reactor Vessel Head Penetration Embedded Flaw Repair Method, April 24, 2021, ADAMS Accession Number ML21114A000.

8.13 Verbal Authorization by the Office of Nuclear Reactor Regulation for Proposed Alternative RA-21-0145, Alternate Repair of a Reactor Vessel Head Penetration 74, Duke Energy Carolinas, LLC, Catawba Unit 2 Docket No. 50-414, April 24, 2021, ADAMS Accession Number ML21117A129.

8.14 Catawba Nuclear Station, Unit 2 - Proposed Alternative Request RA-21-0145 to use Reactor Vessel Head Penetration Embedded Flaw Repair Method (EPID L-2021-LLR-0028),

September 20, 2021, ADAMS Accession Number ML21253A082.

8.15 WCAP-18708-P, Revision 0, Technical Basis for Westinghouse Embedded Flaw Repair of Catawba Unit 2 Reactor Vessel Head Penetration Nozzles and Attachment Welds, December 2021.

Relief Request RA-21-0144 Enclosure Page 8 of 8 8.16 Brunswick Steam Electric Plant, Units 1 and 2 - Request for Alternative to Inservice Inspection of Reactor Pressure Vessel Welds and Nozzle Inner Radius Sections (EPID L-2020-LLR-0091), February 2, 2021, ADAMS Accession No. ML21005A010 8.17 Sequoyah Nuclear Plant, Units 1 and 2 - Alternative Request 18-ISI-1 Regarding Examination of Dissimilar Metal Welds in Reactor Vessel Head (EPID L-2019-LLR-0006),

September 23, 2019, ADAMS Accession No. ML19234A013.

RA-21-0144 Attachment 1 WCAP-18708-P, Technical Basis for Westinghouse Embedded Flaw Repair of Catawba Unit 2 Reactor Vessel Head Penetration Nozzles and Attachment Welds (Proprietary)

RA-21-0144 Attachment 2 WCAP-18708-NP, Technical Basis for Westinghouse Embedded Flaw Repair of Catawba Unit 2 Reactor Vessel Head Penetration Nozzles and Attachment Welds (Non-Proprietary)

Westinghouse Non-Proprietary Class 3 WCAP-18708-NP December 2021 Revision 0 Technical Basis for Westinghouse Embedded Flaw Repair of Catawba Unit 2 Reactor Vessel Head Penetration Nozzles and Attachment Welds

• • • This record was final approved on 12/10/2021, 10:42:40 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 WCAP-18708-NP Revision 0 Technical Basis for Westinghouse Embedded Flaw Repair of Catawba Unit 2 Reactor Vessel Head Penetration Nozzles and Attachment Welds December 2021 Maria Rizzilli*

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RV/CV Design and Analysis Reviewers: Anees Udyawar*

RV/CV Design and Analysis Approved: Lynn A. Patterson, Manager*

RV/CV Design and Analysis

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• • • This record was final approved on 12/10/2021, 10:42:40 AM. (This statement was added by the PRIME system upon its validation)

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WCAP-18708-NP December 2021 Revision 0

• • • This record was final approved on 12/10/2021, 10:42:40 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 iii RECORD OF REVISIONS Revision Date Revision Description December 0 Original Issue 2021 WCAP-18708-NP December 2021 Revision 0

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Westinghouse Non-Proprietary Class 3 iv TABLE OF CONTENTS FOREWORD ................................................................................................................................................ ii 1 INTRODUCTION ........................................................................................................................ 1-1 2 TECHNICAL BASIS FOR APPLICATION OF EMBEDDED FLAW REPAIR TECHNIQUE TO PENETRATION NOZZLES......................................................................................................... 2-1 2.1 ACCEPTANCE CRITERIA ......................................................................................................... 2-2 2.1.1 Axial Flaws ................................................................................................................................... 2-2 2.1.2 Circumferential Flaws................................................................................................................... 2-4 2.2 METHODOLOGY ....................................................................................................................... 2-6 2.2.1 Geometry and Material ................................................................................................................. 2-6 2.2.2 Finite Element Analysis ................................................................................................................ 2-6 2.2.3 Loading Conditions....................................................................................................................... 2-7 2.2.4 Allowable Flaw Size Determination ............................................................................................. 2-9 2.2.5 Stress Intensity Factors ............................................................................................................... 2-10 2.2.6 Fatigue Crack Growth Prediction ............................................................................................... 2-11 2.3 FRACTURE MECHANICS ANALYSIS RESULTS ................................................................. 2-12 2.3.1 Maximum End-of-Evaluation Period Flaw Sizes ....................................................................... 2-12 2.3.2 Allowable Initial Flaw Sizes for Penetration Nozzles ................................................................ 2-12 3 TECHNICAL BASIS FOR APPLICATION OF EMBEDDED FLAW REPAIR TECHNIQUE TO ATTACHMENT J-GROOVE WELD ........................................................................................... 3-1 3.1 ACCEPTANCE CRITERIA ......................................................................................................... 3-1 3.1.1 Section XI Appendix K ................................................................................................................. 3-1 3.1.2 Primary Stress Limits.................................................................................................................... 3-2 3.2 METHODOLOGY ....................................................................................................................... 3-2 3.2.1 Geometry and Material ................................................................................................................. 3-3 3.2.2 Loading Conditions....................................................................................................................... 3-5 3.2.3 Stress Intensity Factors ................................................................................................................. 3-5 3.2.4 J-R curve for Reactor Vessel Closure Head Material.................................................................... 3-6 3.2.5 Applied J-Integral ......................................................................................................................... 3-7 3.2.6 Fatigue Crack Growth Prediction ................................................................................................. 3-8 3.3 FRACTURE MECHANICS ANALYSIS RESULTS ................................................................... 3-9 3.3.1 Results for Applied J-Integral and J-R Curve ............................................................................... 3-9 3.3.2 Results for Fatigue Crack Growth into the Reactor Vessel Head ............................................... 3-14 3.3.3 Results for Fatigue Crack Growth into the Repair Weld ............................................................ 3-15 4

SUMMARY

AND CONCLUSIONS ............................................................................................ 4-1 5 REFERENCES ............................................................................................................................. 5-1 WCAP-18708-NP December 2021 Revision 0

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Westinghouse Non-Proprietary Class 3 1-1 1 INTRODUCTION Leakage and cracks have been reported from the reactor vessel closure head penetration nozzles in a number of plants that resulted in repairs or prompted the replacement of the reactor vessel closure head. The degradation of the reactor vessel closure head penetration nozzles increases the probability of a more significant loss of reactor coolant pressure boundary. This has led to the issuance of various regulatory requirements and guidelines in the United States imposing additional volumetric and surface examinations to supplement the existing visual inspections of the reactor vessel closure head as well as the penetration nozzles. The presence of axial cracks extending above and below the head penetration nozzle attachment J-groove welds was discovered in some of the leaking penetration nozzles. The cause of these axially oriented cracks has been determined to result from primary water stress corrosion cracking (PWSCC) that is driven by both the steady state operating stress and the residual stress resulting from the weld fabrication process. [

]a,c,e As a part of the inspection and repair efforts associated with the reactor vessel closure head inspection program at Catawba Unit 2, engineering evaluations have been performed in this report to support plant-specific use of the Westinghouse embedded flaw repair process in the repair of unacceptable flaws. [

]a,c,e The analysis performed in this report provides technical justification for long term plant operation for Catawba Unit 2 reactor vessel head penetration 74 and all other penetrations where an embedded flaw repair is applied.

[

]a,c,e Engineering evaluations were performed to determine the maximum flaw sizes that would satisfy the requirements in Section XI of the ASME Code [2] and be suitable to support the weld repair process.

The results presented in this report would enable the weld repair team to effectively determine the appropriate repair method.

Section XI repair rules allow the use of grinding to remove flaws, regardless of the edition of the Code.

The only requirement is to ensure that the excavated region still meets the stress limits of the original construction code, which wasSection III. Evaluations were performed in [3] to provide repair guidelines that may be used for removal of defects found on the surfaces of J-groove attachment welds and associated nozzles for the Catawba Unit 2 control rod drive mechanism (CRDM) penetrations.

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Westinghouse Non-Proprietary Class 3 1-2 The technical basis of the embedded flaw repair process is documented in WCAP-15987-P [4], which has been reviewed and accepted by the NRC. The staff also concluded that WCAP-15987-P is acceptable for referencing in licensing applications. As discussed in Appendix C of WCAP-15987-P, Westinghouse has developed the following three repair scenarios/method to address the most common types of flaws during the vessel head inspection:

Scenario 1: Axial or circumferential crack in the penetration nozzle inner surface Scenario 2: Postulated crack encompassing the entire penetration J-groove weld Scenario 3: Axial or circumferential crack in the penetration nozzle outer surface Figure 1-1 shows the repair for Scenario 1, and Figure 1-2 shows the repair for Scenario 2 and 3.

The purpose of this report is to provide plant-specific technical basis for the use of the embedded flaw repair process and to confirm that Catawba Unit 2 meets the criteria for application of the embedded flaw repair process stated in Appendix C of WCAP-15987-P [4]. Engineering evaluations were performed and the results are presented in this report to provide the maximum allowable initial embedded flaw sizes that could be repaired using the Westinghouse embedded flaw repair process and would satisfy the requirements in Section XI of the ASME Code [2]. The ASME Section XI Code of record for Catawba Unit 2 is 2007 Edition with 2008 Addenda [2]. Note that the methodology used in this report from the 2007 Edition with 2008 Addenda is the same up to the 2017 Edition of ASME Section XI Code, which is the most recent ASME Code edition approved by the NRC. The results presented in this report would support the use of the Westinghouse embedded flaw repair process as the repair option for all the Catawba Unit 2 reactor vessel head penetration nozzles. In this report, the technical basis and evaluation results to support the use of embedded flaw repair process for a flawed head penetration are provided in Section 2. The technical basis and evaluation results that support a similar application for a flawed head penetration nozzle attachment J-goove weld are provided in Section 3. The conclusions of this report are in Section 4, with supporting references in Section 5.

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Westinghouse Non-Proprietary Class 3 1-3 52 Repair Weld Figure 1-1 General Schematic of the Embedded Flaw Repair to a Flaw in the Head Penetration Tube Inside Surface WCAP-18708-NP December 2021 Revision 0

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Westinghouse Non-Proprietary Class 3 1-4 Figure 1-2 General Schematic of the Embedded Flaw Repair to a Flaw in the Head Penetration Tube Outside Surface, or to a Flaw in the Attachment Weld (J-Groove Weld)

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Westinghouse Non-Proprietary Class 3 2-1 2 TECHNICAL BASIS FOR APPLICATION OF EMBEDDED FLAW REPAIR TECHNIQUE TO PENETRATION NOZZLES This section provides a discussion on the technical basis for the use of the embedded flaw repair method for a flawed head penetration nozzle (i.e., flaws on the inner diameter (ID) or outer diameter (OD) of the head penetration nozzles (Scenario 1 and Scenario 3)). Such a repair would involve depositing several layers of Alloy 52/52M weld material over the flaw detected on the inside surface of the penetration nozzle or right over the outside surface of the penetration nozzle of interest below the J-groove weld, as well as over the wetted surface of the J-groove weld in the event that an outside surface flaw is detected in the penetration nozzle. Since the Alloy 52/52M repair weld material is more PWSCC resistant than the existing Alloy 600 material, any detected surface flaws in the head penetration nozzles can then be shielded from the primary water environment and are no longer susceptible to primary water stress corrosion cracking.

This is consistent with the current plant operation experiences that no primary water stress corrosion cracking initiation has been observed in Alloy 52/52M weld material so far. The technical basis for the use of the embedded flaw repair method for the flawed head attachment weld (Scenario 2) is provided in Section 3.

[

]a,c,e The evaluation of the embedded flaw repair for the axial or circumferential crack on the penetration inner surface (Scenario 1) or outer surface (Scenario 3) began with the determination of an allowable end-of-evaluation period flaw size based on the acceptance criteria described in Section 2.1 for a flaw postulated to remain in the repaired penetration nozzle. [

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Westinghouse Non-Proprietary Class 3 2-2 2.1 ACCEPTANCE CRITERIA Rapid, non-ductile failure is possible for ferritic materials at low temperatures but is not applicable to the nickel-base alloy head penetration nozzle material, Alloy 600. Nickel-base alloy material is a high toughness material and plastic collapse would be the dominant mode of failure. [

]a,c,e 2.1.1 Axial Flaws For axial flaws the allowable flaw depth is given by [

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Westinghouse Non-Proprietary Class 3 2-3

]a,c,e WCAP-18708-NP December 2021 Revision 0

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Westinghouse Non-Proprietary Class 3 2-4 2.1.2 Circumferential Flaws For circumferential flaws [

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Westinghouse Non-Proprietary Class 3 2-5

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Westinghouse Non-Proprietary Class 3 2-6 2.2 METHODOLOGY The evaluation assumed that an unacceptable flaw has been detected on the surface of a penetration nozzle and that the embedded flaw repair process is used to seal the flaw from further exposure to the primary water environment. The evaluation began with the determination of an allowable end-of-evaluation period flaw size based on the acceptance criteria described in Section 2.1 for a flaw postulated to remain in the repaired penetration nozzle. [

]a,c,e The maximum initial flaw size in a penetration nozzle that can be repaired using the embedded flaw repair process can then be determined [

]a,c,e The following provides a discussion of the geometry, loading conditions, thermal transient stress analysis, and

[ ]a,c,e used in the development of the plant specific technical basis for the embedded flaw repair process.

2.2.1 Geometry and Material There are seventy eight CRDM head penetration nozzles in the reactor vessel upper closure head with the same nozzle geometry but at different locations in the closure head [5.a]. The outside radius and thickness for all Alloy 600 tubes are [ ]a,c,e The CRDM nozzle material is

[ ]a,c,e 2.2.2 Finite Element Analysis The distributions of transient thermal and pressure stresses [

]a,c,e Reference [6] considers the welding residual stresses associated with original nozzle installation.

Subsequent to the welding residual stress analysis, the stresses that result from the [

]a,c,e in the presence of welding residual conditions are calculated. [

]a,c,e including the welding residual stresses associated with original nozzle installation. [

]a,c,e Figure 2-2 shows the location of the stress cuts. [

]a,c,e of the circumferential and axial cracks postulated on the inside or outside of the nozzles.

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Westinghouse Non-Proprietary Class 3 2-7 a,c,e Figure 2-2 Finite Element Model with Analytical Stress Cuts Identified 2.2.3 Loading Conditions The requirement for determining the maximum allowable end-of-evaluation period flaw size using the rules of Section XI is that the governing loadings from the normal, upset (including test), emergency, and faulted conditions be considered. This is necessary because, as discussed in Section 2.1, different safety margins are used for the normal/upset conditions and the emergency/faulted conditions. A lower safety factor is used to reflect the lower probability of occurrence for the emergency/faulted conditions.

[

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[

]a,c,e The thermal transients that occur in the upper head region are relatively mild; Catawba Unit 2 is considered as a Tcold plant and the flow in the upper head region is low compared to other regions of the reactor vessel, which mutes the effects of the operating thermal transients.

The normal, upset (including test), emergency, and faulted transients considered for Catawba Unit 2 reactor vessel analyses, design specifications [8.a and 8.b], and the design cycles of the transients from Table 3-50 of Catawba final safety analysis report (FSAR, Table 3-50) [9] are summarized in Table 2-1. [

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Westinghouse Non-Proprietary Class 3 2-9 Table 2-1 Catawba Unit 2 Design Transients Transient Cycles(1)

Plant Loading at 5% of Full Power/min 13,200(3)

Step Load Decrease of 10% Full Power 2,000 Step Load Increase of 10% Full Power 2,000 Large Step Load Decrease 200 Plant Unloading at 5% of Full Power/min 13,200(3)

Reactor Trip from Full Power 400 Reactor Trip with Cooldown (CD) and Safety Injection (SI) 10 Inadvertent Auxiliary Spray 10 Inadvertent Depressurization 20 Turbine Roll Test 10 Primary Side Leak Test 50 Loss of Flow 80 Loss of Power 40 Loss of Load from Full Power 80 Large Steam Line Break 1 Small Steam line Break 5 Small LOCA 5 Steady State Fluctuations 1,000,000(2)

Plant Heatup 200 Plant Cooldown 200 Notes:

1. Cycles are from Equipment Specifications [8.a and 8.b] unless otherwise noted.

2. The 1,000,000 cycles considered for the infinite steady state fluctuation transient.

3. Transient cycles based on Table 3-50 of Catawba final safety analysis report (FSAR) [9]

2.2.4 Allowable Flaw Size Determination Allowable end-of-evaluation flaw sizes for axial and circumferential flaws with various aspect ratios (flaw length/flaw depth) in a CRDM penetration nozzle are calculated in accordance with the acceptance criteria discussed in Section 2.1. The allowable initial flaw sizes are subsequently determined by adjusting the allowable end-of-evaluation flaw sizes based on the results from the fatigue crack growth evaluation described in Section 2.2.6. Since the repaired flaws are embedded and sealed, they are not subjected to PWSCC.

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Westinghouse Non-Proprietary Class 3 2-10 2.2.5 Stress Intensity Factors One of the key elements in a crack growth analysis is the crack driving force or crack tip stress intensity factor, KI. This is based on the equations available in public literature. Both embedded and surface flaws are analyzed for repaired inside and outside surface flaws.

Outside and Inside Surface Flaws The stress intensity factors (SIF), KI, for the part through-wall surface cracks are calculated based on

[ ]a,c,e The stress distribution profile is represented by a 3rd order polynomial as shown below.

a a 2 a 3

= 0 + 1 + 2 + 3 t t t where:

0, 1, 2, and 3 are the stress profile curve fitting coefficients to be determined; a is the distance from the wall surface where the crack initiates; t is the wall thickness; and is the stress perpendicular to the plane of the crack.

The SIFs can be expressed in the general form as follows:

[

]a,c,e Embedded Flaws The stress intensity factor calculation for an embedded flaw was based on [

]a,c,e This stress intensity factor expression for subsurface (embedded) flaws can be expressed [

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Westinghouse Non-Proprietary Class 3 2-11

]a,c,e 2.2.6 Fatigue Crack Growth Prediction With the application of the embedded flaw repair process, any postulated flaws in the reactor vessel head penetration tubes are sealed from the PWR environment; therefore, the only mechanism for crack growth would be due to fatigue crack growth.

The FCG analysis procedure involves postulating an initial flaw at the region of concern and predicting the growth of that flaw due to an imposed series of loading transients. The applied loads include pressure, thermal transients, and residual stresses. The normal, upset (including test), emergency, and faulted thermal transients as well as the associated design cycles considered in the fatigue crack growth analysis are shown in Table 2-1. The cycles are distributed evenly over 60 years of plant design life. The stress intensity factor range, KI, that controls fatigue crack growth, depends on the geometry of the crack, its surrounding structure, and the range of applied stresses in the region of the postulated crack. Once KI is calculated, the fatigue crack growth due to a particular stress cycle can be determined using a crack growth rate reference curve applicable to the material of the head penetration nozzle. Once the incremental crack growth corresponding to a specific transient is calculated for a small time period, it is added to the original crack size, and the analysis continues to the next time period and/or thermal transient. The procedure is repeated in this manner until all the significant analytical thermal transients and cycles known to occur in a given period of operation have been analyzed.

[

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Westinghouse Non-Proprietary Class 3 2-12

]a,c,e 2.3 FRACTURE MECHANICS ANALYSIS RESULTS 2.3.1 Maximum End-of-Evaluation Period Flaw Sizes The maximum allowable end-of-evaluation period flaw sizes are determined for axial and circumferential surface flaws for postulated flaw aspect ratios (flaw length/flaw depth) of 2, 3, 6, and 10. The allowable flaw sizes are considered for all normal, upset, test, emergency, and faulted conditions and the most limiting allowable flaw sizes from these conditions are summarized in Table 2-2 and will be used in the generation of flaw evaluation charts.

Table 2-2 Maximum Allowable End-of-Evaluation Period Flaw Size Based on Section XI Axial Circumferential Location Aspect Ratio Allowable Flaw Size Allowable Flaw Size (l/a) a/t a (in.) a/t a (in.)

2 0.75 0.469 0.75 0.469 CRDM Nozzle 3 0.75 0.469 0.75 0.469

[ ]a,c,e 6 0.75 0.469 0.58 0.363 10 0.75 0.469 0.46 0.288 Notes: l = flaw length a = flaw depth t = wall thickness 2.3.2 Allowable Initial Flaw Sizes for Penetration Nozzles After the maximum allowable end-of-evaluation period flaw sizes are determined, [

]a,c,e First, the outside and inside surface flaws with aspect ratios of 2, 3, 6, and 10 are postulated. [

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Westinghouse Non-Proprietary Class 3 2-13

]a,c,e The results are also plotted in Figure 2-3 and Figure 2-4.

Table 2-3 Maximum Allowable Initial Flaw Size on CRDM Nozzle for Repair Inside Surface Outside Surface Years Aspect Circumferential Axial Circumferential Axial Location of Ratio Flaw Flaw Flaw Flaw Operation (l/a) a/t a (in.) a/t a (in.) a/t a (in.) a/t a (in.)

2 0.74 0.4625 0.74 0.4625 0.74 0.4625 0.74 0.4625 3 0.74 0.4625 0.72 0.4500 0.73 0.4563 0.68 0.4250 20 6 0.57 0.3563 0.64 0.4000 0.54 0.3375 0.52 0.3250 10 0.45 0.2813 0.52 0.3250 0.42 0.2625 0.41 0.2563 2 0.74 0.4625 0.74 0.4625 0.74 0.4625 0.73 0.4563 Downhill 3 0.74 0.4625 0.69 0.4313 0.72 0.4500 0.66 0.4125 40 Side 6 0.57 0.3563 0.57 0.3563 0.51 0.3188 0.47 0.2938 10 0.45 0.2813 0.47 0.2938 0.39 0.2438 0.40 0.2500 2 0.74 0.4625 0.73 0.4563 0.74 0.4625 0.72 0.4500 3 0.74 0.4625 0.68 0.4250 0.71 0.4438 0.63 0.3938 60 6 0.57 0.3563 0.54 0.3375 0.47 0.2938 0.44 0.2750 10 0.45 0.2813 0.44 0.2750 0.37 0.2313 0.37 0.2313 2 0.74 0.4625 0.73 0.4563 0.74 0.4625 0.72 0.4500 20 3 0.74 0.4625 0.68 0.4250 0.73 0.4563 0.66 0.4125 6 0.57 0.3563 0.53 0.3313 0.53 0.3313 0.46 0.2875 10 0.45 0.2813 0.43 0.2688 0.43 0.2688 0.39 0.2438 2 0.74 0.4625 0.71 0.4438 0.74 0.4625 0.70 0.4375 Uphill 40 3 0.74 0.4625 0.63 0.3938 0.72 0.4500 0.61 0.3813 Side 6 0.57 0.3563 0.46 0.2875 0.52 0.3250 0.42 0.2625 10 0.45 0.2813 0.36 0.2250 0.40 0.2500 0.35 0.2188 2 0.74 0.4625 0.7 0.4375 0.74 0.4625 0.68 0.4250 3 0.74 0.4625 0.59 0.3688 0.71 0.4438 0.57 0.3563 60 6 0.57 0.3563 0.41 0.2563 0.50 0.3125 0.38 0.2375 10 0.45 0.2813 0.32 0.2000 0.38 0.2375 0.32 0.2000 Notes: l = flaw length a = flaw depth t = wall thickness WCAP-18708-NP December 2021 Revision 0

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Westinghouse Non-Proprietary Class 3 2-14 Figure 2-3 Maximum Allowable Initial Flaw Size on CRDM Nozzle for Repair - Downhill Side WCAP-18708-NP December 2021 Revision 0

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Westinghouse Non-Proprietary Class 3 2-15 Figure 2-4 Maximum Allowable Initial Flaw Size on CRDM Nozzle for Repair - Uphill Side WCAP-18708-NP December 2021 Revision 0

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Westinghouse Non-Proprietary Class 3 3-1 3 TECHNICAL BASIS FOR APPLICATION OF EMBEDDED FLAW REPAIR TECHNIQUE TO ATTACHMENT J-GROOVE WELD This section provides a discussion on the technical basis for the use of the embedded flaw repair method for the flawed head attachment weld (Scenario 2). [

]a,c,e A flaw evaluation was carried out by analyzing a planar flaw in the reactor vessel head the size of the J-groove weld size.

3.1 ACCEPTANCE CRITERIA 3.1.1 Section XI Appendix K The evaluation procedure and acceptance criteria used to demonstrate structural integrity of the reactor vessel closure head is contained in Appendix K of ASME Section XI Code [2] as well as Regulatory Guide 1.161 [12]. Although the original purpose of Appendix K was to evaluate reactor vessels with low upper shelf fracture toughness, the general approach in paragraph K-4220 is equally applicable to any region of the reactor vessel where the fracture toughness can be described with elastic plastic parameters. This approach to evaluate the integrity of a nuclear vessel has been developed over several years and has been illustrated with a number of example problems [13] to demonstrate its use. The extension of this methodology to issues other than the low shelf fracture toughness issue is appropriate when service conditions (temperature) promote ductile behavior. The closure head region of the reactor vessel has the operating temperature of about 557 ºF. This would result in ductile behavior and therefore the use of elastic-plastic fracture mechanics method is appropriate.

The acceptance criteria are to be satisfied for each category of transients, namely, Service Load Level A (normal), Level B (upset, including test), Level C (emergency) and Level D (faulted) conditions and two criteria discussed below must be satisfied.

The first criterion is that the crack driving force must be shown to be less than the material toughness as follows:

Japplied < Jmaterial where Japplied is the J-integral value calculated for the postulated flaw under the applicable Service Level condition and Jmaterial is the J-integral characteristic of the material resistance to ductile tearing at a crack extension of 0.1 inch. For Level A and B conditions, a safety factor of 1.15 is conservatively applied to the Japplied per Reg Guide 1.161 [12] and ASME Section XI Appendix K Article K-4220 of ASME Section XI Code [2]. The factor of 1.15 needs only to be applied on pressure, however, in this evaluation it is applied WCAP-18708-NP December 2021 Revision 0

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Westinghouse Non-Proprietary Class 3 3-2 to the J-integral calculated from the transient and residual stresses in addition to the normal operating pressure. For Level C and D conditions, the safety factor on Japplied is 1.0.

The second criterion is that the flaw must also be stable under ductile crack growth as follows [12, Section 1.1.2]:

J applied dJ material a da at Japplied = Jmaterial where, Jmaterial = J-integral resistance to ductile tearing for the material.

J applied

= Partial derivative of the applied J-integral with respect to flaw depth, a a

dJ material

= Slope of the J-R curve da For Level A and B conditions, a safety factor of 1.25 is conservatively applied to the Japplied per Reg Guide 1.161 [12] and ASME Section XI Appendix K Article K-4220 of ASME Section XI Code [2]. The factor of 1.25 needs only to be applied on pressure, however, in this evaluation it is conservatively applied to the transient and residual stresses in addition to the normal operating pressure. For Level C and D conditions, the safety factor on Japplied is 1.0. Flaw stability is verified when the slope of the applied J-integral curve is less than the material J-integral curve at the point on J-R curve where the two curves intersect.

3.1.2 Primary Stress Limits In addition to satisfying the Section XI criteria, the primary stress limits of paragraph NB-3000 in Section III of the ASME Code [14] must be satisfied. The effects of a local area reduction that is equivalent to the area of the postulated flaw in the vessel head attachment weld must be considered by increasing the membrane stresses to reflect the reduced cross section. The allowable flaw depth was determined by evaluating the primary stress of the spherical head with reduced wall thickness using the maximum pressure of [ ]a,c,e for all service conditions. The results show the allowable flaw depth is 2.472 inches.

3.2 METHODOLOGY Since the depth of a flaw in the attachment weld cannot be detected using current technology, the engineering evaluation for the embedded flaw repair process was performed to demonstrate the stability of an assumed hypothetical flaw that encompasses the entire attachment J-groove weld region in the reactor vessel head near the penetration nozzle. The criteria used to demonstrate the stability and structural integrity of the reactor vessel closure head is described in Section 3.1.1 as per the ASME Code [2] and Regulatory Guide 1.161 [12].

After the implementation of the embedded flaw repair process, [

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Westinghouse Non-Proprietary Class 3 3-3

]a,c,e That is, the flaw depth at the end of evaluation period should be below the 2.472 inches as determined in Section 3.1.2 such that primary stress limit of the ASME Code Section III, paragraph NB-3000 [14] is satisfied. In addition, it needs to be shown that the postulated flaw will not grow through the repair layer.

3.2.1 Geometry and Material The reactor vessel head is made of ]a,c,e with the following geometry:

[

]a,c,e The reactor vessel upper head nozzle attachment weld geometry for the nozzles used in this calculation is tabulated in Table 3-1 for the case without the weld fillet as shown in Figure 3-1. The weld dimensions in Table 3-1 are used for the fatigue crack growth and J-integral analyses for postulated flaws in the reactor vessel head. The height and width of the J-groove weld configurations are based on the design dimensions where the weld depths for all penetration nozzles on the uphill and downhill sides are provided based on drawing [5.a]. The weld depth dimension a, as shown in Figure 3-1 and Table 3-1 are the weld depths without the fillet weld. For the embedded flaw FCG through the repair layer, a nominal fillet is considered in the analyses that is representative for the Catawba Unit 2 geometry based on review of other similar head geometries. The weld dimensions in Table 3-1 are used in the fatigue crack growth analysis for the growth of postulated flaws through the weld repair layer. [

]a,c,e these flaw depths with the addition of a nominal fillet bound all penetration row weld depths a for Catawba Unit 2.

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Westinghouse Non-Proprietary Class 3 3-4 Table 3-1 J-Groove Weld Geometries-Without Weld Fillet a,c,e Figure 3-1 Definition of J-Groove Weld Dimensions WCAP-18708-NP December 2021 Revision 0

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Westinghouse Non-Proprietary Class 3 3-5 3.2.2 Loading Conditions For the normal/upset condition, the reactor vessel closure head structural integrity evaluation is performed for all the transients in Table 2-1, and [

]a,c,e For the emergency and faulted condition evaluation, [

]a,c,e There are many head penetrations in the reactor vessel upper head, and [

]a,c,e The distribution of residual, transient thermal, and pressure stresses in the closure head region is obtained from detailed three-dimensional elastic-plastic finite element analyses of the head penetration nozzle region [6]. [

]a,c,e 3.2.3 Stress Intensity Factors J-Groove Weld Double Corner Crack in the Reactor Vessel Head Since the depth of a flaw in the attachment weld cannot be detected using current technology, it is conservatively assumed that the flaw in the attachment weld extends radially over the entire attachment weld. [

]a,c,e The stress intensity factor expression shown above is applicable for a range of flaw shapes, with the depth of the flaw defined as a, and the width of the flaw defined as c, as shown in Figure 3-2. This flexibility is necessary because this expression can be applied to different attachment J-groove weld shapes for Catawba Unit 2 closure head penetrations as shown in Table 3-1. The attachment J-groove weld shapes were based on the J-groove geometry shown in the head penetration nozzle drawings and models for Catawba Unit 2 [5.a]. [

]a,c,e WCAP-18708-NP December 2021 Revision 0

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Westinghouse Non-Proprietary Class 3 3-6 Figure 3-2 Corner Crack Geometry Embedded Flaw in the Reactor Vessel Head

[

]a,c,e The details of the method is discussed in Section 2.2.5.

3.2.4 J-R curve for Reactor Vessel Closure Head Material One of the most important pieces of information for fracture toughness for pressure vessel and piping materials is the J-R curve of the material. The J-R stands for material resistance to crack extension, as represented by the measured J-integral value versus crack extension. Simply put, the J-R curve to cracking resistance is as significant as the stress-strain curve to the load-carrying capacity and the ductility of a material. Both the J-R curve and stress-strain curves are properties of a material.

[

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Westinghouse Non-Proprietary Class 3 3-7

]a,c,e Neutron irradiation has been shown to produce embrittlement that reduces the toughness properties of the reactor vessel ferritic steel material. The irradiation levels are very low in the reactor vessel closure head region and therefore the fracture toughness will not be measurably affected.

3.2.5 Applied J-Integral For small scale yielding, Japplied of a crack can be calculated by the Linear Elastic Fracture Mechanics (LEFM) method based on the crack tip stress intensity factor, KI, calculated as per Section 3.2.3. However, a plastic zone correction must be performed to account for the plastic deformation at the crack tip similar to the approach in Regulatory Guide 1.161 [12]. The plastic deformation ahead of the crack front is then regarded as a failed zone and the crack size is, in effect, increased. The KI-values can be converted to Japplied by the following equation:

2

=

where Kep is the plastic zone corrected K-value, and E=E/(1-2) for plane strain, E = Youngs Modulus, and = Poissons Ratio.

Kep is equal to the elastically calculated KI-value based on the plastic zone adjusted crack depth or size.

The plastic zone size, rp, is calculated by 2

1

=

6 where Sy is the yield strength of the material.

Assume that the crack depth is ao, the Kep can now be calculated based on a new crack length, ao + rp. For small scale yielding, this can be simplified as Kep = f KI

( + )

where = 0 Once the J-applied is calculated, stability for the postulated flaw in the attachment J-groove weld can be determined using the methodology described in Section 3.1.1.

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Westinghouse Non-Proprietary Class 3 3-8 3.2.6 Fatigue Crack Growth Prediction With the application of the embedded flaw repair process, any postulated flaws in the reactor vessel head penetration tubes or the attachment weld are sealed from the PWR environment; therefore, the only mechanism for crack growth would be due to fatigue.

The FCG analysis procedure involves postulating an initial flaw at the region of concern and predicting the growth of that flaw due to an imposed series of loading transients, using the same approach described in Section 2.2.6. The FCG curves used for [

]a,c,e and the embedded flaw beneath the repair weld are discussed below.

FCG Curve for the Reactor Vessel Closure Head: Carbon and Low Alloy Ferritic Steel The crack growth rate curves used in the analyses for [

]a,c,e are taken directly from Appendix A in the ASME Section XI Code [2] for ferritic steel material. With the repair weld any potential flaws in the J-groove weld (Alloy 182) are sealed from the primary water environment and the only applicable growth mechanism is fatigue crack growth in air environment; therefore, the analysis is performed for a surface flaw based on the limiting crack growth rate reference curve of the air environment. This curve is a function of the applied stress intensity factor range (KI) and the R ratio, which is the ratio of the minimum to maximum stress intensity factor during a thermal transient. The crack growth equation is given below:

= 0 ( )

where n is the slope of the log (da/dN) versus log (KI) curve and is equal to 3.07 for subsurface flaws.

Parameter Co is a scaling constant:

0 = 0 <

= 1.99 x 1010 where Kth is the threshold KI value below which the fatigue crack growth rate is negligible and S is a scaling parameter. Both Kth and S are a function of the R ratio (Kmin/Kmax). The calculation of crack tip stress intensity factor range (KI) also changes with R ratio when .

= 5.0 < 0

= 5.0(1 0.8) 0 < 1.0 The calculation of crack tip stress intensity factor range (KI) also changes with R ratio when .

The calculation of S and for different R ratio ranges is summarized below:

• For 0 R 1 S = 25.72(2.88-R)-3.07 and KI = Kmax - Kmin

• For R < 0 and > 1.12 S=1 and KI = Kmax - Kmin

• For -2 R 0 and 1.12 S=1 and KI = Kmax

• For R<-2 and 1.12 S=1 and KI = (1-R)Kmax/3 WCAP-18708-NP December 2021 Revision 0

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Westinghouse Non-Proprietary Class 3 3-9

[

]a,c,e Note that a condition is imposed on A-4300(b)(1) of ASME Code Section XI in 10CFR 50.55a Codes and Standards and a factor of 0.8 is applied to the limit in defined in A-4300 of the ASME Code Section XI [2].

FCG Curve for the Repair Weld, Alloy 52/52M, Below the J-Groove Attachment Weld

[

]a,c,e 3.3 FRACTURE MECHANICS ANALYSIS RESULTS 3.3.1 Results for Applied J-Integral and J-R Curve For the J-integral calculation, the key aspects of the analysis are to demonstrate that the magnitude of J-applied is less than J-material at 0.1 inch crack extension, and the slope of the J-material curve is greater than the slope of the J-applied curve at the intersection of the Jmat and Japplied curves. This evaluation is performed for the postulated flaws encompassing the J-groove welds at all the nozzle locations. The weld dimensions are shown in Table 3-1. The results shows that for all the nozzle locations, the applied J-integral is less than material J-integral at 0.1 inch crack extension, as shown in Table 3-2 and Table 3-3. The slope of the J-material curve is also greater than the slope of the J-applied curve at the intersection of the J-applied and J-material curves for all the locations. Figures 3-3 and 3-4 show the plots for the penetration nozzle locations with the highest J-applied at 0.1 inch crack extension for Level A/B and Level C/D conditions, respectively.

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Westinghouse Non-Proprietary Class 3 3-10 Table 3-2 J-Integral Results for 0.1 inch Crack Extension on Downhill and Uphill Sides -

Level A/B Downhill Uphill Penetration Pen. No. Japplied Jmaterial Japplied Jmaterial Angle (kip-in/in2) (kip-in/in2) (kip-in/in2) (kip-in/in2) 1 0.0 0.418 1.415 0.723 1.415 2-5 11.4 0.445 1.415 0.675 1.415 6-9 16.2 0.456 1.415 0.651 1.415 10-13 18.2 0.462 1.415 0.641 1.415 14-17 23.3 0.485 1.415 0.612 1.415 18-21 24.8 0.491 1.415 0.612 1.415 22-29 26.2 0.496 1.415 0.604 1.415 30-37 30.2 0.513 1.415 0.580 1.415 38-41 33.9 0.536 1.415 0.567 1.415 42-49 35.1 0.542 1.415 0.560 1.415 50-53 36.3 0.548 1.415 0.554 1.415 54-61 38.6 0.561 1.415 0.542 1.415 62-65 44.3 0.596 1.415 0.516 1.415 66-73 45.4 0.610 1.415 0.511 1.415 74-78 48.7 0.636 1.415 0.493 1.415 WCAP-18708-NP December 2021 Revision 0

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Westinghouse Non-Proprietary Class 3 3-11 Table 3-3 J-Integral Results for 0.1 inch Crack Extension on Downhill and Uphill Sides -

Level C/D Downhill Uphill Penetration Pen. No. Japplied Jmaterial Japplied Jmaterial Angle (kip-in/in2) (kip-in/in2) (kip-in/in2) (kip-in/in2) 1 0.0 0.497 1.471 1.114 1.471 2-5 11.4 0.530 1.471 0.959 1.471 6-9 16.2 0.545 1.471 0.912 1.471 10-13 18.2 0.551 1.471 0.891 1.471 14-17 23.3 0.581 1.471 0.836 1.471 18-21 24.8 0.587 1.471 0.839 1.471 22-29 26.2 0.594 1.471 0.823 1.471 30-37 30.2 0.614 1.471 0.780 1.471 38-41 33.9 0.645 1.471 0.759 1.471 42-49 35.1 0.652 1.471 0.748 1.471 50-53 36.3 0.660 1.471 0.737 1.471 54-61 38.6 0.675 1.471 0.718 1.471 62-65 44.3 0.720 1.471 0.679 1.471 66-73 45.4 0.737 1.471 0.672 1.471 74-78 48.7 0.770 1.471 0.646 1.471 WCAP-18708-NP December 2021 Revision 0

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Westinghouse Non-Proprietary Class 3 3-12

(

Figure 3-3 Applied and Material J-Integral versus Crack Depth Curve for the Downhill Case with the Highest Japplied at 0.1 inch Crack Extension - Level A/B Conditions WCAP-18708-NP December 2021 Revision 0

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Westinghouse Non-Proprietary Class 3 3-13

(

Figure 3-4 Applied and Material J-Integral versus Crack Depth Curve for the Downhill Case with the Highest Japplied at 0.1 inch Crack Extension - Level C/D Conditions WCAP-18708-NP December 2021 Revision 0

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Westinghouse Non-Proprietary Class 3 3-14 3.3.2 Results for Fatigue Crack Growth into the Reactor Vessel Head The FCG into the reactor vessel head is considered for the postulated cracks with the initial flaw size based on the J-groove weld depth from Table 3-1.

The fatigue crack growth is performed for the outermost nozzle along with the largest J-groove weld dimensions, in order to bound all the other penetration nozzles. It is assumed that the initial aspect ratio is held constant as the flaw grows through the reactor head wall thickness.

The stress intensity factor is conservatively calculated based on [

]a,c,e and the fatigue crack growth law for the reactor vessel head carbon steel material described in Section 3.2.6 is used. The FCG results are shown in Figure 3-5, which demonstrates that the postulated flaw will not reach the reactor vessel head primary stress limit (2.472 inches) after 60 years of growth.

3 Primary Stress Limit of 2.472 inches 2.5 2

J-groove Flaw Depth (inch)

Uphill Side 1.5 1

Downhill Side 0.5 0

0 10 20 30 40 50 60 Time (Years)

Figure 3-5 Fatigue Crack Growth Prediction into the Reactor Vessel Shell for Postulated Flaws in the J-Groove Welds for the Bounding Penetration Angles on the Downhill and Uphill Sides WCAP-18708-NP December 2021 Revision 0

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Westinghouse Non-Proprietary Class 3 3-15 3.3.3 Results for Fatigue Crack Growth into the Repair Weld The attachment weld (J-groove) repair is performed by [

]a,c,e The attachment weld is thus sealed, and the thickness of the reactor vessel shell is locally increased by [ ]a,c,e In order to determine the durability of the repair weld, an embedded flaw based on the J-Groove weld geometry is postulated, which starts from [ ]a,c,e beneath the free surface. The postulated flaw, which encompasses the entire shape of the J-groove weld, will have an aspect ratio (flaw length/flaw depth) of 2.

This aspect ratio of 2 bounds all the aspect ratios for the uphill and downhill side attachment weld dimensions shown in Table 3-1. For the FCG analysis, the initial total flaw depth (2a) is assumed equal to the maximum uphill and downhill weld depths [

]a,c,e in Table 3-1. The crack growth results are summarized in Table 3-4 and it shows that the structural integrity of the repaired weld layer is expected to be maintained for at least 47 years of service life.

Table 3-4 Growth of Embedded Flaw in J-Groove Weld Remaining Repair Weld Thickness Location Year (inch) 0 [ ] a,c,e 10 [ ] a,c,e Uphill 20 [ ] a,c,e Side 30 [ ] a,c,e 40 [ ] a,c,e 47 [ ]a,c,e 0 [ ] a,c,e 10 [ ] a,c,e 20 [ ] a,c,e Downhill 30 [ ] a,c,e Side 40 [ ] a,c,e 50 [ ] a,c,e 60 [ ] a,c,e WCAP-18708-NP December 2021 Revision 0

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Westinghouse Non-Proprietary Class 3 4-1 4

SUMMARY

AND CONCLUSIONS Engineering evaluations were performed to provide plant specific technical basis for the Westinghouse embedded flaw repair process that is associated with the reactor vessel head penetration nozzle inspection and contingency repair program for Catawba Unit 2.

The technical basis for the use of the embedded flaw repair process if unacceptable flaws are detected in the head penetration nozzles is provided in Section 2. Based on the results in Section 2.3, it is determined that unacceptable axial and circumferential flaws detected on the inside surface or outside surface of a head penetration nozzle can be repaired using the embedded flaw repair process by shielding them from the primary water environment. The maximum allowable initial axial and circumferential flaw sizes that can be repaired using the Westinghouse embedded flaw repair process are shown in Table 2-3 and Figures 2-3 and 2-4 for a plant service life up to 60 years.

The technical basis for the use of the embedded flaw repair process if indications or flaws are found in the head penetration attachment J-groove welds is provided in Section 3. Based on the results shown in Section 3.3, the evaluation documented herein has demonstrated that the embedded flaw repair process is a viable method for repairing flaws found in the attachment J-groove weld. The fracture mechanics evaluation demonstrated that a flaw postulated in the J-groove weld which encompasses the entire attachment J-groove weld shape is stable under the J-integral analysis. Furthermore, the reduced wall thickness considering the 60-year fatigue crack growth of the postulated flaw will meet the reactor vessel head primary stress limit minimum thickness requirement. The fatigue crack growth through the weld overlay repair layer demonstrates that a postulated flaw in the J-groove weld will not grow through the repair layer in less than 47 years. Therefore, it is technically justified to use the embedded flaw repair process as the repair option for the reactor vessel head penetration nozzle attachment J-groove welds since the criteria for application of such a process as stated in Appendix C of WCAP-15987-P Revision 2-P-A [4] is met.

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Westinghouse Non-Proprietary Class 3 5-1 5 REFERENCES

1. Duke Energy, RA-21-0145, Revision to Proposed Alternative to Use Reactor Vessel Head Penetration Embedded Flaw Repair Method, April 24, 2021 (ML21114A000).

2. ASME Boiler & Pressure Vessel Code, 2007 Edition with 2008 Addenda,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components.

3. Westinghouse Letter, LTR-SDA-19-017, Revision 0, Catawba Unit 2 General Reactor Vessel Control Rod Drive Mechanism (CRDM) Penetration J-Groove Weld Repair Dimensional Requirements, March 14, 2019.

4. Westinghouse Report WCAP-15987-P, Revision 2-P-A, Technical Basis for the Embedded Flaw Process for Repair of Reactor Vessel Head Penetrations, December 2003.

5. [

]a,c,e

6. [

]a,c,e

7. [

]a,c,e

8. [

]a,c,e

9. Duke Energy Company, Catawba Nuclear Station Updated Final Safety Analysis Report, Revision 20, April 2018.

10. [

]a,c,e WCAP-18708-NP December 2021 Revision 0

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Westinghouse Non-Proprietary Class 3 5-2

11. [

]a,c,e

12. Regulatory Guide 1.161, Evaluation of Reactor Pressure Vessels with Charpy Upper-Shelf Energy Less Than 50 ft-lb.

13. Development of Criteria for Assessment of Reactor Vessels with Low Upper Shelf Fracture Toughness, Welding Research Council Bulletin 413, July 1996.

14. ASME Boiler & Pressure Vessel Code, 1971 Edition with Addenda through the Winter of 1972,Section III, Rules for Construction of Nuclear Power Plant Component.

15. [

]a,c,e

16. [

]a,c,e WCAP-18708-NP December 2021 Revision 0

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RA-21-0144 Attachment 3 Affidavit Attesting to Proprietary Nature of Information in Attachment 1

Westinghouse Non-Proprietary Class 3 CAW-21-5247 Page 1 of 3 COMMONWEALTH OF PENNSYLVANIA:

COUNTY OF BUTLER:

(1) I, Anthony J. Schoedel, have been specifically delegated and authorized to apply for withholding and execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse).

(2) I am requesting the proprietary portions of WCAP-18708-P, Revision 0 be withheld from public disclosure under 10 CFR 2.390.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged, or as confidential commercial or financial information.

(4) Pursuant to 10 CFR 2.390, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse and is not customarily disclosed to the public.

(ii) The information sought to be withheld is being transmitted to the Commission in confidence and, to Westinghouses knowledge, is not available in public sources.

(iii) Westinghouse notes that a showing of substantial harm is no longer an applicable criterion for analyzing whether a document should be withheld from public disclosure. Nevertheless, public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable

Westinghouse Non-Proprietary Class 3 CAW-21-5247 Page 2 of 3 others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

(5) Westinghouse has policies in place to identify proprietary information. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage (e.g., by optimization or improved marketability).

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

Westinghouse Non-Proprietary Class3 CAW-21-5247 Page3 of3 (6) The attached documents are bracketed and marked to indicate the bases for withholding. The justification for withholding is indicated in both versions by means of lower-case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower-case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (5Xa) through (f) of this Affidavit.

I declare that the avennents of fact set forth in this Affidavit are true and correct to the best of my knowledge, infonnation, and belief.

I declare under penalty of perjury that the foregoing is true and correct.

I)/I (I/).o:)... I Executed on:

-li./(

Anthony J. Schoedel, Manager Advanced Reactors Licensing Engineering