RS-24-088, Relief Requests Associated with the Fifth Inservice Inspection Interval

Relief Requests Associated with the Fifth Inservice Inspection Interval

ML24257A063
Person / Time
Site:	Byron
Issue date:	09/13/2024
From:	Steinman R Constellation Energy Generation
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
RS-24-088
Download: ML24257A063 (1)

Text

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office

RS 088 10 CFR 50.55a(z)

September 13, 2024

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

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

Subject:

Relief Requests Associated with the Fifth Inservice Inspection Interval

Reference:

1) Letter from R. A. Gibbs, (U.S. NRC) to C. M. Crane, (EGC), Byron Station, Unit Nos. 1 and 2 - Evaluation of Relief Request I3R -06 for Control Rod Drive Canopy Seal Welds (TAC No. MD3863 and MD3864), dated March 9, 2007 (ML070520480)

2) Letter from Jacob Zimmerman, (U.S. NRC) to M. J. Pacilio (EGC),

Braidwood Station, Units 1 and 2 and Byron Station, Unit Nos. 1 and 2 -

Relief Requests I3R-09 and I3R-20 Regarding Alternative Requirements for Repair of Reactor Vessel Head Penetrations (TAC No. ME6071, ME6072, ME6073, and ME6074), dated March 29, 2012 (ML120790647)

3) Letter from David M. Gullott, (EGC) to U.S. NRC, Revision to the Third 10-Year Inservice Inspection Interval Requests for Relief for Alternative Requirements for the Repair of Reactor Vessel Head Penetrations, dated September 8, 2014 (ML14251A536)

4) Letter from Justin C. Poole, (U.S. NRC) to Bryan C. Hanson (EGC), Byron Station, Units Nos. 1 and 2, and Braidwood Station, Units 1 and 2 - Relief from the Requirements of the ASME Code (CAC Nos. MF4809, MF4810, MF4811, and MF4812), dated January 21, 2016 (ML16007A185)

5) Letter from David M. Gullott, (EGC) to U.S. NRC, Relief Request for Alternative Requirements for the Repair and Examination of Reactor Vessel Head Penetrations for the Fourth Inservice Inspection Interval, dated August 16, 2016 (ML16229A250)

6) Letter from David M. Gullott, (EGC) to U.S. NRC, Supplemental Response to Request for Additional Information for Byron Station Relief Request I4R-10:

Proposed Alternative Requirements for the Repair and Examination of Reactor Vessel Head Penetrations for the Fourth Inservice Inspection Interval, dated February 13, 2017 (ML17044A294)

September 13, 2024 U. S. Nuclear Regulatory Commission Page 2

7) Letter from Kimberly J. Green, (U.S. NRC) to Bryan C. Hanson (EGC), Byron Station, Units Nos. 1 and 2 - Request for Relief from the Requirements of the ASME Code (CAC Nos. MF8282 and MF8283), dated March 6, 2017 (ML17062A428)

8) Letter from N. L. Salgado, (U.S. NRC) to D. P. Rhoades (EGC), Braidwood Station, Units 1 and 2; Byron Station, Unit Nos. 1 and 2; Calvert Cliffs Nuclear Power Plant, Units 1 and 2; Clinton Power Station, Unit No. 1; Dresden Nuclear Power Station, Units 2 and 3; James A. FitzPatrick Nuclear Power Plant; LaSalle County Station, Units 1 and 2; Limerick Generating Station, Units 1 and 2; Nine Mile Point Nuclear Station, Units 1and 2; Peach Bottom Atomic Power Station, Units 2 and 3; Quad Cities Nuclear Power Station, Units 1 and 2; and R. E. Ginna Nuclear Power Plant - Proposed Alternative to Eliminate Certain Documentation Requirements for Pressure Retaining Bolting (EPIDS L-2021-LLR-0153, -0154, and - 0155), dated August 5, 2021 (ML21216A220)

9) Letter from N.L. Salgado (U.S. NRC) to B.C. Hanson (Exelon Generation Company, LLC), "Byron Station, Unit Nos. 1 and 2; Dresden Nuclear Power Station, Units 2 and 3; James A. Fitzpatrick Nuclear Power Plant; LaSalle County Station, Units 1 and 2; Limerick Generating Station, Units 1 and 2; and Quad Cities Nuclear Power Station, Units 1 and 2 - Request to Use Provisions in the 2013 Edition of the ASME Boiler and Pressure Vessel Code For Performing Non-Destructive Examinations (EPID L-2019-LLR-0080),"

dated April 17, 2020 (ML20099D955)

In accordance with 10 CFR 50.55a, " Codes and standards," paragraphs (z)(1) and (z)(2 ),

Constellation Energy Generation, LLC ( CEG), hereby requests NRC approval of the attached relief requests associated with the Fifth Inservice Inspection (ISI) Interval for Byron Station, Units 1 and 2 (Byron). The F ifth Interval of the Byron ISI Program is currently scheduled to begin on July 16, 2025, and end on July 15, 2037, subject to the allowable changes for inspection intervals in IWA-2430, "Inspection Intervals," and will comply with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI, 2019 Edition. CEG proposes the following relief requests for the Byron F ifth ISI Interval:

I5R-01: Alternative Requirements for Repair/Replacement of Control Rod Drive Mechanism (CRDM) Canopy Seal Welds in Accordance with IWA-4000. There currently are no known degraded canopy seal welds at Byron Station.

I5R-02: Proposed Alternative Requirements for the Repair and Examination of Reactor Vessel Head Penetrations

In Reference 2, the U.S. Nuclear Regulatory Commission (NRC) provided their authorization to implement Relief Requests I3R-09 and I3R-20, Revision 1 as a repair method for degradation identified in reactor vessel head penetrations. In Reference 3, CEG, then known as Exelon Generation Company (EGC), LLC, submitted a relief request that was applicable to the Third ISI Interval and requested inspection frequency relief for the reactor vessel head penetrations repair weld surface examinations (i.e., dye penetrant (PT)) for Braidwood Station, Units 1 and 2 September 13, 2024 U. S. Nuclear Regulatory Commission Page 3

and Byron Station, Units 1 and 2. In Reference 4, the NRC approved the request for the Third ISI Interval for Braidwood Station and Byron Station.

In Reference 5, CEG submitted Relief Request I4R-10, Revision 0 for the Fourth ISI Interval for Byron Station, Units 1 and 2, similar to the relief request that was approved in Reference 4. In Reference 6, CEG submitted Revision 2 of Relief Request I4R-10 based on requests for additional information from the NRC. In Reference 7, the NRC approved the request for the Fourth ISI Interval for Byron Station.

The Attachment 1 relief request is similar to the Reference 6 relief request that was approved in Reference 7. The only differences between the Reference 6 request and the Attachment 1 request are: minor formatting and editorial changes are made and the Applicable Code Edition is updated to reflect the ASME BPV Code,Section XI, 2019 Edition, which is the ASME Code applicable to the Byron Station Fifth ISI Interval.

Differences in the ASME Code have been reconciled and the change to the more recent ASME Code had no impact on the technical basis for the proposed request. The technical information supporting the Reference 2 and Reference 7 relief request approvals remain applicable to support the Attachment 1 request for the Fifth ISI Interval.

I5R-03: Proposed Alternative Concerning ASME Section XI Repair/Replacement Documentation for Replacement of Pressure Retaining Bolting The proposed alternative concerns completion of a Repair/Replacement Plan and Form NIS-2, Owners Repair/Replacement Certification Record. Specifically, CEG proposes to forego preparation and completion of a Repair/Replacement Plan and Form NIS-2 for pressure retaining bolting that is not classified as Examination Category B-G-1, B-G-2, C-D, or E-G.

I5R-04: Request to Permit Continued Application of Certain ASME Section XI 2013 Edition Non-Destructive Examination Requirements

The Fifth Interval of the Byron, Units 1 and 2, ISI Program will comply with the 2019 Edition of the ASME BPV Code. However, this request proposes short -term application of certain 2013 Edition non-destructive examination requirements to maintain Byron requirements consistent with the remainder of the CEG fleet.

The bases for these relief requests are provided in Attachments 1 through 4, respectively.

CEG requests approval of these requests by July 15, 2025, to support implementation of the Byron Station Fifth ISI Interval.

September 13, 2024 U. S. Nuclear Regulatory Commission Page 4

There are no regulatory commitments contained within this letter.

Should you have any questions concerning this letter, please contact Ms. Lisa M. Zurawski at (779) 231-6196.

Respectfully,

Rebecca L. Steinman Sr. Manager - Licensing Constellation Energy Generation, LLC

Attachments:

1. 10 CFR 50.55a Relief Request I5R-01, Revision 0

2. 10 CFR 50.55a Relief Request I5R-02, Revision 0

3. 10 CFR 50.55a Relief Request I5R-03, Revision 0

4. 10 CFR 50.55a Relief Request I5R-04, Revision 0

cc: NRC Regional Administrator, Region III NRC Senior Resident Inspector - Byron Station

ATTACHMENT 1

10 CFR 50.55a Relief Request I5R-01 Alternative Requirements for Repair/Replacement of Control Rod Drive Mechanism (CRDM) Canopy Seal Welds in Accordance with IWA-4000 in Accordance with 10 CFR 50.55a(z)(2)

- Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety -

Revision 0

10 CFR 50.55a Relief Request I5R-01 Alternative Requirements for Repair/Replacement of Control Rod Drive Mechanism (CRDM) Canopy Seal Welds in Accordance with IWA-4000 in Accordance with 10 CFR 50.55a(z)(2)

- Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety -

Revision 0

1. ASME Code Component(s) Affected Code Class: 1

Reference:

IWA-4000 Examination Category: NA Item Number: NA

Description:

Alternative Requirements for Repair/Replacement of Control Rod Drive Mechanism (CRDM) Canopy Seal Welds in Accordance with IWA-4000 Component Number: Reactor CRDM Canopy Seal Welds - Class 1 Appurtenance to the Reactor Vessel

2. Applicable Code Edition The Fifth Interval of the Byron Station (Byron), Units 1 and 2, Inservice Inspection (ISI)

Programs will be based on the American Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel (BPV) Code,Section XI, 2019 Edition.

3. Applicable Code Requirements The CRDM assemblies were designed and fabricated to the ASME Section III, 1974 Edition through Summer 1974 Addenda.

IWA-4000 of ASME Section XI requires that repairs be performed in accordance with the owners original construction Code of the component or system, or later editions of the Code. The canopy seal weld is described in Section III, and a repair to this weld would require the following activities:

a. Excavation of the rejectable indications,

b. A surface examination of the excavated areas,

c. Re-welding and restoration to the original configuration and materials, and

d. Final surface examination.

4. Reason for Request In accordance with 10 CFR 50.55a(z)(2), relief is requested on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

At this time, this request for relief is associated with a contingency repair. There currently are no known degraded canopy seal welds at Byron Station. The principal issues leading to this relief request are the excavation of indications contained within the existing weld, the accompanying dose received during the excavation and examination activities, and the weld material used for the repair or replacement.

Due to the nature of the flaw in the subject canopy seal weld, the excavation of the leaking portion of the weld would result in a cavity that extends completely through wall.

A liquid penetrant (PT) examination of this cavity is required to verify the removal of the rejectable flaw or to verify that the flaw is removed or reduced to an acceptable size.

This PT examination would deposit the penetrant materials onto the inner surfaces of the component. This material would not be readily removed prior to re-welding due to the

Page 1 of 8

10 CFR 50.55a Relief Request I5R-01 Alternative Requirements for Repair/Replacement of Control Rod Drive Mechanism (CRDM) Canopy Seal Welds in Accordance with IWA-4000 in Accordance with 10 CFR 50.55a(z)(2)

- Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety -

Revision 0

inaccessibility of the inside surface. The remaining penetrant material would introduce contaminants to the new weld metal and reduce the quality of the repair weld. The configuration of the canopy assembly would prevent the establishment and maintenance of an adequate back-purge during the welding process and would further reduce the quality of the repair weld.

Components associated with the canopy seal welds were assembled using threaded connections, were torqued and seal welded using consumable inserts. The threaded joints between the cap and rod travel housing, rod travel housing and latch housing, and latch housing and head adaptor provide the primary pressure boundary and structural support for the control rod drive mechanisms. The canopy seal weld overlay is for secondary pressure retention only and does not provide any structural support. The general configuration of the canopy seal welds along with their general location on the control rod drive mechanism assembly are provided in Figures I5R-01-1 and I5R-01-2.

The approximate diameter of the canopy seal welds ranges are 2" (for the upper canopy seal weld), 3.75" (for the middle canopy seal weld), and 6.45" (for the lower canopy seal weld).

The CRDM canopy seal welds are located above the Reactor Vessel Closure Head, which is highly congested and subject to high radiation levels. The high radiological dose associated with a CRDM canopy seal weld repair in strict compliance to these ASME Code requirements would be contrary to the intent of the as low as reasonably achievable (ALARA) radiological controls program. In order to reduce the exposure to personnel involved in the welding process, most of the repair activities would be performed remotely using robotic equipment to the extent practical. However, the required excavation and PT examinations would necessitate hands on access to the canopy weld. Based on expected radiation dose levels and time estimates to perform the excavation and PT examination for a single CRDM repair, the estimated total dose for these activities is estimated to be in excess of 0.600 person-Rem. This dose estimate is consistent with industry experience for similar activities.

IWA-4200 requires that the repair material conform to the original Design Specification or ASME Section III. In this case, the replacement material would have the same resistance to stress corrosion cracking as the original material. Use of the original material does not guarantee that the repaired component will continue to maintain leakage integrity throughout the intended life of the item.

In lieu of performance of PT examinations of CRDM seal weld repairs or replacement, a visual examination will be performed after the welding is completed. In addition, Alloy 52/52M nickel-based weld repair material will be used rather than austenitic stainless steel.

Alloy 52/52M nickel-based weld repair material was selected for the repair because of its resistance to stress corrosion cracking. The suitability of the replacement material will be evaluated for each application and determined to be compatible with the existing component and will provide a leakage barrier for the remainder of the intended life of the CRDM.

Page 2 of 8 10 CFR 50.55a Relief Request I5R-01 Alternative Requirements for Repair/Replacement of Control Rod Drive Mechanism (CRDM) Canopy Seal Welds in Accordance with IWA-4000 in Accordance with 10 CFR 50.55a(z)(2)

- Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety -

Revision 0

The alternative method of repair is being requested to facilitate contingency repair efforts during future outages within the Fifth ISI Interval. The alternative nondestructive examination method is being requested to facilitate examination of a repair of a CRDM canopy seal weld during the Fifth ISI Interval.

Industry experience with failure analyses performed on leaking canopy seal welds removed from service at other plants has attributed the majority of the cases to transgranular stress corrosion cracking (SCC). The size of the opening where the leakage occurs has been extremely small, normally a few thousandths of an inch. The crack orientations vary, but often radiate outward such that a pinhole appears on the surface, as opposed to a long crack. The SCC results from exposure of a susceptible material to residual stress, which is often concentrated by weld discontinuities, and to a corrosive environment, such as water trapped in the cavity behind the seal weld that is mixed with the air initially in the cavity, resulting in higher oxygen content than is in the bulk primary coolant. Based on this operating experience, there were no instances of significant degradation associated with any target surfaces as a result of identified canopy seal leakage.

Should a minor canopy seal weld leak develop during reactor startup, there are no components in the near vicinity of the CRDMs that would be adversely impacted by a canopy seal weld leak. There are also established reactor coolant system (RCS) leak detection methods available to detect RCS leakage from any location. If a leak developed, the leak could be detected by the containment area or process radiation monitoring system; or by the containment sump monitor. In addition, a RCS leakrate surveillance is conducted every shift that would identify leakage.

5. Proposed Alternative and Basis for Use The CRDM canopy seal weld flaws will not be removed, but an analysis of the repaired weldment will be performed in accordance with applicable provisions of ASME Section XI, Nonmandatory Appendix Q prior to entering Mode 4, to assure that the remaining flaw will not propagate unacceptably. The canopy seal weld is not a structural weld, nor a pressure-retaining weld, but provides a seal to prevent RCS leakage if the mechanical joint leaks. The canopy seal weld and associated overlay weld is separate from the structural pressure-retaining threaded segments of the CRDM. The threaded segments are not designed to be leak tight and, therefore, canopy seal welds are utilized as part of the plant design to minimize leakage. The threaded joints between the cap and rod travel housing, rod travel housing and latch housing, and latch housing and head adapter provide the primary pressure boundary and structural support for the control rod drive mechanisms. The proposed overlay simply replaces a leaking section of the canopy seal weld and, because it does not affect the threaded CRDM joint, does not change the probability of a CRDM failure.

The weld buildup is considered a repair in accordance with IWA-4110. Applicability of the original Code of construction or design specification is mandated because the weld is performed on an appurtenance to a pressure-retaining component. The alternative CRDM canopy seal weld repair uses a Gas Tungsten Arc Welding (GTAW) process controlled remotely. Should the need arise, a manual GTAW repair may be utilized.

Page 3 of 8 10 CFR 50.55a Relief Request I5R-01 Alternative Requirements for Repair/Replacement of Control Rod Drive Mechanism (CRDM) Canopy Seal Welds in Accordance with IWA-4000 in Accordance with 10 CFR 50.55a(z)(2)

- Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety -

Revision 0

Welding vendor selection would be based on vendor availability at the time of need, experience welding Alloy 52/52M, and proven performance of the vendor. Based on the accessibility of the canopy seal weld, weld overlay repairs in accordance with applicable provisions of ASME Section XI, Nonmandatory Appendix Q would be performed using an appropriate Welding Procedure Specification (WPS) along with welders or welding operators qualified in accordance with ASME Section IX, Qualification Standard for Welding and Brazing Procedures, Welders, Brazers, and Welding and Brazing Operators, requirements. Since the existing base materials are relatively thin, special consideration to travel speed and heat input would be critical parameters to be addressed during the weld qualification process.

Based on a review of documented industry experience, most of the difficulties attributed to Alloy 52/52M overlay repairs have been associated with overlays onto existing dissimilar metal welds on nozzle-to-safe end configurations or embedded flaw repairs on pressurized water reactor (PWR) upper head penetrations (i.e., overhead welding). All base materials associated with the canopy seal welds are stainless steel. Overhead welding is not a concern as the lower and upper canopy seal welds are oriented horizontally (see 2G position as depicted in Figure QW-461.4(b), Groove Welds in Pipe

- Test Position, of ASME Section IX). The middle canopy seal welds are also oriented horizontally (see 1G position as depicted in Figure QW-461.3(a) of ASME Section IX.

Constellation Energy Generation (CEG) procedure CC-AA-501-1028, Constellation Nuclear Welding Program High Risk/High Value (HR/HV) Welds, will be used to evaluate specific weld repairs and determine whether mock-ups are necessary to ensure a sound weld overlay is successfully applied. All associated canopy seal repair welding activities would be approved in accordance with a repair plan consistent with ASME Section XI, IWA-4150, Repair/Replacement Program and Plan.

A visual examination of the repaired/replaced weld will be performed using methods and personnel qualified to the standards of ASME Section XI VT-1 visual examination requirements. Conduct of the VT-1 visual examination would be dependent on accessibility of the inspection. If the canopy seal repair was associated with a control rod drive mechanism on the outer periphery of the core, it may be possible to perform the inspection through direct visual observation and a mirror. If inspection access is limited, VT-1 visual examinations would be conducted using remote visual equipment such as a video probe or camera equipment that accompanies the welding equipment.

Remote visual equipment resolution will be demonstrated prior to and upon completion of the examination(s) in accordance with ASME Section XI, IWA-2211, VT-1 Examination, and ASME Section V, Nondestructive Examination, Article 9, Visual Examination, requirements. Byron Station will utilize the requirements of ASME Section XI, Table IWA-2211-1, Visual Examinations, for procedure demonstrations required to support the VT-1 visual examinations. The repaired/replaced weld will be examined for quality of workmanship and discontinuities will be evaluated and dispositioned to ensure the adequacy of the new leakage barrier.

The automated GTAW weld repair and alternate VT-1 visual examination methods result in significantly lower radiation exposure because the equipment is remotely operated

Page 4 of 8 10 CFR 50.55a Relief Request I5R-01 Alternative Requirements for Repair/Replacement of Control Rod Drive Mechanism (CRDM) Canopy Seal Welds in Accordance with IWA-4000 in Accordance with 10 CFR 50.55a(z)(2)

- Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety -

Revision 0

after setup. A post-maintenance VT-2 visual examination will be performed at normal operating temperature and pressure during the System Leakage Test.

There is no applicable ASME Section XI, Examination Category or Item Number associated with this configuration as canopy seal welds are not subject to Table IWB-2500-1, Examination Categories, surface or volumetric examinations.

Additionally, replacement seal welds are specifically exempted from post-welding pressure testing per ASME Section XI, IWA-4540, Pressure Testing of Classes 1, 2, and 3 Items Following Repair/Replacement Activities, paragraph (b)(7). The only applicable ASME Section XI NDE requirements associated with the canopy seal welds are those associated with defect removal (i.e., IWA-4422.2.2, Defect Removal Followed by Welding or Brazing). A final surface examination is required in accordance with the Code of Construction (i.e., ASME Section III) and the original Design Specification.

The canopy seal welds are contained within the Class 1 system leakage test boundary and are examined by the VT-2 visual examination method each refueling outage as required by ASME Section XI, Table IWB-2500- 1, Examination Category B-P, All Pressure-Retaining Components, Item Number B15.10, Pressure -Retaining Components. In addition to the Class 1 system leakage test performed at the conclusion of each refueling outage, leakage checks at the beginning of each refueling outage and during forced outages are conducted in accordance with Byron Stations commitment to Generic Letter 88-05, Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants.

The control rod drive mechanism housings are not insulated and are oriented vertically.

Accessibility to conduct a VT-2 visual examination of the lower canopy seal welds is relatively unobstructed once the shroud access doors are opened. Staining, due to RCS leakage, would be evident at low points in this area. The intermediate and upper canopy seals cannot be observed directly and are examined to the extent practical in accordance with ASME Section XI, IWA-5241, Insulated and Noninsulated Components, paragraph (d). Radiation dose rate in this area does not prevent performance of the VT-2 visual examinations.

Repair/Replacement activities, using the process described in this relief request, shall be documented on the required forms (i.e., NIS-2, Form NIS -2 Owners Report for Repair/Replacement Activity). This relief request will be identified on the NIS-2 forms.

The repair documents will be reviewed by the Authorized Nuclear Inspector, and maintained in accordance with the requirements for archiving permanent plant records.

6. Duration of Proposed Alternative Relief is requested for the Fifth ISI Interval for Byron Station, Units 1 and 2.

Page 5 of 8 10 CFR 50.55a Relief Request I5R-01 Alternative Requirements for Repair/Replacement of Control Rod Drive Mechanism (CRDM) Canopy Seal Welds in Accordance with IWA-4000 in Accordance with 10 CFR 50.55a(z)(2)

- Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety -

Revision 0

7. Precedents

• Braidwood Station, Units 1 and 2, Third ISI Interval Relief Request I3R-11 was authorized per Nuclear Regulatory Commission (NRC) Safety Evaluation (SE) dated April 28, 2014 (ML14084A549).

• Braidwood Station, Units 1 and 2, Fourth ISI Interval Relief Request I4R-03 was authorized per NRC SE dated January 17, 2019 (ML18347B419). This Byron Fifth ISI Interval relief request utilizes a similar approach to the previously approved relief request for Braidwood.

• Byron Station, Units 1 and 2, Third ISI Interval Relief Request I3R-06 was authorized by NRC SE dated March 9, 2007 (ML070520480).

8. References None.

Page 6 of 8

10 CFR 50.55a Relief Request I5R-01 Alternative Requirements for Repair/Replacement of Control Rod Drive Mechanism (CRDM) Canopy Seal Welds in Accordance with IWA-4000 in Accordance with 10 CFR 50.55a(z)(2)

- Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety -

Revision 0

Figure I5R-01-1 Structure of CRDM Housing

Page 7 of 8 10 CFR 50.55a Relief Request I5R-01 Alternative Requirements for Repair/Replacement of Control Rod Drive Mechanism (CRDM) Canopy Seal Welds in Accordance with IWA-4000 in Accordance with 10 CFR 50.55a(z)(2)

- Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety -

Revision 0

Figure I5R-01-2 Sectional View of Canopy Seals

Page 8 of 8 ATTACHMENT 2

10 CFR 50.55a Relief Requests I5R-02 Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1)

- Acceptable Level of Quality and Safety -

Revision 0

10 CFR 50.55a Relief Request I 5R-02 Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1)

- Acceptable Level of Quality and Safety -

Revision 0

1. ASME Code Component(s) Affected Component Numbers: Byron Station, Units 1 and 2, Reactor Vessels 1RC01R (Unit 1) and 2RC01R (Unit 2)

Description:

Alternative Requirements for the Repair of Reactor Vessel Head Penetrations (VHPs) and J-groove Welds Code Class: Class 1 Examination Category: ASME Code Case N-729-6 Code Item: B4.60 Identification: Byron Units 1 and 2, VHP Numbers 1 through 78, (P-1 through P-78)

Previous repairs (I3R-14) : Unit 2, P-681 (I3R-19): Unit 1, P-31, P-43, P-64, and P-76 1 (I3R-20): Unit 2, P-61 Drawing Numbers: Various

2. Applicable Code Edition and Addenda Inservice Inspection and Repair/Replacement Programs: American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, 2019 Edition. Examinations of the VHPs are performed in accordance with 10 CFR 50.55a(g)(6)(ii)(D), which specifies the use of Code Case N-729-6, with conditions.

Code of Construction Reactor Pressure Vessel (RPV): ASME Section III, 1971 Edition through Summer 1973 Addenda.

3. 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 in accordance with the following requirements:

(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 by welding or brazing shall be in accordance with IWA-4411.

(1) Defect removal may include removal of all or a portion of the defective item, accompanied by installation of new material, either in accordance with the existing configuration or in a new configuration. Design or configuration changes shall meet

1 This relief request includes lnservice Inspection (ISI) examination requirements for repairs previously completed in accordance with I3R-14, I3R -19, and I3R -20 in the Third ISI Interval.

Page 1 of 9

10 CFR 50.55a Relief Request I 5R-02 Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1)

- Acceptable Level of Quality and Safety -

Revision 0

IWA-4311.

(2) Welding or brazing to restore the minimum required material thickness may be considered defect removal. In this case, IWA-4422 does not apply.

Note that use of the Mitigation of Defects by Modification provisions of IWA-4340 of ASME Section XI, 2011 Edition through the latest edition incorporated by reference in paragraph 10 CFR 50.55a(a)(1)(ii) is subject to the conditions of 10 CFR 50.55a(b)(2)(xxv)(B).

For the removal of defects by welding or brazing, ASME Section XI, IWA-4411 states, in part, the following.

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

The applicable requirements of the Construction Code required by IWA -4411 for the removal or mitigation of defects by welding from which relief is requested are as follows.

Base Material Defect Repairs:

For defects in base material, ASME Section III, NB-4131 requires that the defects are eliminated, repaired, and examined in accordance with the requirements of NB-2500.

These requirements include the removal of defects via grinding or machining per NB-2538. Defect removal must be verified by a Magnetic Particle (MT) or Liquid Penetrant (PT) examination in accordance with NB-2545 or NB-2546, and if necessary to satisfy the design thickness requirement of NB-3000, repair welding in accordance with NB-2539.

ASME Section III, NB-2539.1 addresses removal of defects and requir es defects to be removed or reduced to an acceptable size by suitable mechanical or thermal methods.

ASME Section III, NB-2539.4 provides the rules for examination of the base material repair welds and specifies they shall be examined by the MT or PT methods in accordance with NB-2545 or NB-2546. Additionally, if the depth of the repair cavity exceeds the lesser of 3/8-inch or 10% of the section thickness, the repair weld shall be examined by the radiographic method in accordance with NB-5110 using the acceptance standards of NB -5320.

Weld Metal Defect Repairs (This applies to the CRDM penetration J-Groove weld.)

ASME Section III, NB-4450 addresses repair of weld metal defects.

ASME Section III, NB-4451 states; that unacceptable defects in weld metal shall be eliminated and, when necessary, repaired in accordance with NB-4452 and NB-4453.

ASME Section III, NB-4452 addresses elimination of weld metal surface defects without subsequent welding and specifies defects are to be removed by grinding or machining.

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ASME Section III, NB-4453.1 addresses removal of defects in welds by mechanical means or thermal gouging processes and requires the defect removal to be verified with MT or PT examinations in accordance with NB-5340 or NB-5350 and weld repairing the excavated cavity. In the case of partial penetration welds where the entire thickness of the weld is removed, only a visual examination is required to determine suitability for re-welding.

As an alternative to the requirements above, repairs will be conducted in accordance with the appropriate edition of ASME Section III and the alternative requirements, based on WCAP-15987-P, Revision 2-P-A, Technical Basis for the Embedded Flaw Process for Repair of Reactor Vessel Head Penetrations, December 2003, (Refer to Reference 1, hereafter known as WCAP-15987-P).

4. Reason for Request Constellation Energy Generation (CEG), LLC will conduct examinations of the reactor Vessel Head Penetrations (VHPs) in accordance with Code Case N-729-6, as amended by 10 CFR 50.55a. Flaw indications that require repair may be found on the VHP tube material and/or the J-groove attachment weld(s) on the underside of the reactor vessel head.

Relief is requested from the requirements of ASME Section XI, IWA -4411 to perform permanent repair of future defects that may be identified on the VHPs and/or J-groove attachment weld(s) in accordance with the rules of the ASME Section III Construction Code as described in this relief request.

Specifically, relief is requested from:

• The requirements of ASME Section III, NB-4131, NB-2538, and NB-2539 to eliminate and repair defects in materials.

• The requirements of ASME Section III, NB-4450 to repair defects in weld metal.

5. Proposed Alternative and Basis for Use 5.1 Proposed Alternative CEG proposes to use the less intrusive embedded flaw process (Reference 1) for the repair of VHP(s) as approved by the NRC (Reference 2) as an alternative to the defect removal requirements of ASME Section XI and Section III.

5.1.1 The criteria for flaw evaluation established in 10 CFR 50.55a(g)(6)(ii)(D), which specifies the use of Code Case N-729-6, will be used in lieu of the Flaw Evaluation Guidelines specified by the NRC Safety Evaluation for WCAP-15987-P (Refer to Reference 5).

5.1.2 Consistent with WCAP-15987-P, Revision 2-P-A methodology, the following repair requirements will be performed.

1. Inside Diameter (ID) VHP Repair Methodology

a. An unacceptable axial flaw will be first excavated (or partially excavated) to a maximum depth of 0.125 inches. Although this depth differs from that specified in WCAP-15987-P, the cavity depth is not a critical parameter in the Page 3 of 9 10 CFR 50.55a Relief Request I 5R-02 Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1)

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implementation of a repair on the ID surface of the VHP. The goal is to isolate the susceptible material from the primary water (PW) environment.

The purpose of the excavation is to accommodate the application of at least two (2) weld layers of Alloy 52 or 52M, which is resistant to Primary Water Stress Corrosion Cracking (PWSCC), to meet that requirement. The depth specified in WCAP-15987-P is a nominal dimension and the depth needed to accommodate three (3) weld layers while still maintaining the tube ID dimension. Since two (2) weld layers will be applied, less excavation is required and only 0.125 inches of excavation is necessary. The shallower excavated cavity for two (2) weld layers would mean a slightly thinner weld, which would produce less residual stress.

The excavation will be performed using an Electrical Discharge Machining (EDM) process to minimize VHP tube distortion. After the excavation is complete, either an ultrasonic test (UT) or surface examination will be performed to ensure that the entire flaw length is captured. Then a minimum of two (2) layers of Alloy 52 or 52M weld material will be applied to fill the excavation. The expected chemistry of the weld surface is that of typical Alloy 52 or 52M weldment with no significant dilution. The finished weld will be conditioned to restore the inside diameter and then examined by UT and surface examination to ensure acceptability.

b. If required, unacceptable ID circumferential flaw will be either repaired in accordance with existing code requirements; or will be partially excavated to reduce the flaw to an acceptable size, examined by UT or surface examination, inlaid with Alloy 52 or 52M, and examined by UT and surface examination as described above.

2. Outside Diameter (O D) VHP and J-groove Weld Repair Methodology

a. An unacceptable axial or circumferential flaw in a tube below a J-groove attachment weld will be sealed off with an Alloy 52 or 52M weldment.

Excavation or partial excavation of such flaws is not necessary. The embedded flaw repair technique may be applied to OD axial or circumferential cracks below the J-groove weld because they are located away from the pressure boundary, and the proposed repair of sealing the crack with Alloy 690 weld material would isolate the crack from the environment as stated in Section 3.6.1 of the NRC Safety Evaluation for WCAP-15987-P.

b. Unacceptable radial flaws in the J-groove attachment weld will be sealed off with a 360 degree seal weld of Alloy 52 or 52M covering the entire weld.

Excavation or partial excavation of such flaws is not necessary.

c. If CEG determines an excavation is desired (e.g., boat sample), then

• The excavation will be filled with Alloy 52 or 52M material.

• It is expected that a portion of the indication may remain after the boat sample excavation; however, a surface examination will be performed on the excavation to assess the pre-repair condition.

• Depending on the extent and/or location of the excavation, the repair

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procedure requires the Alloy 52 or 52M weld material to extend at least one half inch outboard of the Alloy 82/182 to stainless steel clad interface.

d. Unacceptable axial flaws in the VHP tube extending into the J-groove weld will be sealed with Alloy 52 or 52M as discussed in Item 5.1.2.2.a above. In addition, the entire J -groove weld will be sealed with Alloy 52 or 52M to embed the axial flaw. The seal weld will extend onto and encompass the portion of the flaw on the outside diameter of the VHP tube.

e. For seal welds performed on the J-groove weld, the interface boundary between the J-groove weld and stainless steel cladding will be located to positively identify the weld clad interface to ensure that all of the Alloy 82/182 material of the J-groove weld is seal welded during the repair.

f. The seal weld that will be used to repair an OD flaw in the nozzles and the J-groove weld will conform to the following.

• Prior to the application of the Alloy 52 or 52M seal weld repair on the RPV clad surface, at least three beads (one layer) of ER309L stainless steel buffer will be installed 360° around the interface of the clad and the J-groove weld metal.

• The J-groove weld will be completely covered by at least three (3) layers of Alloy 52 or 52M deposited 360° around the nozzle and over the ER309L stainless steel buffer. Additionally, the seal weld will extend onto and encompass the outside diameter of the penetration tube Alloy-600 material by at least one half inch.

• The VHP tube will have at least two (2) layers of Alloy 52 or 52M deposited over the flaw on the VHP tube, extending out at least one half inch beyond the flaw, or to the maximum extent allowed by the nozzle geometry (e.g., limited length of the VHP tube).

g. Nondestructive examinations of the finished seal weld repair (i.e., Repair NDE) and during subsequent outages (i.e., ISI NDE) are summarized in the table below.

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Repair Location in Flaw Orientation Repair Repair NDE ISI NDE Original in Original Method Note (2) Note (2)

Component Component VHP Nozzle/Tube Axial or Seal weld UT and UT or Surface ID Circumferential Surface VHP Nozzle/Tube Axial or OD above J-groove Circumferential Note (1) Note (1) Note (1) weld VHP Nozzle/Tube Axial or OD below J-groove Circumferential Seal weld UT or Surface UT or Surface weld UT and UT and Surface, J-groove weld Axial Seal weld Surface, Notes (3) and (4)

Note (3)

UT and UT and Surface, J-groove weld Circumferential Seal weld Surface, Notes (3) and (4)

Note (3)

Notes: (1) Repair method to be approved separately by NRC.

(2) Preservice and lnservice Inspection to be consistent with 10 CFR 50.55a(g)(6)(ii)(D), which requires implementation of Code Case N-729-6 with conditions; or NRG-approved alternatives to these specified conditions.

(3) UT personnel and procedures qualified in accordance with 10 CFR 50.55a(g)(6)(ii)(D), which requires implementation of Code Case N-729-6 with conditions. Examine the accessible portion of the J-groove repaired region. The UT plus surface examination coverage equals to 100%.

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

a. Perform surface examination during the first and second refueling outage after installation or repair of the EFR.

b. When the examination results in 4.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 4.a above shall be applied during the next refueling outage.

5.1.3 J-Groove Weld ISI NDE Requirements Note 4 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 005, Revision 3 (Reference 9) provides the technical bases for Page 6 of 9 10 CFR 50.55a Relief Request I 5R-02 Alternative Requirements for the Repair of Reactor Vessel Head Penetrations in Accordance with 10 CFR 50.55a(z)(1)

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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 PWSCC. EFR installation, examination, and operational history indicate that the EFR performs acceptably.

Examination and removed sample history indicate that the flaws identified shortly after installation of EFR weld material were due to embedded weld discontinuities and not due to service induced degradation. With inspection of the EFR every other cycle of operation, the nozzles are adequately monitored for degradation by ultrasonic examination methods similar to the nozzles without EFR repairs.

CEG projects that the reduction of the PT examination of nozzles would result in a dose savings of approximately 0.4 to 0.7 REM per nozzle examination. The historical radiation dose associated with these examinations is presented in Reference 9, Table 2.

The proposed changes to the inservice examination requirements assure that the EFR repaired nozzles are 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 nozzles will continue to perform their required function.

5.1.4 Reporting Requirements and Conditions on Use CEG will notify NRC of the Division of Component Integrity or its successor of changes in indication(s) or findings of new indication(s) in the penetration nozzle or J-groove weld beneath a seal weld repair, or new linear indications in the seal weld repair, prior to commencing repair activities in subsequent outages.

5.2 Technical Basis for Proposed Alternative As discussed in WCAP-15987-P, the embedded flaw repair technique is considered a permanent repair. As long as a PWSCC flaw remains isolated from the Primary Water (PW) environment, it cannot propagate. Since an Alloy 52 or 52M weldment is considered highly resistant to PWSCC, a new PWSCC flaw should not initiate and grow through the Alloy 52 or 52M seal weld to reconnect the PW environment with the embedded flaw. Structural integrity of the affected J-groove weld and/or nozzle will be maintained by the remaining unflawed portion of the weld and/or the VHP.

Alloy 690 and Alloy 52/52M are highly resistant to stress corrosion cracking, as demonstrated by multiple laboratory tests, as well as over twenty years of service experience in replacement steam generators.

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, as described in WCAP-15987-P. Therefore, fatigue driven crack growth should not be a mechanism for further crack growth after the embedded flaw repair process is implemented.

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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 F is 8.2E-6 in/in/degree F as found in Section II part D. The Alloy 600 base metal has a coefficient of thermal expansion of 7.8E-6 in/in/degree F, a difference of about 5 percent. The effect of this small difference in thermal expansion is that the weld metal will contract more than the base metal when it cools, thus producing a compressive stress on the Alloy 600 tube or J-groove weld. This beneficial effect has already been accounted for in the residual stress measurements reported in the technical basis for the embedded flaw repair, as noted in the WCAP-15987-P.

WCAP-16401-P, Revision 1 (Reference 3) provides the plant-specific analysis performed for Byron and Braidwood Stations using the same methodology as WCAP-15987-P. This analysis provides the means to evaluate a broad range of postulated repair scenarios to the reactor vessel head penetrations and J-groove welds relative to ASME Code requirements for allowable flaw size and service life.

Based on Reference 3, a service life of at least twenty (20) years was determined for flaws in the VHP nozzles and a service life of at least forty (40) years was determined for flaws in the J-groove attachment welds.

The above proposed embedded flaw repair process is supported by applicable generic and plant specific technical bases, and is therefore considered to be an alternative to Code requirements that provides an acceptable level of quality and safety, as required by 10 CFR 50.55a(z)(1).

6. Duration of the Proposed Alternative Relief is requested for the Fifth ISI Interval for Byron Station, Units 1 and 2.

7. Precedents In Reference 8, the NRC provided their authorization to implement Relief Requests I3R- 09 and I3R-20, Revision 1 as a repair method for degradation identified in Reactor Vessel Head Penetrations. In Reference 11, the NRC provided their authorization to implement Relief Requests I3R-09 and I3R-20, Revision 2 during the Third ISI Interval.

The Fourth ISI Interval Relief Request utilized the same approach that was previously approved in Reference 11 for Byron Station Units 1 and 2. In Reference 12, the NRC provided their authorization to implement Relief Request I4R-10, Revision 2. This Fifth ISI Interval Relief Request utilizes the same approach that was previously approved in Reference 12.

8. References

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

2. 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.

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3. Westinghouse WCAP-16401-P, Revision 1, Technical Basis for Repair Options for Reactor Vessel Head Penetration Nozzles and Attachment Welds: Byron and Braidwood Units 1 and 2, January 2017.

4. Letter LTR-NRC- 03-61 from J. S. Galembush (Westinghouse Electric Company) to Terence Chan (U.S. NRC) and Bryan Benney (U.S. NRC), Inspection of Embedded Flaw Repair of a J-groove Weld, dated October 1, 2003.

5. Letter from R. J. Barrett, (U.S. NRC) letter to A. Marion (Nuclear Energy Institute),

Flaw Evaluation Guidelines, dated April 11, 2003.

6. American Society of Mechanical Engineers Boiler and Pressure Vessel Case N-729-6, Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration WeldsSection XI, Division 1.

7. Letter from R. Gibbs, (U.S. NRC) to C. M. Crane (EGC), Byron Station, Unit No.

2 - Relief Request I3R-14 for the Evaluation of Proposed Alternatives for lnservice Inspection Examination Requirements (TAC No. MD5230), dated May 23, 2007 (ML071290011).

8. Letter from Jacob Zimmerman, (U.S. NRC) to M. J. Pacilio, (EGC), Braidwood Station, Units 1 and 2 and Byron Station, Unit Nos. 1 and 2 - Relief Requests I3R- 09 and I3R-20 Regarding Alternative Requirements for Repair of Reactor Vessel Head Penetrations (TAC Nos. ME6071, ME6072, ME6073, and ME6074), dated March 29, 2012 (ML120790647).

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

10. Letter from J. Zimmerman, (U.S. NRC) to M. Pacilio (EGC), Byron Station, Unit No. 1 - lnservice Inspection Relief Request I3R-19: Alternative Requirements for the Repair of Reactor Vessel Head Penetrations (TAC Nos. ME5877 and ME5948), dated February 1, 2012 (ML112990783).

11. Letter fr om Justin C. Poole, (U.S. NRC) to Bryan C. Hanson (EGC), Byron Station, Units Nos. 1 and 2, and Braidwood Station, Units 1 and 2 - Relief from the Requirements of the ASME Code, dated January 21, 2016 (ML16007A185).

12. Letter from Kimberly J. Green, (U.S. NRC) to Bryan C. Hanson (EGC), Byron Station, Unit Nos. 1 and 2 - Request for Relief from the Requirements of the ASME Code (CAC Nos. MF8282 and MF8283), dated March 6, 2017 (ML17062A428).

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10 CFR 50.55a Relief Request I5R-03 Alternative Concerning ASME Section XI Repair/Replacement Documentation for Replacement of Pressure Retaining Bolting in Accordance with 10 CFR 50.55a(z)(1)

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10 CFR 50.55a Relief Request I5R-03 Alternative Concerning ASME Section XI Repair/Replacement Documentation for Replacement of Pressure Retaining Bolting in Accordance with 10 CFR 50.55a(z)(1)

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1. ASME Code Component(s) Affected All ASME pressure retaining bolting that is not classified as Examination Category B-G-1 (Class 1 pressure retaining bolting, greater than 2 in. in diameter), B -G-2 (Class 1 pressure retaining bolting, 2 in. and less in diameter), C-D (Class 2 pressure retaining bolting greater than 2 in. in diameter), or E-G (Class MC pressure retaining bolting).

2. Applicable Code Edition The Fifth Interval of the Byron Station (Byron), Units 1 and 2, Inservice Inspection (ISI)

Programs will be based on the American Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel (BPV) Code,Section XI, 2019 Edition.

3. Applicable Code Requirements IWA-4141 [2019 Edition] requires the Owner to provide or cause to be provided a Repair Replacement Program, a Repair Replacement Plan, and specification requirements for repair/replacement activities.

IWA-4142 [2019 Edition] requires the organization that performs repair/replacement activities shall establish a Quality Assurance Program for control of their activities in accordance with the Repair/Replacement Program and Plans.

IWA-4150(c) [2019 Edition] requires that a Repair/Replacement Plan be prepared in accordance with the Repair/Replacement Program whenever a repair/replacement activity is to be performed.

IWA-4511 [2019 Edition] requires that personnel performing nondestructive examination required by the Construction Code shall be qualified and certified in accordance with the Construction Code identified in the Repair/Replacement Plan.

IWA-6211(d) [2019 Edition] requires the Owner to prepare the Owners Repair/Replacement Certification Record, Form NIS-2 (Form NIS-2) upon completion of all required activities associated with the Repair/Replacement Plan to place the item in service.

IWA-6220 [2019 Edition] provides the requirements for tracking and approval of Repair/Replacement plans and Form NIS-2.

IWA-6350 [2019 Edition] requires that Repair/Replacement Plans and Form NIS-2 be retained.

4. Reason for Request In accordance with 10 CFR 50.55a(z)(1), relief is requested on the basis that the proposed alternative will provide an acceptable level of quality and safety.

Constellation Energy Generation (CEG) is requesting relief from the current ASME Section XI requirements to initiate a Repair/Replacement Plan and complete a Form NIS-2 for certain routine pressure retaining bolting replacement activities. Compliance with this ASME Code administrative requirement results in expending personnel resources that are better used on more safety significant activities. Eliminating this administrative burden will also streamline the planning and post maintenance review processes by involving fewer plant organizations (Engineering, Planning, Work Control, Maintenance, etc.) in the overall required maintenance activities. This

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proposed relief request is strictly limited to the administrative requirements of ASME Section XI and does not eliminate any ASME Section XI or Construction Code technical requirements associated with bolting materials or installation and maintenance requirements.

Based on the above discussion, reasonable assurance is still achieved by performing the remaining technical requirements when bolting is replaced.

5. Proposed Alternative and Basis for Use This relief request proposes to forego preparation and completion of a Repair/Replacement Plan and associated Form NIS-2 for replacement of pressure retaining bolting that is not classified as Examination Category B -G-1, B-G-2, C-D, or E-G. This relief request will not be applied to activities that involve replacement of bolting that has experienced unacceptable service-induced degradation or when involving a design change. Bolting replacement during normal maintenance work activities due to damage or loss will apply this relief request unless involving Examination Category B-G-1, B-G-2, C-D, or E-G components. Unacceptable service-induced degradation that cannot apply this relief request is defined as follows:

• Fractures and crack like flaws not caused by maintenance activities.

• More than one deformed or sheared thread in the zone of thread engagement that is due to a service-induced condition. Threads can often get deformed or sheared during the removal process and may not be considered service-induced degradation.

• Corrosion (e.g., boric acid, raw water) that reduces the bolt cross sectional area by more than 5%.

• Bending, twisting, or deformation of bolts determined to be from a service-induced condition.

• Degradation of protective coatings on bolting surfaces.

Quality Assurance Program and system/component specification requirements remain in place during application of this relief request; therefore, these technical requirements remain unchanged. The specific requirements will not be documented in a Repair/Replacement Plan but are currently implemented through the normal planning, procurement, and maintenance processes.

IWA-4150 Repair/Replacement Program and Plan administrative requirements applicable to this relief request are addressed as follows:

1. The Section XI Repair/Replacement Program remains applicable with editions and addenda(s) defined through application of the Repair/Replacement Program and ISI Plan and are therefore defined. NRC enforcement of Section XI and Construction Code requirements remain applicable because this relief request maintains Code requirements other than a documented Repair/Replacement Plan and NIS-2.

2. The applicable Construction Code Edition, Addenda, Cases, and Owners Requirements are included in plant records that may be retrieved when appropriate. Routine replacement of bolting is achieved through review of current component and procurement records with replacement material being

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of equal or more stringent requirements (e.g., replacing a Class 3 stud with a Class 1 stud).

3. This relief request will not be used for locations that have experienced unacceptable service-induced degradation or for a replacement involving a design change; therefore, there are no NDE detection methods, defect removal methods, welding requirements, acceptance NDE requirements, description of activity, life of component, or code stamping requirements involved.

4. Documentation of the work activity and replacement bolting is achieved through the normal processes of procurement, planning, and maintenance.

Close-out reviews are completed through the normal post-work review process to assure appropriate documentation of work performed and traceability is satisfied. Authorized Inspection Agency Review will remain in place but will not be documented on the Form NIS-2.

5. Availability of records for regulatory review of the Repair/Replacement Program and associated repair/replacement evaluations remain in place and available for review. There will be no Repair/Replacement Plan, but the associated work order and procurement documentation will be available for review by the Authorized Inspection Agency and regulatory authority when requested.

Replacement bolting will receive Construction Code and Owners Requirements NDE as part of the nor mal procurement and receipt inspection processes which identify applicable Construction Code and Owners Requirements. The Construction Code and Owners Requirements for NDE will be documented in the procurement and receipt records.

The current Form NIS-2 provides documented evidence of compliance with Section XI for repair/replacement activities by obtaining Owner and Authorized Inspection Agency signatures. This relief request proposes to forego preparation and completion of the Repair/Replacement Plan and associated Form NIS-2 for replacement of pressure retaining bolting that is not classified as Examination Category B-G-1, B-G-2, C-D, or E-G.

For routine replacement of bolting, where no unacceptable service induced degradation has been identified (i.e., lost bolting, bolting damaged during disassembly, bolting with corrosion that does not reduce the cross sectional area by more than 5%), the only ASME Section XI requirements currently invoked beyond normal work and procurement processes are administrative in nature and include completion of a Repair/Replacement Plan and completion of a Form NIS-2. For these replacements, the Repair/Replacement Plan provides no additional information other than reiteration of the work package instructions for bolting replacement. Similarly, since there are no additional technical requirements under Section XI for routine bolting replacement, completing the Form NIS-2 is merely generating an additional document for activities that are already covered through the work control process and procedures, along with the procurement process and procedures. The Form NIS-2 does not contain any specific information; therefore, it does not provide any information beyond indicating the bolting was replaced in

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accordance with the Owners processes and procedures, which ensure the Owners responsibilities under ASME Section XI have been met.

The proposed alternative will use existing work control, procurement, and records retention processes to assure adequate controls of these routine bolting replacements.

The Authorized Inspection Agency will have access to these site work control systems for review at their discretion. Should the Authorized Inspection Agency choose to review certain completed work orders that fall under this relief request, the work management system will be used to document any agency comments during the work order records review, commonly referred to as the post-work review process. In lieu of completing Repair/Replacement Plans and associated Forms NIS-2, a log will be maintained of work packages where the relief request has been invoked. The Authorized Inspection Agency will have access to the log, providing them the opportunity to review the work package instructions, along with the associated procurement documentation.

As stated above, a replacement bolting Repair/Replacement Plan and associated Form NIS-2 will not be completed for bolting replacement activities described above; therefore, these documents cannot be submitted as plant records. The procurement and work activity records will document technical requirements and work activities to allow subsequent review for adequacy and traceability, thereby meeting the intent of Code requirements and maintaining an acceptable level of quality and safety.

6. Duration of Proposed Alternative The proposed alternative to forego preparation and completion of a Repair/Replacement Plan and associated Form NIS-2 will be applicable as specified in Section 2 above or until such time as the NRC approves a similar administrative requirement relaxation in an NRC-approved applicable Code Edition in 10 CFR 50.55a or a Code Case in Regulatory Guide 1.147.

7. Precedents

• Braidwood Station, Unit Nos. 1 and 2; Byron Station, Unit Nos. 1 and 2; Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2; Clinton Power Station, Unit No. 1; Dresden Nuclear Power Station, Unit Nos. 2 and 3; James A. FitzPatrick Nuclear Power Plant; LaSalle County Station, Unit Nos. 1 and 2; Limerick Generating Station, Unit Nos. 1 and 2; Nine Mile Point Nuclear Station, Unit Nos. 1 and 2; Peach Bottom Atomic Power Station, Unit Nos. 2 and 3; Quad Cities Nuclear Power Station, Unit Nos. 1 and 2; and R.E. Ginna Nuclear Power Plant Fleet Relief Request RS 149, Revision 0, as supplemented, was authorized in an NRC safety evaluation (SE) dated August 5, 2021 (ML21216A220). This Byron Fifth ISI Interval Relief Request utilizes a similar approach that was previously authorized.

• Dresden Nuclear Power Station, Units 2 and 3 and Quad Cities Nuclear Power Station, Units 1 and 2 Sixth ISI Interval Relief Request I6R-06, Revision 0 was authorized in an NRC safety evaluation (SE) dated December 13, 2021 (ML21267A317). This Byron Fifth ISI Interval Relief Request utilizes a similar approach that was previously authorized.

8. References None.

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10 CFR 50.55a Relief Request I5R-04 Alternative to Permit Continued Application of Certain ASME Section XI 2013 Edition NDE Requirements for Short Term Fleet Consistency in Accordance with 10 CFR 50.55a(z)(1)

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10 CFR 50.55a Relief Request I5R-04 Alternative to Permit Continued Application of Certain ASME Section XI 2013 Edition NDE Requirements for Short Term Fleet Consistency in Accordance with 10 CFR 50.55a(z)(1)

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1. ASME Code Component(s) Affected Code Class: 1, 2, 3, and MC

Reference:

ASME Section XI, 2013 Edition ASME Section XI, 2019 Edition Examination Category: All, as related to NDE Methods and Requirements Item Number: All, as related to NDE Methods and Requirements

Description:

Alternative to Permit Continued Application of Certain ASME Section XI 2013 Edition NDE Requirements for Short Term Fleet Consistency Components: Class 1, 2, 3, and MC Component Examinations

2. Applicable Code Edition The Fifth Interval of the Byron Station (Byron), Units 1 and 2, Inservice Inspection (ISI)

Programs will be based on the American Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel (BPV) Code,Section XI, 201 9 Edition.

The Fourth Interval of the Byron, Units 1 and 2, ISI Programs are currently based on the ASME Section XI, 2013 Edition for all n on-destructive examination (NDE) r equirements.

The approval for using the 2013 Edition NDE requirements was made under NRC Safety Evaluation dated April 17, 2020 (ML20099D955). Similarly, all Constellation Energy Generation (CEG), LLC plants are currently standardized on the 2013 Edition for NDE requirements, either via this same relief request or under the normal plant ISI Program ASME Section XI code of record.

3. Applicable Code Requirements CEG is required to update the Byron, Units 1 and 2, ISI Program for the Fifth Interval to the latest Edition of the ASME Section XI, approved by the Nuclear Regulatory Commission (NRC) in 10 CFR 50.55a(a).

10 CFR 50.55a(g)(4)(ii), Applicable ISI Code: Successive code of record intervals, states:

"Inservice examination of components and system pressure tests conducted during successive code of record intervals must comply with the requirements of the latest edition and addenda of the ASME BPV Code incorporated by reference in paragraph (a) of this section 18 months before the start of the code of record interval" In accordance with 10 CFR 50.55a(a)(1)(ii)(C)(56 ), the 2019 Edition is the latest NRC approved version of ASME Section XI as of 18 months prior to the Byron Fifth Interval start date.

4. Reason for Request The Byron, Units 1 and 2, Fifth ISI Interval is currently scheduled to begin on July 16, 2025. The current fleet approval for all CEG plants to be governed by the NDE requirements of the 2013 Edition of ASME Section XI will expire for Byron on this date.

And based on 10 CFR 50.55a(g)(4)(ii) the Fifth Interval ISI Program for Byron will be required to update to the 2019 Edition.

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10 CFR 50.55a Relief Request I5R-04 Alternative to Permit Continued Application of Certain ASME Section XI 2013 Edition NDE Requirements for Short Term Fleet Consistency in Accordance with 10 CFR 50.55a(z)(1)

- Acceptable Level of Quality and Safety -

Revision 0

The CEG fleet of plants currently uses a common Code Edition specifically for NDE requirements such that inservice inspection related NDE activities across the fleet are performed in accordance with a standardized fleet program. The standardized NDE requirements and procedures are currently based on the 2013 Edition of ASME Section XI. While the overall ISI Program requirements are managed in site-specific documents under plant-specific code editions, the NDE program is controlled in common corporate procedures that are implemented across the entire CEG nuclear fleet.

This standardized fleet program is currently based on either the NRC approval referenced in Section 2 or is in accordance with the individual plant ASME Section XI code of record for the stations ISI Program. Byron was included in the April 17, 2020 approval to use ASME Section XI, 2013 Edition for the performance of NDE activities through the end of the current Fourth ISI Interval in order to allow use of common CEG procedures, processes, training, knowledge, and technical skills.

Due to the difference in interval start dates for Byron as compared to the other CEG sites, the upcoming Byron Fifth Interval would require development and implementation of a separate and unique NDE Program with unique supporting procedures affecting multiple organizations. Additionally, CEG would be required to create a unique skill set with unique training applicable only Byron. This minimizes the potential for using shared resources with other CEG sites, presents human performance concerns when applying different criteria, and prevents the use of common site and fleet procedures.

Pursuant to 10 CFR 50.55a(a)(z)(1), CEG requests authorization to continue to utilize the NDE specific requirements in the 2013 Edition of ASME Section XI for Byron, Units 1 and 2 for a period of time consistent with the First Period of the new Fifth Interval. The conclusion of the First Period for Byron, Units 1 and 2 coincides with the time when the balance of the plants in the CEG fleet will update to the latest ASME Section XI NDE requirements approved in 10 CFR 50.55a. The continued use of ASME Section XI 2013 Edition at Byron will be limited to the performance of NDE activities during the First Period and is subject to the applicable conditions contained in 10 CFR 55.55a(b)(2).

5. Proposed Alternative and Basis for Use CEG proposes to continue to utilize the 2013 Edition of ASME Section XI for NDE activities related to performing inservice inspections during the First Period of the Byron Fifth Interval. In implementing this proposed alternative, CEG will comply with all applicable NRC conditions and limitations related to the NDE requirements of the ASME Section XI 2013 Edition specified in 10 CFR 50.55a(b)(2).

Specifically, CEG is requesting the continued use of the provisions in the 2013 Edition of ASME Section XI for NDE activities associated with inservice inspections. These NDE activities are primarily driven from ASME Section XI Subarticles IWA-2100, IWA-2200, and IWA-2300. CEG will comply with all related NDE requirements of the 2013 Edition of ASME Section XI, as they interface with the General Requirements of IWA-1000 as well as certain Mandatory and Nonmandatory Appendices. Refer to Table 1 for a summary of the ASME Section XI NDE requirements related to Mandatory Appendices that will continue to use the requirements of the 2013 Edition of ASME Section XI.

Page 2 of 4 10 CFR 50.55a Relief Request I5R-04 Alternative to Permit Continued Application of Certain ASME Section XI 2013 Edition NDE Requirements for Short Term Fleet Consistency in Accordance with 10 CFR 50.55a(z)(1)

- Acceptable Level of Quality and Safety -

Revision 0

Table 1 - Proposed ASME Section XI Codes1 Code Provision ASME Section XI Edition/Addenda2 Sub-Section Article 2013 Edition IWA-General Requirements IWA-1000 X3 IWA-2000 X4 IWE-Requirements for Class IWE-2300 X5 MC Components Mandatory Appendices I X III X IV X VI X VII X6 VIII Note 7 Notes:

1. CEG will use the Articles and Mandatory Appendices of the 2013 Edition as shown in this table for NDE related activities. Articles, Mandatory Appendices (including Appendix XI, Repair/Replacement Activities for Class 3 Polyethylene Piping), and Nonmandatory Appendices not identified in the table will use the 2019 Edition.

2. CEG will comply with all NRC conditions, limitations, and restrictions specified in 10 CFR 50.55a(b)(2) for the applicable Edition referenced.

3. CEG will use the referenced standards from the 2013 Edition in IWA-1600 in relation to NDE related activities. All other paragraphs will utilize the 2019 Edition.

4. CEG will use IWA-2100, General, IWA-2200, Qualification Methods, and IWA-2300, Qualification of Nondestructive Examination Personnel, from the 2013 Edition.

However, CEG will use IWA-2400, IWA-2500, and IWA-2600 from the 2019 Edition.

5. CEG will use IWE-2300 from the 2013 Edition for requirements applicable to examination methods and qualification of personnel.

6. Appendix VII will be implemented from the 2013 Edition as amended by 10 CFR 50.55a(b)(2)(xviii)(D) to use Table VII -4110- 1 and Subarticle VIII-2200 from the 2010 Edition.

7. As required by 10 CFR 50.55a, CEG implements the latest approved Edition of Appendix VIII.

This request will allow Byron NDE requirements to remain standardized with the rest of the CEG fleet until the balance of the plants update to the latest approved edition of ASME Section XI, nominally at the end of the Byron First Period of the Fifth Interval.

It is important to note that this request does not apply to the balance of the ISI Program including the selection, planning, and scheduling of ISI examinations and tests as defined in IWB-, IWC-, IWD-, IWE-, and IWF-2500 or NRC approved ISI alternatives.

Therefore, ISI examinations and tests will be selected, planned, and scheduled in accordance with 2019 Edition of ASME Section XI which is the Fifth Interval Code of Record.

Page 3 of 4 10 CFR 50.55a Relief Request I5R-04 Alternative to Permit Continued Application of Certain ASME Section XI 2013 Edition NDE Requirements for Short Term Fleet Consistency in Accordance with 10 CFR 50.55a(z)(1)

- Acceptable Level of Quality and Safety -

Revision 0

Continued application of the 2013 Edition of ASME Section XI specifically for NDE requirements (consistent with the other CEG sites) will provide an acceptable level of quality and safety, and will enhance the effective management and implementation of the NDE activities at Byron. Approval of this request supports CEGs ability to maximize efficiencies, minimize human performance and procedure error traps, and optimize the use of internal operating experience as all CEG units will continue to utilize common procedures, processes, training, and knowledge for NDE implementation.

6. Duration of Proposed Alternative Relief is requested for the Fifth ISI Interval, First Inspection Period, for Byron, Units 1 and 2.

7. Precedents

• Exelon Generation Company Fleet Letter Request RS-19-089 (JAFP-19-0084)

"Byron Station, Unit Nos. 1 and 2; Dresden Nuclear Power Station, Units 2 and 3; James A. Fitzpatrick Nuclear Power Plant; LaS alle County Station, Units 1 and 2; Limerick Generating Station, Units 1 and 2; and Quad Cities Nuclear Power Station, Units 1 and 2 - Request to Use Provisions in the 2013 Edition of the ASME Boiler and Pressure Vessel Code for Performing Non-Destructive Examinations (EPID L-2019-LLR-0080)" was approved by the NRC Safety Evaluation (SE) dated April 17, 2020 (ML20099D955).

• Arkansas Nuclear One, Unit 2 (ANO-2) Relief Request ANO2-ISI-021, "Arkansas Nuclear One, Unit 2 - Request for Alternative ANO2-ISI-021 to Permit Continued Application of the 2007 Edition through the 2008 Addenda of the ASME Code (EPID L-2018-LLR-0122)" was approved by the NRC SE dated June 11, 2019 (ML19156A400).

• Dresden Nuclear Power Station (Dresden), Units 2 and 3 Relief Request I6R-10, Dresden Nuclear Power Station, Units 2 and 3 - Authorization and Safety Evaluation for Alternative Request I6R-10 (EPID: L-2022-LLR-0057) was approved by the NRC SE dated January 6, 2023 (ML23004A171).

• Quad Cities Nuclear Power Station (Quad Cities), Units 1 and 2 Relief Request I6R-09, Quad Cities Nuclear Power Station, Units 1 and 2 - Authorization and Safety Evaluation for Alternative Request I6R-09, Revision 0 (EPID L-2022-LLR-0059) was approved by the NRC SE dated February 3, 2023 (ML23033A098).

8. References None

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