AEP-NRC-2024-29, (CNP) Unit 2 - Request for Relief Related to American Society of Mechanical Engineers (ASME) Code Case N-729-6 Supplemental Examination Requirements, ISIR-5-07

(CNP) Unit 2 - Request for Relief Related to American Society of Mechanical Engineers (ASME) Code Case N-729-6 Supplemental Examination Requirements, ISIR-5-07

ML24094A283
Person / Time
Site:	Cook
Issue date:	04/03/2024
From:	Ferneau K Indiana Michigan Power Co, (Formerly Indiana & Michigan Power Co)
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
AEP-NRC-2024-29
Download: ML24094A283 (1)

Text

Indiana Michigan Power INDIANA Cook Nuclear Plant MICHIGAN One Cook Place POWIR - Bridgman, Ml 49106 indianamichiganpower.com An MP Company

BOUNDLESS ENERGY-

April 3, 2024 AEP-NRC-2024-29 10 CFR 50.55a

Docket No.: 50-316

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

Request for Relief Related to American Society of Mechanical Engineers (ASME)

Code Case N-729-6 Supplemental Examination Requirements, ISIR-5-07

Pursuant to 10 CFR 50.55a(z)(2), Indiana Michigan Power Company (l&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Unit 2, requests Nuclear Regulatory Commission approval of the enclosed request for an alternative for CNP Unit 2, based upon the specified Code Case requirements representing a hardship or unusual difficulty without a compensating increase in the level of quality and safety. Enclosure 1 to this letter identifies the affected components, applicable ASME Boiler and Pressure Vessel Code (Code) Case requirements, reason for request, proposed alternative, and basis for use. The alternative is proposed to be applied during the next operating cycle and will conclude at the end of the next refueling outage, U2C29.

As part of the current CNP Unit 2 refueling outage, U2C28, the Unit 2 Reactor Vessel Closure Head (RVCH) visual examination was performed in accordance with the requirements in ASME Section XI Code Case N-729-6, Table-1. This subsequent examination was a commitment as a part of relief request ISIR-5-06 that was accepted on January 4, 2023 (ML22363A562). During this subsequent examination, it was discovered that relevant conditions of boric acid deposits, discoloration and entrained residues exist at one nozzle on the Unit 2 RVCH.

Code Case N-729-6, Paragraph -3142.2, requires nozzles with relevant conditions to have supplemental examinations consisting of a volumetric examination of the nozzle tube and surface examination of the partial-penetration weld or surface examination of the nozzle tube inside surface, the partial-penetration weld, and nozzle tube outside surface below the weld, in accordance with Paragraph -3200(b). As described in Enclosure 1 of this request, l&M is requesting an alternative to the specified requirements of Code Case N-729-6, Paragraph -3142.2, pursuant to 10 CFR 50.55a(z)(2), as the provisions that require supplemental examinations represent a hardship or unusual difficulty without a compensating increase in the level of quality and safety. l&M submitted the first relief request for CNP Unit 2 on May 4, 2021 (ML21124A237). The safety evaluation was received on May 12, 2021 (ML21130A008). l&M submitted a subsequent relief request for CNP Unit 2 on October 24, 2022 (ML22297A211 ). The safety evaluation was received on January 4, 2023 (ML22363A562).

U. S. Nuclear Regulatory Commission AEP-NRC-2024-29 Page2

Based on the repeat nature of this leakage, l&M has taken corrective actions to prevent reoccurrence of thermocouple sealing assembly (TECSA) leakage contacting the RVCH in the future by updating the procedure to include necessary steps to verify compression of the gaskets for all TECSA seals, increased the torque used on the seals based on vendor input, and developed activities to capture and clean any leakage identified during reactor startup. Further, l&M plans to replace the TECSA seals during the U2C29 refueling outage in the Fall of 2025 with a new seal design.

Approval of the proposed relief is requested prior to entry into Mode 2, which is scheduled to occur on or about April 26, 2024.

There is one new commitment in this letter, as specified in Enclosure 2. Should you have any questions, please contact Mr. Michael K. Scarpello, Director of Regulatory Affairs, at (269) 466-2649.

Sincerely,

~Fial~

Site Vice President

RAW/sjh

Enclosures:

1. Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2) Request for an Alternative to American Society of Mechanical Engineers (ASME) Code Case N-729-6 for Replacement Reactor Vessel Closure Head Penetration Nozzles.

2. Regulatory Commitment for Reactor Vessel Closure Head (RVCH) Inspection.

c: EGLE - RMD/RPS J. B. Giessner - NRC Region Ill N. Quilico - MPSC NRC Resident Inspector R. M. Sistevaris - AEP Ft. Wayne S. P. Wall - NRC, Washington D.C.

A. J. Williamson - AEP Ft. Wayne

j Enclosure 1 to AEP-NRC-2024-29

Indiana Michigan Power Company - Donald C. Cook Nuclear Plant-Unit 2 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2)

Request for an Alternative to the American Society of Mechanical Engineers (ASME) Code Case N-729-6 for Replacement Reactor Vessel Closure Head Penetration Nozzles

1. ASME Code Component Affected

Components I Head with nozzles and partial-penetration welds of Primary Water Stress Corrosion Cracking Numbers: (PWSCC) resistant materials ASME Boiler and Pressure Vessel Code (Code),

Code Class: Class 1 ASME Section XI 2013 Edition and no addenda

ASME Section XI, Division 1, Code Case N-729-6,

References:

Alternative Examination Requirements for Pressurized Water Reactor (PWR) Reactor Vessel Upper Heads with Nozzles Having Pressure-retaining Partial-Penetration Welds,Section XI, Division 1 Table IWB-2500, Category B-P Examination Category: Table 1 of ASME Code Case N-729-6 Item Number( s ): 84.30 Pressure retaining components

Reactor Vessel Closure Head (RVCH) with Description : nozzles and partial-penetration welds of primary water stress corrosion cracking PWSCC-resistant materials

Penetration - 76 Unit/ Donald C. Cook Nuclear Plant, Unit 2 (CNP)/ 5th Inspection Ten-Year In-service Inspection (ISi) Interval Interval (March 1, 2020 to February 28, 2030)

Applicability: to AEP-NRC-2024-29 Page2

2. Applicable Code Edition and Addenda

The Fifth Ten-year ISi interval Code of Record for CNP is the 2013 Edition of ASME Code,Section XI, no addenda.

Examinations of the RVCH and penetration nozzles are performed in accordance with ASME Code Case N-729-6, Alternative Examination Requirements for Pressurized Water Reactor (PWR) Reactor Vessel Upper Heads with Nozzles having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1 (Reference 1 ), as conditioned by 10 CFR 50.55a(g)(6)(ii)(D).

10 CFR 50.55a(g)(6)(ii)(D) requires, in part, that licensees of pressurized water reactors shall augment the ISi program with ASME Code Case N-729-6 subject to the conditions specified in paragraphs (g)(6)(ii)(D)(2) through (8) of this section.

3. Applicable Code Requirement

10 CFR 50.55a(g)(6)(ii)(D)(1) requires:

(D)Augmented ISi requirements: Reactor vessel head inspections-(1) Implementation. Holders of operating licenses or combined licenses for pressurized-water reactors as of or after June 3, 2020, shall implement the requirements of ASME BPV Code Case N-729-6 instead of ASME BPV Code Case N-729-4, subject to the conditions specified in paragraphs (g)(6)(ii)(D)(2) through (8) of this section, by no later than one year after June 3, 2020. All previous NRC approved alternatives from the requirements of paragraph (g)(6)(ii)(D) of this section remain valid.

Paragraph-3141 of Code Case N-729-6 states, regarding in-service visual examinations (VE):

(a) The VE required by -2500 and performed in accordance with IWA-2200 and the additional requirements of this Case shall be evaluated by comparing the examination results with the acceptance standards specified in -3142.1.

(b) Acceptance of components for continued service shall be in accordance with-3142.

(c) Relevant conditions for the purposes of the VE shall include evidence of reactor coolant leakage, such as corrosion, boric acid deposits, and discoloration.

Paragraph -3142.1 Acceptance by VE of Code Case N-729-6 states:

(a) A component whose VE confirms the absence of relevant conditions shall be acceptable for continued service.

(b) A component whose VE detects a relevant condition shall be unacceptable for continued service until the requirements of (1 ), (2), and (c) below are met.

(1) Components with relevant conditions require further evaluation. This evaluation shall include determination of the source of the leakage and correction of the source of leakage in accordance with -3142.3.

(2) All relevant conditions shall be evaluated to determine the extent, if any, of degradation. The boric acid crystals and residue shall be removed to the extent necessary to allow adequate examinations and evaluation of degradation, and a subsequent VE of the previously obscured surfaces shall be performed, prior to return to AEP-NRC-2024-29 Page 3

to service, and again in the subsequent refueling outage. Any degradation detected shall be evaluated to determine if any corrosion has impacted the structural integrity of the component. Corrosion that has reduced component wall thickness below design limits shall be resolved through repair/replacement activity in accordance with IWA-4000.

(c) A nozzle whose VE indicates relevant conditions indicative of possible nozzle leakage shall be unacceptable for continued service unless it meets the requirements of-3142.2 or -

3142.3.

Paragraph-3142.2 Acceptance of Supplemental Examination of Code Case N-729-6 states:

A nozzle with relevant conditions indicative of possible nozzle leakage shall be acceptable for continued service if the results of supplemental examinations [-3200(b )] meet the requirements of -3130.

Paragraph -3142. 3 Acceptance by Corrective Measures or Repair/Replacement Activity of Code Case N-729-6 states:

(a) A component with relevant conditions not indicative of possible nozzle leakage is acceptable for continued service if the source of the relevant condition is corrected by a repair/replacement activity or by corrective measures necessary to preclude degradation.

(b) A component with relevant conditions indicative of possible nozzle leakage shall be acceptable for continued service if a repair/replacement activity corrects the defect in accordance with IWA-4000.

Paragraph 3200(b) Supplemental Examinations of Code Case N-729-6 states:

(b) The supplemental examination performed to satisfy -3142.2 shall include volumetric examination of the nozzle tube and surface examination of the partial-penetration weld, or surface examination of the nozzle tube inside surface, the partial penetration weld, and nozzle tube outside surface below the weld, in accordance with Figure 2, or the alternative examination area or volume shall be analyzed to be acceptable in accordance with Mandatory Appendix I. The supplemental examinations shall be used to determine the extent of the unacceptable conditions and the need for corrective measures, analytical evaluation, or repair/ replacement activity.

4. Reason for Request

Indiana Michigan Power Company (l&M), the licensee for CNP Unit 2, is requesting approval of an alternative to the specified requirements of ASME Code Case N-729-6, Paragraph -3142.2, pursuant to 10 CFR 50.55a(z)(2), as the provisions that require a supplemental examination represent a hardship or unusual difficulty without a compensating increase in the level of quality and safety.

l&M performed the VE of the RVCH nozzle penetrations during the current Unit 2 refueling outage (U2C28) in accordance with ASME Code Case N-729-6, Table 1. During U2C28, the RVCH nozzle penetrations were examined in the as-found condition using ASME Code Case N-729-6. During the examination, compressed air was used to remove the lightly adhered accumulated residues. Boric acid deposits, discoloration and entrained residues that remained after light cleaning were determined to be relevant conditions in the as-found state. One penetration, specifically penetration 76, was determined to have relevant conditions pursuant to Code Case N-729-6, Paragraph -3141(c). to AEP-NRC-2024-29 Page4

The l&M qualified examiner concluded that the relevant condition did not have active leakage characteristics because the pattern of residue on the nozzle was not consistent with the traditional patterns seen in Control Rod Drive Mechanism (CROM) nozzle leaks based on Reference 2.

The leakage is understood to be from the in-core instrumentation (ICI) thermocouple sealing assemblies (TECSA). Although the source of leakage provides the likely cause of the conditions, based on the guidance in Regulatory Issue Summary (RIS) 2018-06 (Reference 3), it could not be absolutely refuted that the relevant conditions identified in the as-found examination were not masking relevant conditions indicative of nozzle leakage. A review of previous inspection results performed in U2C26 and U2C27 noted light debris but no evidence of degradation or active leakage specifically related to penetration 76. The previous inspection was performed after the issuance of RIS 2018-06.

A review of previous inspection results noted discoloration/staining and debris but no evidence of degradation or active leakage. The previous inspections were performed after the issuance of RIS 2018-06, and the noted conditions required relief requests, which were approved in May 2021 and January 2023 (References 5 and 13). After the previous inspections, all of the nozzles having relevant conditions were cleaned, and as-left inspections were performed. The as-found condition of the Unit 2 RVCH in U2C28 is similar to the condition identified during the as-left inspection, except for 1 previously acceptable nozzle having relevant conditions. It is noted that the three outages do not have any nozzles with relevant conditions in common with the other, indicating that all of the nozzles that previously had relevant conditions have been proven to not be leaking.

The relevant conditions on penetration 76 require consideration, per RIS 2018-06, that some or all of the relevant conditions possibly came from the nozzle. Code Case N-729-6, Paragraph -3142.2, requires that nozzles with relevant conditions undergo supplemental examinations consisting of a volumetric examination of the nozzle tube and surface examination of the partial-penetration weld or surface examination of the nozzle tube inside surface, the partial-penetration weld, and nozzle tube outside surface below the weld, in accordance with Paragraph -3200(b).

Cleaning with steam and a subsequent bare metal visual (BMV) examination were completed in U2C28, and all relevant conditions were removed. No degradation was noted by the subsequent BMV examination. The steam cleaning returns the unit to service with a reactor head free of relevant conditions that will support implementation of an effective subsequent visual examination during the next outage which, as discussed below, is the alternative proposed in support of this relief request.

Removing boric acid deposits, discoloration and entrained residues also precludes RVCH degradation during the next operating cycle.

While these actions are insufficient to absolutely determine that the nozzle is free from reactor coolant pressure boundary leakage, they support a conclusion that no significant degradation of the RVCH has occurred as a result of the identified TECSA seal leakage.

5. Proposed Alternative and Basis for Use

10 CFR 50.55a(z)(2) requires demonstrating that compliance with the specified requirements represent a hardship or unusual difficulty without a compensating increase in the level of quality and safety. This section discusses the identified hardship, proposed alternative, and supporting basis that shows the supplemental volumetric and surface examinations result in a hardship with limited increase in quality and safety. to AEP-NRC-2024-29 Page 5

Identified Hardship per 10 CFR 50.55a(z)(2)

In order to perform the supplemental volumetric and surface examinations required in Code Case N-729-6, Paragraph -3200(b ), it would be necessary to mobilize equipment and approximately 16 qualified personnel to the site. The supplemental examinations require access to the underside of the highly contaminated RVCH. The bottom of the RVCH also exposes personnel to elevated dose rates not accounted for in the dose plans for U2C28. The additional dose for this work described above is estimated to be approximately 754 mRem. Providing access to the underside of the RVCH is an infrequently performed evolution that has not been performed at l&M in approximately 17 years.

While manageable, additional risk is incurred in setting up an atypical configuration where the RVCH rests levelly at a higher elevation on a head stand extension and Delrin blocks to provide access for the examination equipment and personnel. Setting up the Delrin blocks in the correct configuration to prevent damage to the RVCH will require additional training, time, and resources.

It is estimated that mobilization of contract personnel and equipment, and completion of the required supplemental examinations would take approximately ten days. This would add extra duration to the U2C28 outage, and all remaining outage activities associated with reactor assembly and startup would be delayed by at least ten days, subject to the availability of equipment and personnel to perform the examinations.

Additionally, performing these supplemental examinations will require l&M to meet all of the "needed" elements within MRP-384 based on the requirements of NEI 03-08 (References 6 and 7). This would require l&M to plan all of the examinations, train site personnel, set up the RVCH for examination, and perform all necessary oversight of the vendor before the vendor comes onsite to perform the examinations, on an expedited basis. This oversight would require l&M to work with the vendor to plan, create appropriate mock-ups, review all vendor procedures and qualifications, and create two different sets of training for the vendor/oversight personnel related to all of the topics identified in MRP-384. Because this would typically be done in advance of a refueling outage over a period of weeks, this again increases risk. MRP-384 includes the following training topics for oversight personnel:

• Examination procedure scope, limitations, and requirements

• Examination techniques

• Data analysis software

• Data analysis process

• Fabrication and geometrical indications, repair indications, and flaw indications

• Signal response characteristics

• Techniques for resolving and sizing flaws

• Documentation of the examination results (e.g., calibration, indications, lack of coverage, etc.)

The topics for Site Specific training would include the following:

• Site examination scope of work

• Procedure requirements and applicable technical justifications

• Plant-specific review of component and examination history

• Lessons learned from previous examinations

• Repair history, geometry, and previous indications

• Coverage, access restriction issues, or other factors that could impact the examination

• Industry and site-specific operational experience

• PWSCC flaw signal responses from field data

• Fabrication reflector signal responses from field data

• Leak path responses from field data

• Plant-specific mockup data, where applicable to AEP-NRC-2024-29 Page6

• Data quality expectations related to o Probe contact o Scanning surface distortion o Configuration-specific anomalies

Once the vendor personnel are on site and have set up all equipment, there will be additional "needed" elements to follow, including independent analysis of the data and comparing the current outage data to past examination data. l&M would be required to ensure that vendor personnel performing these examinations are properly qualified. Currently, l&M has one in-house Quality Control individual who is qualified to perform or provide oversight of all training related to ultrasonic examinations.

For these reasons, the supplemental examinations represent a hardship or unusual difficulty, pursuant to 10 CFR 50.55a(z)(2).

Proposed Alternative

As an alternative to performing supplemental examinations required by Paragraph -3142.2, l&M proposes performing the Code Case required BMV examination of the CNP Unit 2 RVCH in the next refueling outage (U2C29) in accordance with the latest revision of Code Case N-729 endorsed in 10 CFR 50.55a. The examination will be conducted in accordance with Paragraph -3140, and the results will be evaluated in accordance with Paragraph -3142. l&M considers this to be a regulatory commitment (See Enclosure 2 of this letter). As discussed above, l&M performed steam cleaning of affected areas of the RVCH and a post-cleaning BMV examination. The subsequent examination did not reveal any RVCH degradation or further relevant conditions. These actions in conjunction with the corrective actions discussed below are intended to provide a high level of confidence that l&M will have the necessary baseline conditions to support performing an effective subsequent visual examination as proposed in this alternative.

Basis for Use and Limited Increase in Quality and Safety

The items listed below provide the information l&M used to provide reasonable assurance that the relevant conditions are not indicative of RVCH nozzle leakage. The volumetric and surface examinations would confirm l&M's conclusion that the identified conditions are not nozzle leakage.

However, in the unlikely event that a nozzle leak exists or develops, the proposed alternative, the evaluation of the structural integrity of the replacement PWSCC resistant RVCH, and the CNP leakage detection program serve as a basis for concluding that nozzle leaks occurring during the next operating cycle would be detected prior to developing into a safety concern.

In U1C30 l&M submitted a Relief Request (Reference 8) due to identified TECSA seal leakage and reactor head vent disassembly practices as the source of leakage onto Unit 1 's RVCH where 17 penetrations were identified with relevant conditions. Due to the effectiveness of the corrective actions no relevant conditions were identified in the subsequent BMV during U1 C31.

In the Relief Request submitted for Unit 2 during U2C26 (Reference 4) l&M identified the discoloration and accumulated debris on the Unit 2's RVCH to be due to the ventilation system and cleanup of leakage from poor head vent piping disassembly practices. In performing the subsequent BMV for Reference 4 in U2C27, the corrective actions were effective in that all penetrations that had relevant conditions in U2C26 no longer had relevant conditions in U2C27, while two new penetrations had relevant conditions related to TECSA seal leakage.

During the current refueling outage, U2C28, one penetration with relevant conditions was identified and is discussed in this request. This penetration is a different penetration than those identified in U2C26 and U2C27. The implemented corrective actions still prove to be moderately effective as the quantitative number of relevant conditions is lower than previously identified. The relevant condition to AEP-NRC-2024-29 Page 7

identified in U2C28 is consistent with those identified in U2C27 as being a result of leakage from TECSA seals, but the penetrations identified in U2C27 no longer had relevant conditions when inspected in U2C28, validating that corrective actions taken were moderately effective.

Based on a review of the corrective action program and the current leakage path, TE CSA seal leakage contributed solely to the U2C28 relevant conditions. Based on the repeat nature of this leakage, l&M has taken corrective actions to prevent reoccurrence of TECSA seal leakage contacting the RVCH in the future by updating the procedure to include necessary steps to verify compression of the gaskets for all TECSA seals, increased the torque used on the seals based on vendor input, and developed activities to capture and clean any leakage identified during reactor startup. Further, l&M plans to replace the TECSA seals in the U2C29 refueling outage in the Fall of 2025 with seals of a different design. Based on this, the supplemental volumetric and surface examinations do not provide a compensating increase in quality or safety. Each item mentioned above is discussed in detail below.

• Review of the previous examination shows absence of degradation on the RVCH.

The penetration nozzles on the Unit 2 RVCH were examined in refueling outage U2C26 (Spring 2021 ), consistent with Code Case N-729-6. For the U2C26 inspection, Code Case N-729-6 and the supplemental guidance in RIS 2018-06 were used to determine if the identified discoloration and accumulated residues represented relevant conditions. Based on the guidance in RIS 2018-06, the noted relevant conditions that could not be removed by light cleaning could not be dispositioned regardless of the identified source. Following carbon dioxide cleaning, l&M performed a BMV examination; the BMV showed that no significant degradation existed on the affected areas of the RVCH. Relief was granted for those relevant conditions (Reference 5), and a commitment to re-inspect the RVCH resulted in inspecting the RVCH in U2C27. For all of the penetrations that had relevant conditions in U2C26, the subsequent examination revealed no evidence of through-wall leakage or a degraded condition for those nozzles.

During the subsequent examination in U2C27 (Fall 2022) per Code Case N-729-6, two different nozzles had relevant conditions identified. l&M assessed the identified relevant conditions and, for the reasons stated in that relief request, concluded that the relevant conditions were the result of a source other than RVCH penetration leakage. However, using the direction provided in RIS 2018- 06, the possibility of RVCH penetration leakage could not be completely refuted since the relevant conditions could have been masking RVCH penetration leakage. Following carbon dioxide cleaning, l&M performed a BMV examination; the BMV showed that no significant degradation existed on the affected areas of the RVCH.

Relief was granted for those relevant conditions (Reference 13), and a commitment to re-inspect the RVCH resulted in inspecting the RVCH in U2C28. For all of the penetrations that had relevant conditions in U2C27, the subsequent examination revealed no evidence of through-wall leakage or a degraded condition for those nozzles.

The source of the relevant conditions discovered in U2C28 is consistent with, but involves a different nozzle penetration, than those previously identified in U2C27. As with the U2C27 relevant conditions, l&M posits that the relevant conditions discovered in U2C28 do not represent a degraded condition of the RVCH. l&M assessed the identified relevant conditions and, for the reasons stated in this relief request, concluded that the relevant conditions are the result of a source other than RVCH penetration leakage. However, again using the direction provided in RIS 2018- 06, the possibility of RVCH penetration leakage cannot be completely refuted since the relevant conditions could be masking RVCH penetration leakage.

• The pattern of residue is not consistent with RVCH nozzle leaks, and the source causing the relevant conditions has been identified. to AEP-NRC-2024-29 Page 8

Each penetration-to-head interface (annulus) is closely scrutinized in accordance with guidance in Reference 2, and a determination is made as to whether there are any boric acid deposits, corrosion, or discoloration on or close to the annulus. Small and newly formed leak paths may result in a minimal amount of boron deposit buildup. An active leak will produce a localized buildup of light-colored boric acid crystals. If leakage has occurred over several outages the deposits can resemble popcorn, stalagmites, or spaghetti. In some cases, the "spaghetti" has even formed into a ball. In general, buildup tends to be seen most frequently on the downhill side of a penetration because the leak runs downhill and the boron is deposited as the water evaporates.

The characteristics of the identified relevant conditions of boric acid deposits, discoloration and entrained residues are not consistent with industry examples of deposits from penetration leaks. The l&M as-found examination identified localized deposits running down the nozzles from above and onto the annulus and then the vessel head. This is indicative that the leakage originated elsewhere. l&M has identified the source of leakage that caused the identified conditions as the TECSA seals. The corrective actions to address this leakage source are discussed below. The penetration is characterized as having relevant conditions because the guidance in RIS 2018-06 directs this action. l&M initiated additional corrective actions to prevent boric acid leaking onto the head if the TECSAs leak in the future.

l&M took corrective actions to resolve leakage from TECSA seals in response to conditions identified during previous outages. A discussion of the corrective actions to prevent the reoccurrence of the leakage onto the RVCH is below.

Actions Already Taken

• Preclude future TECSA leakage

o In 2018, a new parts vendor was utilized to provide the TECSA components. l&M identified issues with existing part quality and vendor instruction. The new parts, which were receipt inspected against quality requirements, were installed beginning in U1 C30. These parts were used for the U1C31, U2C26, U2C27, and U2C28 outages.

o Beginning in U1C29, the methodology for the TECSA inspection was changed to allow for performing the TECSA inspections when the RVCH is removed and on the stand. This allows greater access during the inspection and cleaning of the seating surfaces. Following implementation of this new inspection method leakage was not observed on the subsequent BMV for Unit 1 in U1 C31, so the corrective actions were effective. These practices were used for the U2C26 and U2C27 outages and were moderately effective in that different and fewer nozzles were impacted by leakage originating from TECSA seals.

• If leakage is observed, the following additional actions will be taken:

o l&M monitors the TECSA seal locations, as well as the RVCH, during start-up with specific targeted walk downs performed at 300 pounds (#), 1000#,

and Normal Operating Pressure and Normal Operating Temperature.

During these walk downs personnel verify no leakage, install leakage to AEP-NRC-2024-29 Page9

collection devices ( special blankets and cloths that are wrapped around the seals) to preclude any leakage from getting to the RVCH, record any observed leakage (including locations and quantity), obtain photographs of the leakage, if possible, clean any accessible leakage and deposits, and report leakage to the Outage Control Center and Operations. These conditions are also be documented in the Corrective Action Program.

Additional Corrective Actions

• Per vender recommendation the TECSA seal assembly procedure was revised to include steps to verify gasket compression and raise the allowed torque value to ensure proper gasket compression can be achieved.

• Develop and implement a modification during U1 C33 and U2C29 to replace the TECSA seals with a different style of seal that has demonstrated low to no leakage during reactor startup.

This modification could not be implemented during U2C28 due to the analysis and qualification for parts not completing in time for implementation.

• RVCH Structural Integrity

l&M replaced the Unit 2 RVCH in 2007. The replacement RVCH is constructed with PWSCC resistant materials with an outer surface of SA-508 Grade 3 manganese molybdenum low alloy steel. The CROM and ICI nozzles are Alloy 690. Evaluations were performed and documented in MRP-375 (Reference 9) to demonstrate the acceptability of extending the inspection intervals for ASME Code Case N-729-1, item B4.40 components. Based on plant service experience, factor of improvement (FOi) studies using laboratory data, deterministic study results, and probabilistic study results, MRP-375 supported extended inspection intervals. This information documents the structural suitability of the RVCH for extended periods of time.

Per MRP-375, much of the laboratory data indicated an FOi of 100 for Alloy 690/52/152 versus Alloy 600/182/82 (for equivalent temperature and stress conditions) in terms of crack growth rates. In addition, laboratory and plant data demonstrate an FOi in excess of 20 in terms of the time to PWSCC initiation. This reduced susceptibility to PWSCC initiation and growth supports elimination of all volumetric examinations throughout the plant service period, and by extension, supports not performing volumetric examinations during U2C28. The next volumetric examination for Unit 2 RVCH is scheduled for U2C29 (Fall of 2025).

Deterministic calculations demonstrate that the alternative volumetric re-examination schedule is sufficient to detect any PWSCC before it could develop into a safety significant circumferential flaw that approaches the large size (i.e., more than 300 degrees of circumferential extent) necessary to produce a nozzle ejection. The deterministic calculations also demonstrate that any base metal PWSCC would likely be detected prior to a through-wall flaw occurring. Probabilistic calculations based on a Monte Carlo simulation model of the PWSCC process, including PWSCC initiation, crack growth, and flaw detection via ultrasonic testing, show a substantially reduced effect on nuclear safety compared to a RVCH with Alloy 600 nozzles examined per current requirements.

As documented in MRP-375, the resistance of Alloy 690 and corresponding weld metals Alloy 52 and 152 is demonstrated by the lack of PWSCC indications reported in these materials, in up to 24 consecutive years of service for thousands of Alloy 690 steam generator tubes, and more than 22 consecutive years of service for thick-wall and thin-wall Alloy 690 applications.

This operating experience includes service at pressurizer and hot-leg temperatures higher than those on the RCS. to AEP-NRC-2024-29 Page 10

Based on the above information, the RVCH nozzles and attachment welds are less susceptible to the initiation and growth of PWSCC flaws. Due to reduced susceptibility of Alloy 690 and low growth rates of PWSCC, MRP-375 provides reasonable assurance of low probability for nozzle leaks and that the structural integrity of the RVCH nozzles will be maintained over the next operating cycle.

Additionally, following water and steam cleaning, l&M performed a BMV examination. The subsequent BMV showed that no significant degradation exists on the affected areas of the RVCH. The proposed alternative to perform a subsequent BMV examination of the RVCH in accordance with the current version of Code Case N-729 during the next refueling outage will either confirm that no leakage from the reactor vessel pressure boundary is occurring or will identify any possible leakage before it could challenge the structural integrity of the RVCH.

Additional corrective actions taken to prevent future leakage from the TECSA seals are expected to minimize potential degradation during the next cycle of operation.

• CNP Technical Specifications (TS) require monitoring of operational leakage.

The CNP leakage detection program serves two distinct purposes related to this relief request.

The first purpose is to support the conclusion that a RVCH nozzle leak does not currently exist. The operational leakage for CNP was reviewed during the last cycle. The unidentified leakage over this entire period was between O and 0.02 gallons per minute (gpm). There was no increase in RCS leakage that would be indicative of a through-wall leak of the RVCH nozzles.

Second, the CNP leakage detection program would detect increases in operational leakage consistent with the formation of nozzle leaks during the cycle. Increased operational leakage would be identified and addressed prior to challenging the structural integrity of the RVCH.

CNP has operational RCS leakage requirements established in TS 3.4.13. The Pressurized Water Reactor Owners Group developed WCAP-16465-NP (Reference 10) to provide standardized action levels and response guidelines that address increasing unidentified RCS leakage less than TS limits. CNP has adopted these industry standard administrative requirements which create three-tiered action levels. Each tier is described in detail below.

Tier 1 Action Level

1. One seven day rolling average of daily Unidentified RCS leak rate greater than or equal to 0.1 gpm.

2. Nine consecutive daily Unidentified RCS leak rates greater than baseline mean µ.

Tier One Action Guidelines if any Tier One Action Level is exceeded,

• Confirm dates, times and data.

• Evaluate trend of affected parameter (Pressurizer Level, Volume Control Tank Level, Tave, and Pressurizer Pressure).

• Evaluate trend of associated Tier One triggers.

• Run confirmatory leak rate calculation with different times (Confirmatory leak rate calculation cannot overlap the initial calculation).

• Check for abnormal trends for other leakage indicators (Containment Radiation Monitors, Dew Point, and Containment Sump Level, etc). to AEP-NRC-2024-29 Page 11

If initial indication is confirmed, then perform the following:

• Increase frequency of leakage testing.

• Perform increased frequency sampling.

• Initiate an Action.

Tier 2 Action Level

1. Two consecutive daily Unidentified RCS leak rates greater than or equal to 0.15 gpm.

2. Two of three consecutive daily Unidentified RCS leak rates greater than or equal to

µ+2o.

Tier Two Action Guidelines if any Tier Two Action Level is exceeded,

• Perform Tier One response.

• Commence a leak investigation:

o Review recent plant evolutions to determine any "suspect" source( s ).

o Evaluate changes in other leakage detection indications.

o Initiate outside containment walk-downs of the chemical and volume control system (CVCS), safety injection (SI), and residual heat removal (RHR) systems.

• Identify the source of the increase in leakage:

o Check any components or flow paths recently changed or placed in service, shutdown, vented, drained, filled, etc.

o Check any maintenance activity that may have resulted in decreasing or increasing leakage.

o Check any filters recently alternated or changed. Inspect filter housing for gasket leakage. Check seal injection filters and reactor coolant filter for signs of leakage.

Tier 3 Action Level

1. One daily Unidentified RCS leak rate greater than or equal to 0.3 gpm.

2. One daily Unidentified RCS leak rate greater than or equal to µ+3o.

Tier Three Action Guidelines if any Tier Three Action Level is exceeded,

• Perform Tier One and Tier Two responses.

• If the increased leak rate is indicated inside containment, then perform the following:

o Begin planning for a containment entry while carrying out other actions. Obtain proper approval for containment entry.

o Obtain samples from the Containment Sump, Reactor Cavity and Pipe Tunnel (annulus) sumps (during respective pump out) and analyze for activity, a larger than expected boric acid concentration and other unexpected chemicals.

o Evaluate other systems for indications of leakage (Component Cooling Water, Service Water, etc.).

o Obtain a containment atmosphere sample for indications of RCS leakage.

• Identify source of the leak.

• Quantify the leakage. to AEP-NRC-2024-29 Page 12

• Initiate plan to correct the leak.

• Monitor containment airborne radiation levels as well as area radiation monitors. Sample the containment atmosphere for indications of RCS leakage.

• Monitor other containment parameters (temperature, pressure, dew point, etc.}.

• Monitor RTime screen "Quick Leak Rate" for any potential leakage propagation.

• If the leak source is found and isolated or stopped, then re-perform RCS leak rate calculation.

Summary

Based on RIS 2018-06, leakage from the RVCH penetrations cannot be completely excluded.

Evaluation of the as-found condition of the RVCH supports the conclusion that it is unlikely that the identified relevant conditions are the result of RVCH penetration leakage. The assessments that have been performed, including evaluation of the form and location of the boric acid deposits, discoloration and entrained residues, support the conclusion that the identified conditions are the result of leaking TECSA seals on the RVCH.

Steam cleaning of the affected areas of the RVCH has been performed in order to remove existing relevant conditions. Additional corrective actions have been taken intended to eliminate or mitigate leakage from the TECSA seals. These actions will provide a baseline for performing visual examinations in the next outage. The substantial margins in structural integrity for the RVCH penetrations provide a high level of confidence that CNP Unit 2 can be operated safely through the next cycle of operation. Therefore, the above evaluations and proposed alternative visual examination during the next outage provide a high level of confidence in continued integrity of the RVCH. The additional personnel exposure and limited availability of personnel to perform the additional volumetric and surface examinations represent a hardship without a compensating increase in the level of quality and safety.

6. Duration of Proposed Alternative

The proposed alternative will be utilized until the end of refueling outage U2C29.

7. Precedent

This request is comparable to that submitted for the CNP U1 C30 (Reference 8), U2C26 (Reference 4}

and U2C27 (Reference 12) refueling outages, approved by the NRC via References 5, 11 and 13.

Similar to Unit 1 and Unit 2 previous outages, l&M is requesting relief from the specified requirements of Code Case N-729-6, Paragraph -3142.2, pursuant to 10 CFR 50.55a(z)(2} due to relevant conditions on the RVCH identified during inspections for Unit 2. CNP evaluated the RVCH from both Unit 1 and Unit 2 and determined that the indications were from sources other than nozzle leakage, and the subsequent visual examinations performed for those penetrations with relevant conditions revealed no further relevant conditions.

Both CNP Units 1 and 2 have replacement heads constructed from PWSCC resistant Alloy 690 materials. l&M utilized the same process that was used for both units to determine that the suspected source of the leakage is not RVCH nozzle leakage. l&M is proposing the same Code Case alternative that was accepted for CNP Units 1 and 2 (References 5, 11 and 13}. to AEP-NRC-2024-29 Page 13

8. References

1. ASME Boiler and Pressure Vessel Code Case N-729-6, "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration WeldsSection XI, Division 1," dated March 2016.

2. Boric Acid Corrosion Guidebook, "Managing Boric Acid Corrosion Issues at PWR Power Stations," Revision 2, dated July 2012.

3. NRC Regulatory Issue Summary (RIS) 2018-06, "Clarification of the Requirements for Reactor Pressure Vessel Upper Head Bare Metal Visual Examinations," dated December 10, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No.

ML18178A137).

4. Indiana Michigan Power Company letter, "Request for Relief related to American Society of Mechanical Engineers (ASME) Code Case N-729-6 Supplemental Examination Requirements, ISIR-5-05," dated May 4, 2021 (ADAMS Accession No. ML21124A237).

5. NRC Letter to Indiana Michigan Power, "Donald C. Cook Nuclear Plant, Unit 2 - Relief Request ISIR-5-05 Related to ASME Code Case N-729-6 Supplemental Examination Requirements of Reactor Vessel Closure Head Penetration Nozzles (EPID L-2021-LLR-0033

[COVID-19))," dated May 12, 2021 (ADAMS Accession No. ML21130A008).

6. MRP-384, "Materials Reliability Program: Guideline for Nondestructive Examination of Reactor Vessel Upper Head Penetrations," dated December 2019.

7. NEI 03-08, "Guideline For The Management Of Materials Issues," dated October 22, 2020.

8. Indiana Michigan Power Company letter, "Request for Relief related to American Society of Mechanical Engineers (ASME) Code Case N-729-6 Supplemental Examination Requirements, ISIR-5-04," dated October 5, 2020 (ADAMS Accession No. ML20279A713).

9. MRP-375, "Material Reliability Program: Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles," dated February 2014.

10. WCAP-16465-NP, "Pressurized Water Reactor Owners Group Standard RCS Leakage Action Levels and Response Guidelines for Pressurized Water Reactors," Revision 0, dated September 2006.

11. NRC Letter to Indiana Michigan Power Company, "Donald C. Cook Nuclear Plant, Unit 1 -

Relief Request ISIR-5-04 Related to ASME Code Case N-729-6 Supplemental Examination Requirements of Reactor Vessel Closure Head Penetration Nozzles (EPID L-2020-LLR-0133

[COVID-19))," dated February 12, 2021 (ADAMS Accession No. ML21034A155).

12. Indiana Michigan Power Company letter, "Request for Relief related to American Society of Mechanical Engineers (ASME) Code Case N-729-6 Supplemental Examination Requirements, ISIR-5-06," dated October 24, 2022 (ADAMS Accession No. ML22297A211)

13. NRC Letter to Indiana Michigan Power Company, "Donald C. Cook Nuclear Plant, Unit 2 -

Relief Request ISIR-5-06 Related to ASME Code Case N-729-6 Supplemental Examination Requirements of Reactor Vessel Closure Head Penetration Nozzles (EPID L-2022-LLR-0073" dated January 4, 2023 (ADAMS Accession No. ML22363A562) to AEP-NRC-2024-29 Page 14

Figure 1:

Reactor Vessel Closure Head Penetration with Identified Relevant Conditions

U2 REACTOR VESSEL CLOSURE HEAD to AEP-NRC-2024-29 Page 15

Attachment 1 to Enclosure 1 to AEP-NRC-2024-29 Relevant Condition Discussion

GENERAL (Reference Indiana Michigan Power Company Drawing DC-13745 at the end of this attachment}. The relevant conditions reported during the as-found Visual Examination of the Reactor Vessel Closure Head (RVCH) in refueling outage U2C28 were characterized as boric acid deposits, discoloration and entrained residues.

The penetration is located around the RVCH perimeter near the periphery at approximately 70°. The insulation package installed above the RVCH is comprised of a series of metal panels with butted seam joints and cut outs for each of the 53 Control Rod Drive Mechanism (CROM) and 5 In-Core Instrument (ICI) tubes with thermocouple sealing assemblies (TECSA) on top of them. The seams are aligned with straight courses of CROM and ICI tubes. The seams and tube cut outs in the insulation panels are the pathways for the Reactor Coolant System water that spilled from the TECSAs. As the spillage reaches the RVCH, it drops onto and flows down the surface around the tube. The five ICI tubes are peripherally located and are lower on the concavity of the RVCH. Therefore, TECSA leakage is localized to the associated ICI penetration and the adjacent surfaces lower on the RVCH.

All boric acid deposits, discoloration and entrained residues were successfully removed via steam cleaning with no relevant conditions identified after cleaning and no degradation noted around the penetration.

PENETRATION 76 Relevant Condition - boric acid deposits, discoloration and entrained residues RVCH Location - - Azimuth - 70° The location of the boric acid deposits, discoloration and entrained residue is all around the penetration. See Attachment 2 to Enclosure 1 for photographs of the penetration as-found and as-left in U2C27 and as found and as-left in U2C28. to AEP-NRC-2024-29 Page 16

Excerpt from DC-13745:

VUEDEDESSUS JOPYIEW

p A + to AEP-NRC-2024-29 Page 17

Attachment 2 to Enclosure 1 to AEP-NRC-2024-29 Supporting Reactor Vessel Closure Head (RVCH) Photographs of Penetration 76

The photographs labeled U2C27 As-Found and As-Left were taken of Penetration 76 during the previous outage bare metal visual (BMV) examination when no relevant conditions were found. After the BMV, the thermocouple sealing assembly (TECSA) atop Penetration 76 was installed. Several Containment leak inspections were performed, and no leakage was identified.

At the beginning of U2C28 in March 2024, the subsequent BMV was performed of the RVCH. The photographs labeled as U2C28 As-Found were taken of the penetration after light cleaning with compressed air during the BMV. The relevant conditions were identified at penetration 76 during the as-found visual examination.

After the BMV was completed, cleaning with steam was performed on penetration 76 and the adjacent areas where boric acid residue, discoloration and entrained residues were identified. The photographs labeled U2C28 As-Left were taken during the subsequent BMV of the penetration after cleaning. The photographs show no degradation of the RVCH and penetration.

Additional photographs are included that show the path of the TE CSA leakage from the TE CSA to the RVCH surface. The leakage at Penetration 76 originated at the TECSA, dripped down the tube and off the insulation onto the surface of the RVCH, with discolored leakage around and below the penetration. to AEP-NRC-2024-29 Page 18

As-Found Photographs of Penetration 76 during U2C27 to AEP-NRC-2024-29 Page 19

As-Found Photographs of Penetration 76 during U2C27 to AEP-NRC-2024-29 Page 20

As-Found Photographs of Penetration 76 during U2C27 to AEP-NRC-2024-29 Page 21

As-Found Photographs of Penetration 76 during U2C27 to AEP-NRC-2024-29 Page 22

As-Left Photographs of Penetration 76 during U2C27 to AEP-NRC-2024-29 Page 23

As-Left Photographs of Penetration 76 during U2C27

l\\fm(f)}Jf}lilDR

-OW) to AEP-NRC-2024-29 Page 24

As-Left Photographs of Penetration 76 during U2C27 to AEP-NRC-2024-29 Page 25

As-Left Photographs of Penetration 76 during U2C27 to AEP-NRC-2024-29 Page 26

As-Found Photographs of Penetration 76 during U2C28 to AEP-NRC-2024-29 Page 27

As-Found Photographs of Penetration 76 during U2C28 to AEP-NRC-2024-29 Page 28

As-Found Photographs of Penetration 76 during U2C28 to AEP-NRC-2024-29 Page 29

As-Found Photographs of Penetration 76 during U2C28 to AEP-NRC-2024-29 Page 30

As-Left Photographs of Penetration 76 during U2C28 to AEP-NRC-2024-29 Page 31

As-Left Photographs of Penetration 76 during U2C28 to AEP-NRC-2024-29 Page 32

As-Left Photographs of Penetration 76 during U2C28 to AEP-NRC-2024 -29 Page 33

Additional Photographs of ICI Tube at Penetration 76 during U2C28

Above: Evidence of TE CSA leakage dripping down the ICI Tube. to AEP-NRC-2024-29 Page 34

Additional photographs of ICI Tube at Penetration 76 during U2C28

Above: Evidence of TE CSA leakage dripping down the ICI Tube. to AEP-NRC-2024-29 Page 35

Additional photographs of ICI Tube at Penetration 76 during U2C28

Above: Evidence of TE CSA leakage dripping down the ICI Tube. to AEP-NRC-2024-29 Page 36

Additional photographs of ICI Tube at Penetration 76 during U2C28

Above: Evidence of TE CSA leakage dripping down the ICI Tube.

Enclosure 2 to AEP-NRC-2024-29 Regulatory Commitment for Reactor Vessel Closure Head (RVCH) Inspection

REGULATORY COMMITMENT

The following table identifies an action committed to by Indiana Michigan Power Company (l&M) in this document. Any other actions discussed in this submittal represent intended or planned actions by l&M.

They are described to the U. S. Nuclear Regulatory Commission (NRC) for the NRC's information and are not regulatory commitments. All commitments discussed in this table are one-time commitments.

Commitment Scheduled Completion Date (if applicable):

Indiana Michigan Power Company will perform a bare metal visual inspection of the Donald C. Cook Nuclear Plant (CNP) Unit 2 Reactor Vessel Closure Head in the next refueling outage in Unit 2 Cycle 29 Refueling accordance with the latest revision of Code Case N-729 endorsed Outage in 1 O CFR 50.55a. This commitment is related to the CNP Unit 2 relief request ISIR-5-07.