AEP-NRC-2022-61, Request for Relief Related to American Society of Mechanical Engineers (ASME) Code Case N-729-6 Supplemental Examination Requirements, ISIR-5-06

Request for Relief Related to American Society of Mechanical Engineers (ASME) Code Case N-729-6 Supplemental Examination Requirements, ISIR-5-06

ML22297A211
Person / Time
Site:	Cook
Issue date:	10/24/2022
From:	Ferneau K American Electric Power, Indiana Michigan Power Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
AEP-NRC-2022-61
Download: ML22297A211 (1)

Text

m INOIANA Indiana Michigan Power MICHIGAN Cook Nuclear Plant One Cook Place POWER* Bridgman, Ml 49106 A unit of American Electric Power lndianaMichiganPower.com October 24, 2022 AEP-NRC-2022-61 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-5-06 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, U2C28.

As part of the current CNP Unit 2 refueling outage, U2C27, 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-5-05 that was accepted on May 12, 2021 (ML21130A008). During this subsequent examination, it was discovered that relevant conditions of boric acid deposits, discoloration and entrained residues exist at two different nozzles 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 similar relief request for CNP Unit 1 on October 5, 2020 (ML20279A713). The safety evaluation was received on February 12, 2021 (ML21034A155).

Approval of the proposed relief is requested prior to entry into Mode 2, which is scheduled to occur on or about November 6, 2022.

U.S. Nuclear Regulatory Commission AEP-NRC-2022-61 Page 2 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,

~~

Kelly Ferneau Site Vice President SJM/kmh

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.

U.S. Nuclear Regulatory Commission AEP-NRC-2022-61 Page 3 c: R. J. Ancona - MPSC EGLE - RMD/RPS J. B. Giessner - NRC Region Ill M. G. Menze - AEP Ft. Wayne NRC Resident Inspector R. M. Sistevaris - AEP Ft. Wayne S. P. Wall - NRC, Washington D.C.

A. J. Williamson - AEP Ft. Wayne

Enclosure 1 to AEP-NRC-2022-61 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 Head with nozzles and partial-penetration welds of Components I Primary Water Stress Corrosion Cracking Numbers:

(PWSCC) resistant materials Code Class: ASME Boiler and Pressure Vessel Code (Code),

Class 1 ASME Section XI 2013 Edition and no addenda ASME Section XI, Division 1, Code Case N-729-6, Alternative Examination Requirements for

References:

Pressurized Water Reactor (PWR) Reactor Vessel Upper Heads with Nozzles Having Pressure-retaining Partial-Penetration Welds,Section XI, Division 1 Examination Table IWB-2500, Category 8-P Category:

Table 1 of ASME Code Case N-729-6 Item Number( s ): 84.30 Pressure retaining components Reactor Vessel Closure Head (RVCH) with nozzles and partial-penetration welds of primary

Description:

water stress corrosion cracking PWSCC-resistant materials Penetrations - 77, 78 Unit / Inspection Donald C. Cook Nuclear Plant, Unit 2 (CNP)/ 5th Interval Ten-Year In-service Inspection (ISi) Interval Applicability: (March 1, 2020 to February 28, 2030)

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 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).

to AEP-NRC-2022-61 Page2 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 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.

to AEP-NRC-2022-61 Page 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 Fig. 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 I 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 (U2C27) in accordance with ASME Code Case N-729-6, Table 1. During U2C27, 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 lightly adhered accumulated residues as were seen during the U2C26 VE. Boric acid deposits, discoloration and entrained residues that remained after light cleaning were determined to be relevant conditions in the as-found state. Two penetrations, specifically penetrations 77 and 78, were determined to have relevant conditions pursuant to Code Case N-729-6, Paragraph -3141(c).

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

Samples from the relevant conditions were obtained in an attempt to better categorize the source.

to AEP-NRC-2022-61 Page 4 The deposits on the Unit 2 RVCH were analyzed by Chemistry personnel, and their age was determined to be older than one year. 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 noted discoloration, staining, and debris but no evidence of degradation or active leakage. The previous inspection was performed after the issuance of RIS 2018-06, and the noted conditions required a relief request, which was approved in May 2021 (References 4 and 5). After the previous inspection, all of the nozzles having relevant conditions were cleaned, and an as-left inspection was performed. The as-found condition of the Unit 2 RVCH in U2C27 is similar to the condition identified during the as-left inspection, except for two previously acceptable nozzles having relevant conditions. It is noted that the two outages do not have any nozzles with relevant conditions in common with each other, indicating that all of the nozzles that previously had relevant conditions have been proven to not be leaking.

The relevant conditions on the two specific nozzles in the scope of this relief request require consideration, per RIS 2018-06, that some or all of the relevant conditions possibly came from the nozzles. 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 water, CO2. and a subsequent bare metal visual (BMV) examination were completed in U2C27, and all relevant conditions were removed. No degradation was noted by the subsequent BMV examination. The water and CO2 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 all 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 nozzles are 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.

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 appropriately 16 to AEP-NRC-2022-61 Page 5 qualified personnel to the site on an emergent basis. 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 U2C27. The additional dose for this work described above is estimated to be approximately 754.0 mRem. Providing access to the underside of the RVCH is an infrequently performed evolution that has not been performed at l&M in fifteen years. While manageable, additional risk is incurred in setting up an atypical configuration where the RVCH rests levelly at a higher elevation on 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 nine days. This would add extra duration to the U2C27 outage, and all remaining outage activities associated with reactor assembly and startup would be delayed by at least nine 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-2022-61 Page 6

• 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 (U2C28) 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 water and CO2 cleaning of affected areas of the RVCH head 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 conclude that the relevant conditions are likely 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. During the current U2 refueling outage, U2C27, two penetrations with relevant conditions were identified and are discussed in this request. The corrective actions still prove to be moderately effective as the number of relevant conditions is significantly lower. 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 a subsequent BMV for Reference 4, the corrective actions were effective in that all penetrations that did have leakage in U2C26 no longer have relevant conditions in U2C27, with two new penetrations currently having relevant conditions.

to AEP-NRC-2022-61 Page 7 Based on a review of the corrective action program and the current leakage paths, TECSA seal leakage contributed solely to the U2C27 relevant conditions. Based on this repeat type of leakage, l&M has taken additional corrective action to prevent reoccurrence of TE CSA seal leakage contacting the RVCH in the future, updating the procedure to include necessary steps to verify compression of the gaskets for all TECSA seals, as well as increasing the torque used on the seals based on vendor input. 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.

• 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. 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.

l&M posits that the relevant conditions discovered in U2C27 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, 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 indications has been identified.

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, or drops from above directly onto the vessel head. This is indicative that the leakage originated elsewhere. l&M has identified the source of leakage that caused the identified conditions, the TECSA seals. The corrective actions to address this leakage source are discussed below.

to AEP-NRC-2022-61 Page 8 l&M obtained samples of the residue and performed chemical composition analysis of the deposits. The residue was determined to be at least one year old based on the chemical analysis, indicating that this leakage is not due to recent or fresh through-wall leakage. These two penetrations are listed as 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 head is below.

Actions Already Taken o Preclude future TECSA leakage

• 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 U1 C31, U2C26, and U2C27 outages.

• Beginning in U1 C29, 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 U 1C31, so the corrective actions were effective. These practices were used for the U2C26 outage and were effective to a degree.

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

• 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, record any observed leakage (including locations and quantity), obtain photos of the leakage, if possible, and report leakage to the Outage Control Center and Operations. These conditions are also be documented in the Corrective Action Program.

• If environmental conditions permit, l&M personnel will clean accessible boric acid deposits identified during the specified walk downs utilizing wetted lint-free cloths.

to AEP-NRC-2022-61 Page 9 Additional Corrective Actions Initiated for U2C27 o Per vender recommendation the TECSA seal assembly procedure is being revised to include steps to verify gasket compression and raise the allowed torque value to ensure proper gasket compression can be achieved.

• 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 CRDM 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 84.40 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 U2C27.

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.

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 the low probability of current nozzle leaks and that the structural integrity of the RVCH nozzles will be maintained over the next operating cycle.

to AEP-NRC-2022-61 Page 10 Additionally, following water and CO2 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 enable either confirmation 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.

• The CNP Technical Specifications (TS) require monitoring of operational leakage 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 0.03 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 (Pzr Level, VCT Level, Tave, and Pzr 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).

If initial indication is confirmed, then perform the following:

• Increase frequency of leakage testing.

• Perform increased frequency sampling.

• Initiate an Action.

to AEP-NRC-2022-61 Page 11 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.

• 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.

to AEP-NRC-2022-61 Page 12 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.

Water and CO2 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. Given that, 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 U2C28.

7. Precedent This request is comparable to that submitted for the CNP U1 C30 (Reference 8) and U2C26 refueling outages (Reference 4), approved by the NRC via Reference 5 and 11.

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 and 11 ).

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).

to AEP-NRC-2022-61 Page 13

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 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).

to AEP-NRC-2022-61 Page 14 Figure 1:

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Attachment 1 to Enclosure 1 to AEP-NRC-2022-61 Relevant Condition Discussion GENERAL (Reference Indiana Michigan Power Company Drawing DC-13745 at the end of this attachment). The two relevant conditions reported during the as-found Visual Examination of the Reactor Vessel Closure Head (RVCH) were characterized as boric acid deposits, discoloration and entrained residues.

Both penetrations are located around the RVCH perimeter near the periphery in two different quadrants, from 135 degree (0 ) to 270°. 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 two tubes. 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 carbon dioxide cleaning with no relevant conditions identified after cleaning and no degradation noted around these penetrations.

PENETRATION 77 Relevant Condition - boric acid deposits, discoloration and entrained residues RVCH Location - - Azimuth 160° The location of the boric acid deposits, discoloration and entrained residues is all around the penetration. See Attachment 2 to Enclosure 1 for photos of the penetration as left U2C26, as found U2C27 and as-left U2C27.

PENETRATION 78 Relevant Condition - boric acid deposits, discoloration and entrained residues RVCH Location - - Azimuth 250° The location of the boric acid deposits, discoloration and entrained residues is all around the penetration. See Attachment 3 to Enclosure 1 for photos of the penetration as left U2C26, as found U2C27 and as left U2C27.

to Enclosure 1 to AEP-NRC-2022-61 Page 2 Excer t from DC-137 45:

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Attachment 2 to Enclosure 1 to AEP-NRC-2022-61 Supporting Reactor Vessel Closure Head (RVCH) Photographs of Penetration 77 The photos labeled as U2C26 were taken of Penetration 77 during the previous outage's bare metal visual (BMV) examination when no relevant conditions were found. After the BMV, on May 11, 2021, the thermocouple sealing assembly (TECSA) atop Penetration 77 was installed. On May 14, 2021, containment leak inspections, at 300 pounds per square inch gauge (psig) (when the reactor is started up,) were performed, and leakage in the form of weeping was identified at this penetration, with l&M personnel inspecting the leaking seals three more times as a follow-up action while the penetration leaked. This penetration continued to leak until May 19, 2021, when it was documented in the corrective action program that the seals were dry. At the beginning of U2C27 in October 2022, another BMV was performed of the RVCH . The photos labeled as U2C27 AS LEFT were taken of the penetration after cleaning with compressed air during the BMV. After the BMV was completed, some initial cleaning with hot water was performed on the penetration, and the unlabeled photos show the leak path of the TECSA leakage from the TECSA to the RVCH surface. The photos labeled U2C27 POST CO2 CLEANING were taken during the subsequent BMV of the penetration after cleaning the penetration with CO2, and the photos show no degradation of the RVCH and penetration.

The leakage at Penetration 77 was heavier than 78, with leakage originating at the TECSA, dripping down the tube and off the insulation onto the surface of the RVCH, with discolored leakage all around and below the penetration.

As-Found/As-Left Photos of Penetration 77 during U2C26:

to Enclosure 1 to AEP-NRC-2022-61 Page2 As-Found/As-Left Photos of Penetration 77 during U2C26 (continued):

to Enclosure 1 to AEP-NRC-2022-61 Page 3 As-Found Photos of Penetration 77 during U2C27:

to Enclosure 1 to AEP-NRC-2022-61 Page4 As-Found Photos of Penetration 77 during U2C27 (continued):

Additional photos of CRDM Risers during U2C27:

Above: Evidence ofTECSA leakage dripping down the CROM riser.

to Enclosure 1 to AEP-NRC-2022-61 Page 5 Additional photos of CRDM Risers during U2C27 (continued):

Above: Evidence ofTECSA leakage dripping down the CRDM riser.

to Enclosure 1 to AEP-NRC-2022-61 Page 6 Additional photos of CROM Risers during U2C27 (continued):

Above: Evidence ofTECSA leakage from above, dripping down the riser onto the RVCH.

Above: Evidence ofTECSA leakage from above, dripping down the riser onto the RVCH.

to Enclosure 1 to AEP-NRC-2022-61 Page 7 Photos of Penetration 77 from subsequent BMV after cleaning with CO2 during U2C27:

to Enclosure 1 to AEP-NRC-2022-61 Page 8 Photos of Penetration 77 from subsequent BMV after cleaning with CO2 during U2C27 (continued):

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Attachment 3 to Enclosure 1 to AEP-NRC-2022-61 Supporting Reactor Vessel Closure Head (RVCH) Photographs of Penetration 78 The photos labeled as U2C26 were taken of Penetration 78 during the previous outage's bare metal visual (BMV) examination when no relevant conditions were found. After the BMV, on May 11, 2021, the thermocouple sealing assembly (TECSA) atop Penetration 78 was installed. On May 14, 2021, containment leak inspections, at 300 pounds per square inch gauge (psig) (when the reactor is started up), were performed, and leakage in the form of weeping was identified at this penetration, with l&M personnel inspecting the leaking seals three more times as a follow-up action while the penetration leaked. This penetration continued to leak until May 19, 2021, when it was documented in the corrective action program that the seals were dry. At the beginning of U2C27 in October 2022, another BMV was performed of the RVCH. The photos labeled as U2C27 AS LEFT were taken of the penetration after cleaning with compressed air during the BMV. After the BMV was completed, some initial cleaning with hot water was performed on the penetration, and the unlabeled photos show the leak path of the TECSA leakage from the TECSA to the RVCH surface. The photos labeled U2C27 POST CO2 CLEANING were taken during the subsequent BMV of the penetration after cleaning the penetration with CO2, and the photos show no degradation of the RVCH and penetration.

The leakage at Penetration 78 was lighter than 77, with leakage originating at the TECSA, dripping down the tube and off the insulation onto the surface of the RVCH, with splashes and droplets of leakage all around the penetration.

As-Found/As-Left Photos of Penetration 78 during U2C26:

to Enclosure 1 to AEP-NRC-2022-61 Page 2 As-Found/As-Left Photos of Penetration 78 during U2C26 (continued):

to Enclosure 1 to AEP-NRC-2022-61 Page 3 Photos of Penetration 78 after blowing air on the penetration during U2C27:

to Enclosure 1 to AEP-NRC-2022-61 Page 4 Photos of Penetration 78 after blowing air on the penetration during U2C27 {continued):

to Enclosure 1 to AEP-NRC-2022-61 Page 5 Additional photos of CROM Riser during U2C27:

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Above: Evidence ofTECSA leakage dripping down the CRDM riser.

to Enclosure 1 to AEP-NRC-2022-61 Page6 Additional photos of CROM Riser during U2C27 (continued):

Above: Evidence ofTECSA leakage dripping down the CRDM riser.

Above: Evidence of TECSA leakage dripping down the CRDM riser.

to Enclosure 1 to AEP-NRC-2022-61 Page 7 Additional photos of CROM Riser during U2C27 (continued):

Above: Evidence ofTECSA leakage from above, dripping down the riser onto the RVCH.

Above: Evidence ofTECSA leakage from above, dripping down the riser onto the RVCH.

to Enclosure 1 to AEP-NRC-2022-61 Page 8 Photos of Penetration 78 from subsequent BMV after cleaning with CO2 during U2C27:

to Enclosure 1 to AEP-NRC-2022-61 Page 9 Photos of Penetration 78 from subsequent BMV after cleaning with CO2 during U2C27 (continued):

Enclosure 2 to AEP-NRC-2022-61 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 Unit 2 Cycle 28 Refueling inspection of the Donald C. Cook Nuclear Plant (CNP) Unit 2 Outage Reactor Vessel Closure Head in the next refueling outage in accordance with the latest revision of Code Case N-729 endorsed in 10 CFR 50.55a. This commitment is related to the CNP Unit 2 relief request ISIR-5-06.