AEP-NRC-2020-65, Request for Relief Related to American Society of Mechanical Engineers (ASME) Code Case N-729-6 Supplemental Examination Requirements, ISIR-5-04

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Request for Relief Related to American Society of Mechanical Engineers (ASME) Code Case N-729-6 Supplemental Examination Requirements, ISIR-5-04
ML20279A713
Person / Time
Site: Cook American Electric Power icon.png
Issue date: 10/05/2020
From: Lies Q
Indiana Michigan Power Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2020-65
Download: ML20279A713 (42)
v • d • e


Text

cm INDIANA Indiana Michigan Power MICHIGAN Cook Nuclear Pia nl POWER 0 One Cook Place Bridgman. Ml 49106 A unit ofAmerican, Electric Power lndlanaMichlganPower.com October 5, 20~0 AEP-NRC-2020-65 10 CFR 50.55a Docket No.: 50-315 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-5-04 Pursuant to 10 CFR 50.55a(z)(2), Indiana Michigan Power Company (l&M), the licensee for D.C.

Cook (CNP) Unit 1, requests Nuclear Regulatory Commission approval of the enclosed request for an alternative for CNP Unit 1, based upon the specified Code Case requirements representing a hardship or unusual difficulty without a cqmpensating 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 atthe end of the next refueling outage, U1 C31. *

  • As part of the current CNP Unit 1 refueling outage, U1 C30, the Unit 1 Reactor Vessel Closure Head (RVCH) visual examination was performed In accordance with the requirements in Code Case N-729-6, Table-1. During this examination, it was discovered that relevant conditions of corrosion, boric acid deposits, and discoloration exist on the Unit 1 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 the enclosure 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.

Approval of the proposed relief is requested prior to entry into Mode 2, which is scheduled to occur on or about October 14, 2020.

U.S. Nuclear Regulatory Commission AEP-NRC-2020-65 Page2 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.

?~A-Li Si~,

a. Shane Lies Site Vice President DLW/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-2020-65 Page3 cc:

R. J. Ancona - MPSC EGLE - RMD/RPS J..B. Giessner - NRC Region Ill NRC Resident Inspector

. S. P. Wall - NRC Washington, D.C.

A. J. Williamson - AEP Ft. Wayne, w/o enclosures

Enclosure 1 to AEP-NRC-2020-65 Indiana Michigan Power Company- Donald C. Cook

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 /

Primary Water Stress Corrosion Cracking Numbers:

(PWSCC) resistant materials American Society of Mechanical Engineers Code Class: (ASME) Boller 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,

References:

Alternative Examination Requirement~ for PWR Reactor Vessel Upper Heads with Nozzles Having Pressure-retaining Partial-Penetration Welds,Section XI, Division 1 Examination Table IWB-2500, category B-P Category:

Table 1 of ASME Code Case N-729-6 Item Number{s): B4.30 Pressure retaining components Reactor Vessel Closure Head (RVCH) with nozzles and partial-penetration welds of primary w~ter stress corrosion cracking PWSCC-resistant

Description:

materials Penetrations - 1, 7, 8, 10, 22, 23, 39, 46, 59, 62, 64, 70, 74, 75, 76, 77, 79 Unit / Inspection D.C. Cook, Unit 1 (CNP)/ 5th 10-Year In-service Interval Inspection (ISi) Interval (March 1, 2020 to Applicability: February 28, 2030)

2. Applicable Code Edition and Addenda

The Fifth 10-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 to AEP-NRC-2020-65 Page2 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)(il)(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.

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 subse~uent VE of the previously obscured surfaces shall be performed, pr;ior 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.

to AEP-NRC-2020-65 Page 3 *

(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 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/ replacement aciMfy. . - .

4. Reason for Request

Indiana Michigan Power Company (l&M), the licensee for CNP Unit 1, 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 VE of the RVCH nozzle penetrations during the current Unit 1 refueling outage (U1C30) in accordance w_ith ASME Code case N-729-6, Table 1. During U1C30, the RVCH head nozzle penetrations*were examined in the as-found condition using ASME Code Case N-729-6. 29 penetrations and general RVCH surfaces were initially determined to have relevant conditions pursuant to Code Case N-729-6, Paragraph-3141{c). Specifically, Nozzles 1, 8, 10, :18, 22, 39, 46, 51, 52, 53, 54, 56, 57, 59, 60, 63, 64, 66, 67, 68, 69, 70, 72, 73, 74, 75, 76, 77, and 79, as well as general RVCH surfaces were identified as having relevant conditions.

After cleaning using steam and soft nylon brushes, the number of penetrations that continu~d to exhibit relevant conditions was reduced to 17. The general RVCH surfaces continued to exhibit to AEP-NRC-2020-65 Page4 relevant conditions are well. The 17 specific nozzles that are within the scope of this relief request are Nozzles 1, 7, 8, 10, 22, 23, 39, 46, 59, 62, 64, 70, 74, 75, 76, 77, and 79. l&M deemed that these were relevant conditions per Code Case N-729-6 and the supplemental guidance in Regulatory Information Summary (RIS) 2018-06 (Reference 2). It is noted that Penetrations 7, 23, and 62 were not included in the initially identified relevant conditions. The noted discoloration on these three penetrations is thought to have been obscured during the first inspection and made visible following light cleaning.

  • 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 Mechanisms (CROM) nozzle leaks based on Reference 9. A review of previous inspection results showed similar levels of corrosion, boric acid deposits, and discoloration, although the U1 C30 inspection did identify more nozzles with relevant conditi,;,ns. The leakage is understood to be from the in-core instrumentation (ICI) thermocouple sealing assembly (TECSA) and outage worker practices related to head vent piping removal and refueling activities.

Although the sources of leakage provide the likely cause of the conditions, based on the guidance in RIS 2018-06, it could not absolutely be refuted that the relevant conditions identified in the as-found examination were not masking relevant conditions indicative of nozzle leakag~.

The relevant conditions on the 17 specific* nozzles in the scope of this relief request require consideration, per RIS 2018-06, that some or all of the leakage possibly came from the nozzles. Code Case N-729-6, Paragraph -3142.2, requires that nozzles with relevant conditions indicative of possible nozzle leakage 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).

Note that during the public conference call on October 2, 2020, l&M discussed with the Nuclear Regulatory Commission (NRC) the possibility of also requesting relief from Code case N-729-6, Paragraph -3142.1 (b)(2). At that time, l&M was still determining if mobilization of contractors and specialized equipment to perform carbon dioxide (CO2) head cleaning was available and appropriate given the considerations discussed below. The CO2 cleaning is required to meet Code Case N-729-6, Paragraph -3142.1 (b)(2). After the conference call, l&M determined that the required equipment and contractors were available and that this cleaning was appropriate. CO2 cleaning and subsequent bare metal visual (BMV) examination will be completed prior to startup. The C<Ji cleaning is being performed because it provides the best possible baseline surface for future inspections. Removing all residual boric acid also precludes.RVCH degradation from this source during the next operating cycle. l&M concluded that the burden associated with performing CO2 cleaning is justified given the importance of returning the unit to service with a clean reactor head that will support implementation of an effective visual examination during the next outage which, as discussed below, is the alternative proposed in support of this relief request.

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 is currently occurring.

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

Enclosure 1 to AEP-NRC-2020-65 Page5

  • safety. This section discusses the identified hardship, proposed alternative and supporting basis that shows the. supplemental volumetric examination provides a hardship with limited increase in quality ana 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 approximately 18 qualified personnel to the site on an emergent basis. The supplemental volumetric examinations require access to the underside of the highly contaminated RVCH which would expose personnel to elevated dose rates not previously planned for this refueling outage. The additional dose for this work described above Is estimated to be approximately 4.1 Rem.

Also, it is estimated that mobilization of contract personnel and equipment, and completion of the required supplemental examinations would take approximately three w~eks, depending upon availability of resources. This would add extra duration to the U1 C30 outage, which would require the large number of individuals associated with the outage to be put into an undesirable situation due to the current COVID-19 Public Health Emergency. l&M has taken extensive measures in the current U1 C30 outage to limit the number of individuals inside the protected area to prevent the spread of COVID-19. All non-essential work was deferred to reduce the number of required interactions between l&M and contract employees.. l&M has mandated remote work locations and strict mask and social distancing policies, and has also minimized the nu.mber of contractors supporting the outage. Adding an additional 18 supplemental workers and the necessary l&M oversight would increase the risk of spreading COVID-19 due to increased time on site and close working conditions required for performing examinations. Additionally, all remaining outage activities associated with reactor assembly and startup, would also be delayed by three weeks depending on the availability of equipment and personnel to *perform the examinations. This would require retention of l&M and contract employees thus extending the duration of increased outage personnel onsite and potential risk of COVID-19 Infection.

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 j RVCH in the next refueling outage (U 1C31) in accordance with the latest revision of Code Case N-729 endorsed in 1O.

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 is taking the action to perform CO2 ~leaning of the RVCH head and post cleaning BMV examinations. These actions in conjunction with the corrective actions discussed below to address the identified sources of leakage are intended to provide a high level of confidence that l&M will have the necessary baseline conditions to support performing effective visual examinations 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 leakage. The volumetric examination would confirm l&M's conclusion to AEP-NRC-2020-65 Page6 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.

l&M has also taken corrective action to prevent reoccurrence of the likely sources of leakage from contacting the RVCH in the future. Based on this, the supplemental volumetric examination does 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 1 RVCH were examined in refueling outage u1c21*

(Spring 2016), consistent with Code case N-729-1. Although fewer nozzles with relevant conditions were identified, the previous examinations showed similar levels of debris and discoloration. The three relevant conditions identified in the U1 C27 inspection were all

  • dispositioned as "discoloration on the head by sources from above." A review of the previous examinations also showed no evidence that would be indicative of a through wall leak or a

. degraded condition, based on the guidance provided in Reference 9.

For the U1 C30 inspection, Code Case N-729-6 and the supplemental guidance In RIS 2018-06 were used to determine if the identified discoloration and boric acid deposits represent relevant conditions. RIS 2018-06 which has *provided clarification as to expectations for classifying relevant conditions was not yet issued to the industry at the time of the U1 C27 inspection. Based on the guidance in RIS 2018-06, the noted relevant conditions that could not be removed by light cleaning can no longer be dispositioned regardless of the identified

. source.

l&M posits that the relevant conditions discovered in U 1C30 do not represent a degraded condition of.the RVCH. l&M assessed the identified relevant conditions and, for the rea~ons stated in this relief request, concluded that the relevant conditions are most likely the result of sources other than RVCH head penetration leakage. However, the possibility of RVCH penetration leakage cannot be completely refuted. *

  • The pattern of residue is not consistent with RVCH nozzle leaks and two sources of leakage causing the relevant indications have been identified.

Each _penetration-to-head Interface (annulus) is closely scrutinized in accordance with guidance in Reference 9 and a determination is made as to* whether there are any boric acid -

deposits 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 conditions/deposits are not consistent with Industry examples of deposits from penetration leaks. l&M as-found inspection identified broad, to AEP-NRC-2020-65 Page7 diffuse deposits near the annulus or on the vessel head, which were sometimes located between penetrations. This is indicative that the leaks originated elsewhere. l&M has identified two sources of leakage that likely caused the identified conditions, TECSA seals and vent piping. These leakage sources and corrective actions to address them are discussed below.. Deposits of this type were found to cover portions of the head like a tightly-adhering coating or taking the form of loose, granular material mainly on the uphill side of the penetration. The as-found deposits exhibited some discoloration due to age.

It should be noted that l&M did not perform chemical composition analysis of the deposits.

The likely source of the leakage is Reactor Coolant System (RCS) fluid from the reactor head vent piping flange connection and the TECSA seals deposited over several cycles. Therefore, chemical analysis would.not be expected to provide meaningful results.

l&M attributes the leakage on the head to two identified sources.

o TECSA seal- leakage The thermocouple columns are a part of Upper Internal Equipment and allow the core thermocouples to pass through the RVCH. The TECSA mounts on the thermocouple nozzle adapter on the RVCH. The TECSA function is to maintain the RCS pressure boundary and support the thermocouple column. The TECSA have been prone to leak at low pressures. l&M has documented TECSA leakage in the corrective action program on previous occasions. Discussions with the seal vendor indicated that "transient weepage" can to occur from the TECSA at lower pressures and the leakage stops at normal operating pressure. The weepage from the TECSA seals has flowed

  • down the ICI tubes, over the penetrations arid down the RVCH. The TECSA leakage contributes to the conditions identified on Nozzle Penetrations 74, 75, 76, 77, and 79.

o Outage Worker Practices A dedicated nozzle, near the center of the reactor head, connects to vent piping, which vents to the upper containment volume to provide reactor vessel head venting of non-condensable gas while* maintaining adequate core cooling and containment integrity.

The vent pipe is removed upon entering into each outage. When removing the heaa vent piping, water has been shown to leak from the flanged connection and associated hose onto the head. The leakage has resulted in multiple flow lines of light corrosion and diffuse boric acid deposits. Previous cleaning of these flow lines has left patches of discoloration on general areas and around the CRDM penetrations in the flow paths.

During removal, the orientation of the flanged connection and piping is not always in.

the exact same location. Variations in worker preferences, during each outage, has caused multiple spill locations. The worker practices are identified.to contribute to all of the 17 identified penetrations and general areas on the RVCH.

A best effort was made to map the expected flows from the TECSA and vent piping to determine where that leakage would be expected to deposit on the RVCH. The mapping efforts showed a reasonable correlation with the as-found conditions on RVCH. See Figure 1 at the end of Enclosure 1 for locations of the specific penetrations with identified conditions. Attachment 1 to Enclosure 1 provides a flow path discussion and insulation diagrams. Attachment 2 to Enclosure 1 provides photographs of insulation arrangement and penetrations.

to AEP-NRC-2020-65 Page8 l&M has already taken corrective actions to resolve leakage from the two identified sources. Further corrective actions have also been initiated to prevent the reoccurrence of the leakage onto the RVCH head.

  • 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, are being installed beginning In U1 C30.

Beginning in U1 C29, the metho_dology 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 in the subsequent startup.

l&M also initiated a corrective action to determine why the continuing TECSA leakage was not corrected earlier.

o Preclude leakage due to outage worker practices

  • l&M recently changed contractors *supporting vent pipe removal and have implemented corrective actions to ensure prevention of spillage on the head.

l&M oversight continually reinforces the standard *that vent pipe leakage is unacceptable.

l&M initiated a corrective action to provide training to eliminate worker practices that lead to vent water inadvertently contacting the head.

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#, 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 Command Center and Operations. These conditions will 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.

  • RVCH Structural Integrity l&M replaced the Unit 1 RVCH in 2006. 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 3) 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 to AEP-NRC-2020-65 Page9 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 U1C30.

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 ge,nerator 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. The general surface areas had no identified wastage and t!terefore maintain structural integrity.

Additionally, following CO2 cleaning, during the current refueling outage l&M will be performing a BMV examination. l&M expects that the BMV examination will show no significant degradation. The proposed alternative to perform a 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. Corrective actions taken to control future leakage from the seals and the vent pipe 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 for the previous 15 months. The unidentified leakage over this entire period was between O and 0.05 gpm. There was no increase in RCS leakage that would be indicative of a through wall leak of the RVCH nozzles.

to AEP-NRC-2020-65 Page 10 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* 8)

  • 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 (7) day rolling average of daily Unidentified RCS leak rates greater than or equal to (0.1) gpm.
2. Nine (9) consecutive daily Unidentified RCS leak rates greater than baseline mean (µ).

i 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.
  • Tier,2 Action Level
1. Two (2) consecutive daily Unidentified RCS leak rates greater than or equal to (0.15) gpm.
2. Two (2) of three*(3) consecutive daily Unidentified RCS leak rates greater than or equal to

(µ+2a).

  • 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.

I *

  • Identify the source of the increase in leakage:

o Check any components ~~ flow path~ recently changed or placec:I in service, shutdown, vented, drained, filled, etc.

Enclosure 1 to AEP-NRC-2020-65 Page 11 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 (1) daily Unidentified RCS leak rates greater than or equal to (0.3) 9pm.
2. One (1) daily Unidentified RCS leak rate greater than or equal to (µ+3a).

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.

o Sample Containment Ventilation condensate for activity.

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

Summary While consistent with RIS 201 a:.oe, leakage from the RVCH penetrations cannot be excluded,*

evaluations of the as-found condition of the RVCH and other identified sources of leakage support the conclusion that it is unlikely that the identified conditions ar~ the result of RVCH penetration leal(age. Rather, the assessments that have been performed, including evaluation of the form and location of the deposits, along with mapping of flow paths, support the conclusion that the identified conditions are most likely the result of leakage from the TECSA seals and vent piping. CO2 cleaning of the RVCH will be performed in order to remove existing deposits and corrective actions have been

  • taken to eliminate or mitigate leakage from the seals and vent piping. These actions will provide a baseline for performing visual examinations in the next outage. The substantial margins in structural integrity for the RVCH penetrations and shell provide a high level of confidence that CNP Unit 1 can be operated safety through the next cycle of operation. Given that the above evaluations and proposed alternative visual examinations during the next outage provide a high level of confidence in continued integrity of the RVCH, the additional personnel exposure and risks to staff from COVID-19
  • associated with performing additional volumetric and surface examinations represent a hardship without a compensating increase in the level of quality and safety.

to AEP-NRC-2020-65 Page 12

6. Duration of Proposed Alternative

The proposed altemative will be utilized until the end of refueling outage U1C31 .

7. Precedent This request is comparable to that submitted by Arkansas Nuclear One Unit 1 (ANO) (Reference 5) and Palo Verde Nuclear Generating Station Unit 1 (Palo Verde) (Reference 4), approved by the NRC via References 6 and 7.

Similar to ANO and Palo Verde, 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. ANO, Palo Verde, and l&M all have replacement heads constructed from PVVSCC resistant Alloy 690 materials. Similar to the precedent submittals, l&M determined that evidence supported that the suspected source of the leakage is not RVCH nozzle leakage. Additionally, l&M Is proposing the same Code Case alternative that was accepted for ANO and Palo Verde.

8. References
1. ASME Boiler and Pressure Vessel Code Case N-729-6, Altemative Examination Requirements for PVVR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration WeldsSection XI, Division 1," March 2016.
2. NRC Regulatory Issue Summary (RIS) 2018--06, "Clarification of the Requirement for Reactor Pressure Vessel Upper Head Bare Metal Visual Examinations," December 10, 2018 *

(Agencywide Documents Access and Management System (ADAMS) Accession No. ML18178A137). .

3. MRP-375, *Material Reliability Program: Technical Basis for Reexamination Interval for Alloy 690 PVVR Reactor Vessel Top Head Penetration Nozzles,* February 2014.
4. Arizona Public Service Company letter, "Relief Request 57- Request for Alternative to American Society of Mechanical Engineers code Case N-729-4 for Replacement Reactor Vessel Closure Head Penetrations Nozzles," October 26, 2017 (ADAMS Accession No. ML17299B333).
5. Entergy Operations, Inc. letter, "Request for Relief related to American Society of Mechanical Engineers (ASME) Code Case N-729-4 Supplemental Examination Requirements ANO1-ISl-033," October 31, 2019 (ADAMS Accession No. ML19304A290).
6. NRC letter, "Relief Request No. 57 to Approve Altemate Requirements for the Reactor Pressure Vessel Head Nozzles to Perform a Bare Metal Examination Per ASME Code Case N-729-4,"

February 16, 2018 (ADAMS Accession No. ML18040A331 ).

7. NRC letter, "Relief Request for ANO1-ISl-033 Related to ASME Code Case N-729-4 Supplemental Examination Requirements of Reactor Vessel Closure Head Penetration Nozzles (EPIC L-2019-LLR-0100)," April 23, 2020 (ADAMS Accession No. ML20107J317).
8. WCAP-16465-NP, "Pressurized Water Reactor Owners Group Standard RCS Leakage Action Levels and Response Guidelines for Pressurized Water Reactors," Revision 0, September 2006.
9. Boric Acid Corrosion Guidebook, Revision 2, "Managing Boric Acid Corrosion Issues at PWR Power Stations," July 2012.

to AEP-NRC-2020-65 Page 13 Figure 1:

Reactor Vessel Closure Head Penetrations with Identified Conditions SEE NOTE 2

- - - - - V O I R NOTA 2 180° BOTTQMYJEW VUE DE oessous I I

/

Attachment 1 to Enclosure 1 Flow Path Discussion with Insulation Diagram GENERAL (Reference I & M Drawings DC-12954 and DC-12637 at the end of this attachment)

The insulation package installed above the Reactor Vessel Closure Head (RVCH) is comprised of a series of metal panels with butted seam joints and cut-outs for each of the 53 Control Rod Drive Mechanisms (CROM) and five In-Core Instrument (ICI) tubes (penetrations 74-77, 79). The seams are aligned with straight courses of CROM and ICI tubes. The seams and cut-outs in the Insulation panels are the pathways for the Reactor Coolant System (RCS) water spilled from the RVCH vent line to reach the RVCH. Once the spillage reaches the RVCH, it flows down the surface where it either Intersects with lower penetrations or deposits in the expanses between them. The five ICI tubes are peripherally located and are lower on the concavity of the RVCH. Therefore, ICI thermocouple sealing assembly (TECSA) leakage is localized to the associated ICI penetration and the adjacent surfaces lower on the RVCH.

All but two of CROM penetrations (8 and 64) with relevant conditions are grouped together In roughly one quadrant of the RVCH from 270 degree ( 0 ) to 0°. The remaining two CROM penetrations are either located Immediately downhill of a penetration where spillage has occurred or share an insulation panel seam with other penetrations where spillage has occurred.

PENETRATION 1 Relevant Condition - Discoloration Location - Top Dead Center of RVCH Leakage pathway(s)- insulation panel seam, CROM tube cut-out Penetration 1 is located at the highest point on the RVCH. From this point, the RCS spillage from the insulation seam and the associated CRDM tube cut-out flowed down the RVCH predominately toward azimuth 315° and Penetrations 10, 46 and 70; toward azimuth 0° and Penetrations 7, 23 and 59; and toward azimuth 90° and Penetration 8. Since the spillage flowed away from the highest point, only discoloration at the annulus of the penetration remains.

PENETRATION 7 Relevant Condition - Discoloration Location - Azimuth 0° Leakage pathway(s)- in flow path directly downhill from Penetration 1, insulation panel seam, CRDM tube cut-out

  • Penetration 7 is located immediately downhill of Penetration 1. Spillage flowing from there, from the insulation seam, and from the associated CROM tube cut-out has resulted in discoloration at-the annulus of the penetration.

PENETRATION 8 Relevant Condition - Discoloration, Boric Acid Deposits Location -Azimuth 90°Leakage pathway(s) - in flow path directly downhill from Penetration 1, insulation panel seam, CRDM tube cut-out Penetration 8 is Immediately downhill of Penetration 1. Spillage flowing from there, from the insulation seam, and the associated CROM tube cut-out has resulted in discoloration and light boric acid deposits at the annulus of the penetration.

to Enclosure 1 to AEP-NRC-2020-65 Page2 PENETRATION 10 Relevant Condition - Areas of Corrosion, Boric Acid Deposits Location -Azimuth 315° Leakage pathway(s) - in flow path directly downhill from Penetration 1, insulation panel seam, CRDM tube cut-out Penetration 10 is immediately downhill of Penetration 1. Spillage flowing from there; from the insulation seam shared with Penetrations 22, 23, and 64; and from the associated CRDM tube cut-out has resulted in discoloration and light boric acid deposits at the annulus of the penetration.

PENETRATION 22 Relevant Condition- Discoloration Location - Azimuth 270° Leakage pathway(s)- insulation panel seam near head vent hose storage area, CRDM tube cut-out Penetration 22 shares an insulation panel seam with Penetrations 10, 23, 64, and 75. Spillage from the insulation seam shared with Penetrations 10, 23, and 64 and from the associated CRDM tube cut-out has resulted in discoloration at the annulus of the penetration.

PENETRATION 23 Relevant Condition - Discoloration Location - Azimuth o*

Leakage pathway(s}- In flow path directly downhill from Penetrations 1 and 7, insulation panel seam, CRDM tube cut-out Penetration 23 is located Immediately downhill of Penetrations 1 and 7. Spillage flowing from there; from the insulation seam shared with Penetrations 10, 22, 64, and 75; and from the associated CRDM tube cut-out has resulted in discoloration at the annulus of the penetration.

PENETRATION 39 Relevant Condition - Discoloration Location - =Azimuth 350° Leakage pathway(s)- in flow path downhill from Penetration 1, insulation panel seam, CRDM tube cut-out Penetration 39 is located downhill of Penetration 1 and shares an insulation panel seam with Penetration 59. Spillage flowing from Penetration 1, from the insulation seam shared with Penetration 59, and from the associated CRDM tube cut-out has resulted in discoloration at the annulus of the penetration.

PENETRATION 46 Relevant Condition - Discoloration Location -Azimuth 315° Leakage pathway(s)- in flow path directly downhill from Penetrations 1 and 10, Insulation panel seam, CRDM tube cut-out Penetration 46 is immediately downhill of Penetrations 1 and 10. Spillage flowing from there, from the insulation seam shared with Penetration 70, and from the associated CROM tube cut-out has resulted in discoloration at the penetration.

to Enclosure 1 to AEP-NRC-2020-65 Page3 PENETRATION 59 Relevant Condition - Areas of Corrosion, Discoloration Location - Azimuth o*

Leakage pathway(s) - In flow path directly downhill from Penetrations 1, 7, and 23; insulation panel seam; CROM tube cut-out Penetration 59 Is located immediately downhill of Penetrations 1, 7, and 23. Spillage flowing from there, from the insulation seam shared with Penetration 39, and from the associated CROM tube cut-out has resulted in areas of corrosion and discoloration at the penetration.

PENETRATION 62 Relevant Condition - Discoloration between 62 and 58 Location - *Azimuth 280° Leakage pathway(s) - insulation panel seam The discoloration between Penetrations 62 and 58 does not emanate from Penetration 62. It is caused by water that has spilled somewhere further up the RVCH In the vicinity of Penetration 38 and flowed downward. Penetration 38 shares an insulation panel seam with Penetration 58.

Spillage from the insulation panel seam shared with Penetration 58 has resulted in discoloration spread over the RVCH surface.

PENETRATION 64 Relevant Condition - Discoloration Location - *Azimuth 20° Leakage pathway(s) - insulation panel seam, CRDM tube cut-out Penetration 64 shares an insulation panel seam with Penetrations 10, 22, 23, and 75. Spillage from the insulation seam shared with Penetrations 10, 22, 23, and 75 and from the associated CRDM tube cut-out has resulted in discoloration at the annulus of the penetration.

PENETRATION 70 Relevant Condition - Areas of Corrosion, Discoloration Location - Azimuth 315° Leakage pathway(s)- in flow path directly downhill from Penetration 1, 10, and 46; insulation panel seam; CRDM tube cut-out Penetration 70 is immediately downhill of Penetrations 1, 10, and 46. Spillage flowing from there, from the insulation seam shared with Penetration 46, and from the associated CRDM tube cut-out has resulted in ar~as of corrosion and discoloration at the penetration.

PENETRATION 74 Relevant Condition - Boric Acid Deposits Location -Azimuth 337.5° Leakage pathway(s) -TECSA, ICI tube cut-out Penetration 74 is an ICI penetration, with a known history of TECSA leakage (as evidenced by the staining down the ICI tube). This leakage, along with spillage from the associated ICI tube cut-out, has resulted in light boric acid deposits at the penetration.

to Enclosure 1 to AEP-NRC-2020-65 Page4 PENETRATION 75 Relevant Condition - Areas of Corrosion, Boric Acid Deposits, Discoloration Location - Azimuth 22.s*

Leakage pathway(s) - TECSA, insulation panel seam, ICI tube cut-out Penetration 75 is an ICI penetration, with a known history of TECSA leakage (as evidenced by the staining down the ICI tube). This leakage, along with spillage flowing from the insulation seam shared with Penetrations 10, 23, and 64 and the associated ICI tube cut-out, has resulted in areas of corrosion, light boric acid deposits, and discoloration at the penetration.

PENETRATION 76 Relevant Condition - Areas of Corrosion, Boric Acid Deposits Location - Azimuth 67.5° Leakage pathway(s) - TECSA, ICI tube cut-out Penetration 76 is an ICI penetration, with a known history of TECSA leakage (as evidenced by the staining down the ICI tube). This leakage, along with spillage from the associated ICI tube cut-out, has resulted in areas of corrosion and light boric acid deposits at the penetration.

PENETRATION 77 Relevant Condition - Boric Acid Deposits, Discoloration Location -Azimuth 157.5° Leakage pathway(s) - TECSA, ICI tube cut-out Penetration 77 is an ICI penetration, with a known history of TECSA leakage (as evidenced by the staining down the ICI tube). This leakage, along with spillage from the associated ICI tube cut-out, has resulted in light boric acid deposits and discoloration at the penetration.

PENETRATION 79 Relevant Condition - Boric Acid Deposits Location -Azimuth 247.5° Leakage pathway(s)-TECSA, ICI tube cut-out Penetration 79 is an ICI penetration, with a known history of TECSA leakage (as evidenced by the staining down the ICI tube). This leakage, along with spillage from the associated ICI tube cut-out, has resulted in light boric acid deposits at the penetration.

RVCH SURFACES Relevant Condition - Discoloration Location - various Leakage pathway(s) - various Similar to the Relevant Condition at Penetration 62, this discoloration is caused by spillage occurring somewhere further up the RVCH and flowing downward. This spillage has resulted in discoloration spread over the RVCH surface.

to Enclosure 1 to AEP-NRC-2020-65 Pages Excerpt from DC-12954

Attachment Page 6 1 to Enclosure 1 to AEP-NRC-2020-65 Excerpt from OC-12637 110

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Penetration 1 Penetration 1 As Found As left

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Penetration 8 Penetration 8 As Found As Left

Penetration 10 As Found Penetration 10 As Left

Penetration 22 Penetration 22 As Left As Found

Penetration 23 Penetration 39 Penetration 39 As Found As left

Penetration 46 Penetration 46 As Found As Left

Penetration 59 Penetration 59 As Found As Left

Penetration 64 Penetration 70 Penetration 70 As Found As Left

Penetration 74 Penetration 74 As Left As Found

Penetration 75 Penetration 75 As Left As Found

Penetration 76 Penetration 76 As Found As left

Penetration n Penetration n As Left As Found

Penetration 79 As Found Penetration 79 As left

View of underside of insulation panels showing leakage through seams Enclosure 2 to AEP-NRC-2020-65 Regulatory Commitment for Reactor Vessel Closure Head (RVCH) Inspection REVISED REGULATORY COMMITMENTS 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 aoolicable) l&M will perform a bare metal visual (BMV) Inspection of the CNP Unit 1 Cycle 31 Refueling Unit 1 RVCH In the next refueling outage in accordance with the outage latest revision of Code Case N-729 endorsed in 10 CFR 50.55a.

This commitment is related to the D.C. Cook (CNP) Unit 1 relief request ISIR-5-04.

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