ML14135A320

Request for Additional Information Concerning the Reactor Vessel Internals Afing Management Program Submittal

ML14135A320
Person / Time
Site:	Cook
Issue date:	06/06/2014
From:	Thomas Wengert Plant Licensing Branch III
To:	Weber L Nuclear Generation Group
Wengert T
References
TAC MF0050, TAC MF0051
Download: ML14135A320 (7)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 6, 2014 Mr. Lawrence J. Weber Senior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, Ml 49106

SUBJECT:

DONALD C. COOK NUCLEAR PLANT, UNITS 1 AND 2- REQUEST FOR ADDITIONAL INFORMATION CONCERNING THE REACTOR VESSEL INTERNALS AGING MANAGEMENT PROGRAM SUBMITTAL (TAC NOS.

MF0050 AND MF0051)

Dear Mr. Weber:

By letter dated October 1, 2012 (Agencywide Documents Access and Management System Accession No. ML12284A320), Indiana Michigan Power (I&M, the licensee) submitted an aging management program (AMP) for the Donald C. Cook Nuclear Plant, Units 1 and 2 reactor vessel internals.

The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the subject submittal and determined that additional information is needed to complete the review, as described in the enclosed request for additional information (RAI). The NRC staff clarified the draft RAI in conference calls conducted on October 23, 2013, January 14, 2014, and May 2, 2014. On May 2, 2014, your staff agreed to respond to this RAI in accordance with the following schedule:

RAI-1 : 60 days from the date of this letter RAI-2 : October 10, 2014 RAI-3: October 10, 2014 RAI-4: October 10, 2014 RAI-5: August 8, 2014 RAI-6: October 10, 2014 RAI-7 : 60 days from the date of this letter RAI-8: 90 days from the date of this letter

L. Weber Please feel free to contact me at (301) 415-4037 if you need any further clarification of the questions in the enclosure.

Sincerely, Thomas J. Wengert, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and 50-316

Enclosure:

Request for Additional Information cc: Distribution via ListServ

REQUEST FOR ADDITIONAL INFORMATION (RAI)

REGARDING AGING MANAGEMENT PROGRAM FOR REACTOR VESSEL INTERNALS DONALD C. COOK NUCLEAR PLANT. UNITS 1 AND 2 INDIANA MICHIGAN POWER COMPANY DOCKET NOS. 50-315 AND 50-316 By letter dated October 1, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12284A320), Indiana Michigan Power (I&M , the licensee) submitted an aging management program (AMP) for the Donald C. Cook Nuclear Plant (CNP), Units 1 and 2 reactor vessel internals (RVI) . The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the submittal and requests the following information to complete its review.

In Sections 2.1 and 2.2 of the submittal, I&M states "Some examples of the results ... " in reference to reviewing operating experience (OE) and records search of plant-specific RVI AMP relevant information. Confirm that these examples include any and all relevant OE and records that will impact the CNP RVI AMPs.

RAI-2

In Section 4.4.2.1 of the submittal , I&M states that a response to Action Item 1 will be submitted prior to the period of extended operation for each unit. Justify that the Pressurized Water Reactor Owners Group (PWROG)-provided results of project PA-MSC-0938 are fully bounding for CNP or identify any gaps between these results and those for CNP. The response to RAI Question 3 should be incorporated into this response.

As discussed in References 1 and 2, provide the following information related to verification of the applicability of MRP-227-A to CNP, Units 1 and 2:

(a) Do the CNP, Units 1 and 2 RVI have non-weld or bolting austenitic stainless steel components with 20 percent cold work or greater, and subject to operating stresses greater than 30 kilopounds per square inch? If so, perform a plant-specific evaluation to determine the aging management requirements for the affected components .

(b) Have CNP, Units 1 and 2 ever utilized a typical design or fuel management that could make the assumptions of MRP-227 -A regarding core loading/core design non-representative for the plant, including power changes/uprates? If so, describe how the differences were reconciled with the assumptions of MRP-227 -A, or provide a plant-specific AMP for the affected components as appropriate.

Enclosure

In Section 4.4.2.2 of the submittal, I&M states that Action Item 2 will be satisfied through action taken by the PWROG. It is unclear from the licensee's response if it was understood that this Action Item requires an applicant/licensee to evaluate whether any plant-specific components that should be included in the CNP AMP were overlooked in MRP-227-A. The NRC staff requests that I&M perform the plant-specific actions described by Action Item 2 and provide the results.

In Section 4.4.2.3 of the submittal , I&M states that information will be provided to the NRC concerning the strategy for managing split pins prior to the period of extended operation for each unit. The NRC staff noted that the licensee intends to investigate whether replacement of the split pins is necessary. The staff requests that the licensee:

(a) Provide a brief summary of the previous inspections thus far that were performed on the split pins including the type of inspections, frequency of inspections, and the results of the inspection. Confirm whether the split pins are binned under the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) ,Section XI examination category.

(b) If cracking was observed in the split pins, describe the corrective action that was taken to prevent recurrence of the aging degradation. If replacement of the split pins is necessary, the NRC staff requests that the licensee state the type of material that will be used for replacement.

In Section 4.4.2.7 of its submittal, I&M states that responses to Action Item 7 will be submitted prior to the period of extended operation for each unit.

(a) The NRC staff requests that the licensee identify all components for which analyses will be provided under Action Item 7.

The cast austenitic stainless steel (CASS) lower support column bodies are prone to thermal embrittlement, neutron embrittlement, and irradiation-assisted stress corrosion cracking . Since CASS materials with delta ferrite content greater than 20 percent are susceptible to thermal embrittlement, establishing the delta ferrite content in each column is essential to assess the extent of aging degradation due to thermal embrittlement in these columns. In addition, the casting method (i.e., static or centrifugal) will affect the occurrence of thermal embrittlement.

The value of the delta ferrite content can be obtained from a certified material test report (CMTR) for each column .

(b) The NRC staff requests that the licensee provide the delta ferrite content for each lower support column from CMTRs, and provide the casting method for each column .

RAI-7

Sections 2.1 and 2.2 of I&M's submittal describe a number of baffle-former and barrel-former bolt failures that have occurred at CNP, Units 1 and 2.

(a) Please discuss the root cause of the failed bolts.

(b) Given the relative frequency of these bolt failures at CNP, Units 1 and 2, above and beyond general industry experience, justify the use of the generic MRP-227 -A guidelines for bolt inspection, as opposed to augmenting the CNP AMP to require more comprehensive bolt inspections.

Section 2.1 .3 of I&M's submittal discusses indications that were found in the CNP, Unit 1 lower radial support structure (LRSS) clevis insert bolts while performing the ASME Code,Section XI in service inspections in 2010.

(a) Please discuss the safety significance of the LRSS clevis insert bolts. Specifically, address the potential impact of clevis insert bolting failures on the capability to safely shut down the reactor.

(b) Please provide a summary of activities related to aging management of the clevis insert bolts, dowell pins, or other clevis insert alignment and interfacing components that have exhibited age-related degradation. The summary should discuss the following items:

• inspection findings related to degradation of the clevis insert components;

• repair and/or replacement activities for clevis insert components, including the number and type of clevis insert components replaced, the material type or specification for the replacement components; and

• methods for ensuring the functionality of the repaired configuration; Additionally, please provide any report(s) documenting root cause analyses and supporting metallurgical analyses of the failed clevis insert components.

(c) Please state whether subsequent inspections of clevis insert components were performed during the 2013 outage (when clevis insert component replacement activities occurred) to detect additional component degradation, beyond that found in 2010.

For any clevis insert component inspections performed in 2013, please discuss the examination method (e.g., VT-3) and results, noting whether any additional degradation to the clevis insert components was found beyond that identified in 2010. These discussions may reference the applicable sections of the reports requested in part (a) of RAI Question 8 above.

Please detail the aging management approach for the clevis insert components going forward . Specifically, identify whether CNP will continue to perform inspections of the

clevis insert components in accordance with the MRP-227 -A guidelines and the ASME Code,Section XI (VT-3 visual examination every 10-year inservice inspection interval),

or whether a more comprehensive examination method and/or shorter examination cycle will be used. If the clevis insert inspection criterion delineated in MRP-227-A is not modified for CNP, please provide a technical explanation for the adequacy of the current VT -3 visual examination method every 10 years to detect cracking before it results in component failure.

(d) Please state when the last ASME Code,Section XI inservice inspection was performed for the clevis inserts at CNP, Unit 2, and discuss whether there were any findings of age-related degradation.

Considering the OE related to clevis insert bolt degradation at CNP, Unit 1, please indicate whether any repair/replacement activities were performed for potentially susceptible clevis insert bolts at CNP, Unit 2, or provide explanation for not replacing susceptible bolting.

References

1. Meeting Summary EPRI-Westinghouse January 22-23, 2013, February 21, 2013 (ADAMS Accession No. ML13042A048).

2. February 25, 2013 Summary of Teleconference with Electric Power Research Institute (EPRI) and Westinghouse Electric Company, March 15, 2013 (ADAMS Accession No. ML13067A262).

ML14135A320 *via memorandum OFFICE LPL3-1/PM LPL3-1/LA NRR/EVIB/BC* LPL3-1/BC LPL3-1/PM NAME TWengert MHenderson SRosenberg RCarlson TWengert DATE 05/30/14 05/19/14 02/26/14 06/05/14 06/6/14