ML14253A316
| ML14253A316 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 09/04/2014 |
| From: | Gebbie J P Indiana Michigan Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML14253A310 | List: |
| References | |
| AEP-NRC-2014-59 | |
| Download: ML14253A316 (7) | |
Text
INDIANA MICHIGAN POWERO A unit of Amenrican Electric Power September 4, 2014 Indiana Michigan Power Cook Nuclear Plant One Cook Place Bridgman, MI 49106 Indiana Michigan Power.com AEP-NRC-2014-59 10 CFR 50.4 Docket Nos.: 50-315 50-316 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Donald C. Cook Nuclear Plant Units 1 and 2 SECOND RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION CONCERNING THE REACTOR VESSEL INTERNALS AGING MANAGEMENT PROGRAM
References:
- 1) Letter from J. P. Gebbie, Indiana Michigan Power Company (I&M), to U. S. Nuclear Regulatory Commission (NRC), "Donald C. Cook Nuclear Plant Units 1 and 2, Transmittal of Reactor Vessel Internals Aging Management Program," dated October 1, 2012. Agencywide Documents Access and Management System (ADAMS) Accession No. ML12284A320.
- 2) Letter from T. J. Wengert, NRC, to L. J. Weber, I&M, "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)," dated June 6, 2014. ADAMS Accession No. ML14135A320.
- 3) Letter from J. P. Gebbie, I&M, to NRC, "Donald C. Cook Nuclear Plant Units 1 and 2-First Response to Request for Additional Information Concerning the Reactor Vessel Internals Aging Management Program," dated July 30, 2014.This letter provides Indiana Michigan Power Company's (I&M), licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, response to Requests for Additional Information (RAI) by the U. S. Nuclear Regulatory Commission (NRC) regarding CNP's Reactor Vessel Internals Aging Management Program.By Reference 1, 1&M submitted the CNP Reactor Vessel Internals Aging Management Program. By Reference 2, the NRC transmitted RAIs regarding the program. Reference 3 provided I&M's response to Reference 2, RAI-1, RAI-5, and RAI-7. Enclosure 1 to this letter provides response to Reference 2, RAI-8. Response to RAI-2, RAI-3, RAI-4, and RAI-6 will be provided in accordance with the schedule for response provided by Reference
- 2. Enclosures 2 and 3 contain the Lower Radial Support System Root Cause Evaluation and supporting metallurgical analysis, respectively.
U.S. Nuclear Regulatory Commission AEP-NRC-2014-59 Page 2 There are no new commitments made in this submittal.
Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Manager, at (269) 466-2649.Sincerely, Joel P. Gebbie Site Vice President DMB/kmh
Enclosures:
- 1. Donald C. Cook Nuclear Plant Response to Request for Additional Information Regarding The Reactor Vessel Internals Aging Management Program 2. I&M CAP Document AR 2010-1804-10, Root Cause Evaluation Attachment, "Rx Vessel Core Support Lug Bolting Anomalies" 3. Babcock & Wilcox Report, S-1473-002, Revision 0, "Examination of Clevis Bolts Removed from D. C. Cook Nuclear Plant" c: M. L. Chawla, NRC Washington, D.C.J. T. King -MPSC MDEQ- RMD/RPS NRC Resident Inspector C. D. Pederson, NRC Region III A. J. Williamson -AEP Ft. Wayne ENCLOSURE 1 TO AEP-NRC-2014-59 DONALD C. COOK NUCLEAR PLANT REACTOR VESSEL INTERNALS AGING MANAGEMENT PROGRAM RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING THE REACTOR VESSEL INTERNALS AGING MANAGEMENT PROGRAM List of Acronyms: ADAMS Agencywide Documents Access and Management System AMP Aging Management Program ASME American Society of Mechanical Engineers CNP Donald C. Cook Nuclear Power Plant EPRI Electric Power Research Institute I&M Indiana & Michigan Power LRSS Lower Radial Support System MRP Materials Reliability Program NRC Nuclear Regulatory Commission PWROG Pressurized Water Reactor Owners Group RAI Request for Additional Information RCE Root Cause Evaluation RFO Refueling Outage RVI Reactor Vessel Internals By letter dated October 1, 2012 (ADAMS Accession No. ML1 2284A320), I&M, the licensee for CNP, submitted an AMP for CNP, Units 1 and 2, RVI to the NRC. By letter dated June 6, 2014 (ADAMS Accession No. ML14135A320), the NRC staff reviewed the submittal and requested additional information to complete its review. By letter dated July 30, 2014, the responses to RAI-1, RAI-5, and RAI-7 were provided to the NRC in AEP-NRC-2014-56.
The response to RAI-8 is provided in this enclosure.
The remaining RAls (RAI-2, RAI-3, RAI-4, and RAI-6) will be provided at a later date per the schedule described in the June 6, 2014 letter from the NRC.
Enclosure 1 to AEP-NRC-2014-59 Page 2 RAI-8 Section 2.1.3 of I&M's submittal discusses indications that were found in the CNP, Unit I lower radial support structure (LRSS) clevis insert bolts while performing the ASME Code, Section Xl 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 X1 (VT-3 visual examination every 1 0-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 X1 inservice inspection was performed for the clevis inserts at CNP, Unit 2, and discuss whether there were any findings of age-related degradation.
Enclosure 1 to AEP-NRC-2014-59 Page 3 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.Response to RAI-8 The LRSS RCE and supporting metallurgical analysis are provided for information only, as Enclosure 2 and 3 of this letter, respectively.
A description of the CNP Unit 1 LRSS can be found in the LRSS RCE in Section 2.1. A summary of the CNP Unit 1 LRSS clevis insert bolt degradation discovery and repair is described below.Previously a meeting on LRSS clevis insert bolt degradation was held on March 27, 2014, between the NRC, EPRI MRP, PWROG, and a number of utilities including I&M. Items discussed at the meeting included the CNP Unit 1 LRSS RCE, potential LRSS failure modes, and safety significance.
It was communicated that following the meeting, additional industry communication regarding the LRSS would be issued by Westinghouse.
Westinghouse published Technical Bulletin (TB) TB-14-5, Revision 0, "Reactor Internals Lower Radial Support Clevis Insert Cap Screw Degradation," on August 25, 2014. This document will be referred to as the "Technical Bulletin" in the remainder of the response.
The document provides a summary of the Operating Experience (OE) as well as root cause findings and the applicability of these findings on Westinghouse pressurized water reactor designs. In addition, the document provides a review of safety implications related to the OE and inspection recommendations for consideration in aging management programs.(a) The LRSS clevis insert bolt degraded condition did not result in the loss of the ability of the LRSS to perform its intended design function, or challenge the ability to shut down the reactor. Even with the postulated failure of all bolts and dowel pins, clevis inserts are predicted to remain in place and functional.
This is discussed in the LRSS RCE in Section 2.8.The clevis inserts have several redundant means of attachment including bolts, pins, an interference fit into lugs, and the geometry of the assembled system. This is the expected order of failure as presented by the PWROG in the meeting held on March 27, 2014. The LRSS RCE determined that the primary failure was the LRSS clevis bolts as described in Section 1.3. The dowel pin failure was secondary to the clevis bolt failures as described in Section 2.4.7 of the LRSS RCE.Although bolt failures do not challenge the ability of the LRSS to perform its design function, they are important to the plant from a commercial perspective.
If the LRSS clevis bolts, dowel pins, and interference fit are defeated, then a commercial concern exists. When the core barrel is removed for maintenance in this condition, then a clevis insert may not be fully restrained.
A clevis insert may then be allowed to displace from its lugs. Repair or replacement of a LRSS clevis insert is economically challenging.
It should be noted that the core barrel is not removed unless the reactor is completely defueled.(b) The CNP Unit 1 LRSS clevis insert bolts and dowel pins were visually inspected, by VT-3, during the ULC23 RFO in the spring of 2010. The inspection revealed that 7 of 48 LRSS Enclosure 1 to AEP-NRC-2014-59 Page 4 clevis insert bolts showed wear between the bolt head and dowel pin, indicating failure. In addition, 1 of 12 LRSS dowel pins had broken tack welds, was rotated and displaced.
An Operability Determination Evaluation was documented, in CNPs corrective action program, following the discovery and the unit was returned to service while a determination was made for the best course of action. A repair methodology was developed over the following two fuel cycles.A commercial minimum bolt pattern analysis was performed by the, reactor Original Equipment Manufacturer to determine the minimum number of bolts required at each clevis insert to ensure that the installed bolts would remain functional for the remainder of the design life when no credit was taken for original bolts left in place. The replacement bolt installation was overseen by quality assurance and quality control requirements.
A pre-service visual inspection, by VT-3, was performed before the unit was returned to service.I&M replaced a commercial minimum bolt pattern in each CNP Unit 1 LRSS clevis insert. A total of 28 of 48 clevis insert bolts were replaced during the U1C25 RFO in the spring of 2013. Replacement bolts were a two piece design. The bolts were fabricated from one material and a crimp cup locking device attached to the head of the bolt was fabricated from a different material.
The replacement bolts were fabricated from SB-637 UNS N07750, Type 2 in accordance with Section II and Section III of the ASME Boiler & Pressure Vessel (B&PV) Code, 2004 Edition, no Addenda, and Code Case N-60-5,Section III, Division 1 of the ASME B&PV Code (commonly known as X-750). Replacement bolt locking devices were fabricated from SA-479 Type 304L in accordance with Section II and Section III of the ASME B&PV Code, 2004 Edition, no Addenda.(c) In the Spring of 2013, a pre-maintenance visual inspection was performed on the LRSS clevis inserts during the Ul C25 RFO immediately upon removal of the lower internals before any replacement equipment was mobilized in the reactor vessel. The pre-maintenance VT-3 revealed that 8 of 48 LRSS clevis insert bolts and 1 of 12 dowel pins had visual indication of failure. These results revealed 1 additional LRSS clevis bolt with visual indication compared to the 2010 inspection.
LRSS wear surfaces were inspected on the clevises attached to the vessel wall, and the radial keys attached to the core barrel. No abnormal or excessive wear was observed on any surfaces.
No dislocation or shifting of components was observed, and all inserts appeared to be fully seated in their originally installed location.The aging management approach for the LRSS clevis insert bolts and dowel pins follows the guidance provided in MRP-227-A, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines," the ASME B&PV Code, and plant specific programs.
Although none of this guidance explicitly requires the inspection of the LRSS clevis insert bolts or dowel pins, in practice these components have been inspected at both CNP units.Considering the low safety significance of LRSS clevis insert bolt and dowel pin failures, no augmentation has been made to the inspection requirements or aging management approach for these components.
This is discussed further in the LRSS RCE in Section 2.4.9. The Technical Bulletin includes inspection recommendations, but no additional requirements.
I&M has not had adequate time to review and evaluate the Technical Bulletin between the time of its issuance and the due date of this RAI response.
The Technical Enclosure 1 to AEP-NRC-2014-59 Page 5 Bulletin has been entered into CNP's Corrective Action Program for evaluation.
I&M continues to participate in industry programs related to reactor vessel internals aging management, including LRSS clevis insert bolts.(d) The last ASME Code,Section XI In-Service Inspection was performed for the clevis inserts at CNP, Unit 2, in 1996 with no relevant indications.
I&M obtained relief for inspections requiring removal of the core barrel for CNP Unit 2 during the 2009 ASME Code, Section Xl In-Service Inspection.
Therefore, the LRSS was not inspected during the 2009 RFO.However, the core barrel was removed for other work in 2010, so the opportunity was taken to perform a VT-3 visual inspection of the CNP Unit 2 LRSS clevis inserts. No relevant indications have been identified in the CNP Unit 2 LRSS.I&M continues to manage the LRSS clevis inserts, bolts, and dowel pins, for both CNP units, using the current requirements and programs.
The low safety significance of bolt failure does not warrant pre-emptive replacement.
This is discussed in the LRSS RCE in Section 2.3. The Technical Bulletin includes inspection recommendations, but no additional requirements.
I&M has not had adequate time to review and evaluate the Technical Bulletin between the time of its issuance and the due date of this RAI response.
The Technical Bulletin has been entered into CNP's Corrective Action Program for evaluation.
I&M continues to participate in industry programs related to RVI aging management, including LRSS clevis insert bolts.