ML13263A338

RAI Set 14

ML13263A338
Person / Time
Site:	Sequoyah
Issue date:	09/26/2013
From:	Plasse R License Renewal Projects Branch 1
To:	James Shea Tennessee Valley Authority
References
TAC MF0481, TAC MF0482
Download: ML13263A338 (7)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 September 26, 2013 Mr. Joe W. Shea Vice President, Nuclear Licensing Tennessee Valley Authority P.O. Box 2000 Soddy-Daisy, TN 37384

SUBJECT:

REQUESTS FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2, LICENSE RENEWAL APPLICATION (TAC NOS. MF0481 AND MF0482)- SET 14.

Dear Mr. Shea:

By letter dated January 7, 2013, Tennessee Valley Authority submitted an application pursuant to Title 10 of the Code of Federal Regulations (CFR) Part 54, to renew the operating license DPR-77 and DPR-79 for Sequoyah Nuclear Plant, Units 1 and 2, for review by the U.S. Nuclear Regulatory Commission (NRC) staff. The staff is reviewing the information contained in the license renewal application and has identified, in the enclosure, areas where additional information is needed to complete the review.

These requests for additional information (RAis), outlined in the Enclosure were discussed with Henry Lee, and a mutually agreeable date for the response to RAI B.1.34-8 is within 60 days from the date of this letter, and for the rest of the enclosed RAis the mutually agreeable date for response is within 30 days from the date of this letter. If you have any questions, please contact me at 301-415-1427 or by e-mail at Richard.Piasse@nrc.gov.

Sincerely,

~

Richard A. Plasse, Project Manager Projects Branch 1 Division of License Renewal Office of Nuclear Reactor Regulation Docket Nos. 50-327 and 50-328

Enclosure:

Requests for Additional Information cc: Listserv

ML13263A338 OFFICE LA:RPB1 :DLR PM:RPB1 :DLR BC:RPB1 :DLR PM: RPB1 :DLR NAME Y. Edmonds E Sayee Y Diaz-Sanabria R Plasse DATE 9/26/2013 9/26/2013 9/26/2013 9/26/2013 SEQUOYAH NUCLEAR PLANT, UNITS 1 AND 2 LICENSE RENEWAL APPLICATION REQUESTS FOR ADDITIONAL INFORMATION RAI 8.0.4-1 a

Background:

In its July 29, 2013, response to RAI B.0.4-1, the applicant provided additional information on its programmatic activities for the ongoing review of operating experience. The applicant provided this information to support consistency of these activities with the areas described in Appendix B to the Generic Aging Lessons Learned (GALL) Report, which the NRC established in Final License Renewal Interim Staff Guidance, LR ISG 2011 05, "Ongoing Review of Operating Experience," dated March 16, 2012.

Based on the response to RAI B.0.4-1, the staff could not ascertain whether certain aspects of the applicant's operating experience review activities are fully consistent with Appendix B to the GALL Report. Specifically:

1) Appendix B to the GALL Report states that NRC and industry guidance documents and standards applicable to aging management should be considered as sources of operating experience, and there should be written plans and expectations for finding and processing these sources. The applicant stated that it uses its Operating Experience Program to monitor industry operating experience based on Institute of Nuclear Power Operations (INPO) guidelines. The INPO guidelines identify certain sources of NRC and industry operating experience, but there are other sources that could be applicable to aging management. The applicant did not describe how it will identify and evaluate these other sources.

2) Appendix B to the GALL Report states that any adverse trends associated with age-related degradation trend codes should be entered into the corrective action program for evaluation. The applicant stated that its Corrective Action Program includes trend codes for identifying aging management and license renewal items, and it periodically reviews these codes for trends. However, the applicant did not describe how it will address adverse trends.

3) Appendix B to the GALL Report states that personnel responsible for implementing the aging management programs and processing plant-specific and industry operating experience should receive training on age-related degradation and aging management topics. The applicant stated that it trains personnel based on the complexity of the job performance requirements and assigned responsibilities. However, the applicant did not indicate that this training specifically includes topics on age-related degradation and aging management.

ENCLOSURE

2 Request:

If demonstrating consistency with Appendix B to the GALL Report, address the following:

1. Describe the written plans and expectations for finding and processing sources of NRC and industry guidance documents and standards applicable to aging management that are not covered by the sources listed in the INPO guidelines.

2. Indicate whether adverse trends identified from review of the aging management and license renewal trend codes will be entered into the Corrective Action Program for evaluation.

3. Demonstrate that personnel training will include topics on age-related degradation and aging management.

If not demonstrating consistency with Appendix B to the GALL Report, justify why the operating experience review activities provide for the adequate consideration of operating experience involving age-related degradation and aging management to maintain the effectiveness of the aging management programs (AMPs) and activities.

Identify any necessary enhancements to existing activities based on the response to this request. Provide the schedule for implementing these enhancements and a justification if implementation is later than the date when the renewed operating license is scheduled to be issued, if approved.

RAI 3.5.1-87

Background:

LRA Table 3.5.1, item 3.5.1-87, states "[v]ibration, flexing of the joint, cyclic shear loads, thermal cycles and other causes can cause partial self-loosening of a fastener. These causes of loosening are minor contributors in structural steel and steel component threaded connections and are eliminated by initial preload bolt torquing. The LRA further states:

"SQN uses site procedures and manufacturer recommendations to provide guidance for proper torquing of nuts and bolts used in structural applications ..... Therefore, loss of preload due to self-loosening is not an aging effect requiring management for structural steel and steel component threaded fasteners within the scope of license renewal."

The ASME Section XI, Subsection IWF Program described in the GALL Report, program element, "parameters monitored or inspected," states "Structural bolts are monitored for corrosion and loss of integrity of bolted connections due to self-loosening and material conditions that can affect structural integrity." Based on this, the staff's position is that the potential loss of preload due to self-loosening from vibration, flexing of the joint, cyclic shear loads, thermal cycles and other causes is an aging effect requiring management.

Request:

Provide sufficient technical basis for concluding loss of preload due to self-loosening is not an aging effect requiring management, or identify an AMPs to manage this aging effect.

RAI 3.5.1-57

Background:

LRA Table 3.5.1, item 3.5.1-57 addresses constant and variable load spring hangers; guides; stops exposed to air-indoor, uncontrolled or air-outdoor environments, which will be managed for aging effects of loss of mechanical function due to corrosion, distortion, dirt, overload, fatigue due to vibratory and cyclic thermal loads. The LRA states that this item is not applicable because loss of mechanical function due to distortion, dirt, overload, and fatigue due to vibratory and cyclic thermal loads is not an aging effect requiring management because such failures typically result from inadequate designs rather than aging effects.

The staff notes that LRA Section 8.1.17, "lnservice Inspection- IWF (ISI-IWF) Program" states

"[v]isual examinations are conducted to determine the general mechanical and structural condition or degradation of component supports such as verification of clearances, settings, physical displacements, loose or missing parts, debris, corrosion, wear, erosion, or the loss of integrity at welded or bolted connections."

The GALL Report, identifies loss of mechanical function in Class 1 piping and components (such as constant and variable load spring hangers, guides, stops, sliding surfaces, and vibration isolators) fabricated from steel or other materials, as an aging effect that can occur through the combined influence of a number of aging mechanisms. Such aging mechanisms are not limited to loss of material due to corrosion, but also include distortion, dirt, overload, fatigue due to vibratory and cyclic thermal loads, or elastomer hardening.

The loss of mechanical function due to distortion, dirt, overload, and fatigue due to vibratory and cyclic thermal loads is not solely the result of inadequate design or events. As stated above, the GALL report identifies loss of mechanical function for Class 1 piping and components due to distortion, dirt, overload, fatigue due to vibratory and cyclic thermal loads as an aging effect to be managed.

Request:

Provide the staff with sufficient technical basis for concluding loss of mechanical function due to distortion, dirt, overloads, and fatigue due to vibratory and cyclic thermal loads is not an aging effect requiring management, or provide an AMP to manage this aging effect.

RAI 8.1.34-8

Background:

LRA Table 3.1.2-2, Reactor Vessel Internals, indicates that the clevis insert bolts are nickel alloy and that cracking will be managed by the Reactor Vessel Internals Program in the "no additional measure" inspection category. Appendix A to Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guideline (MRP-227-A) indicates that failures of Alloy X-750, precipitation-hardenable nickel-chromium alloy, clevis insert bolts were reported by one Westinghouse designed plant in 2010. Furthermore, the staff noted that these clevis insert bolts failed because of cracking, which is an aging effect that was not addressed in MRP-227-A.

The staff noted that the only aging mechanism requiring management by MRP-227 -A for the clevis insert tolts is wear and the bolts are categorized as an "Existing Programs" component.

Thus, under MRP-227 -A, the clevis insert bolts will be inspected in accordance with the ASME Code, Section Xllnservice Inspection Program to manage the effects due to wear only.

The staff noted that the ASME Code,Section XI specifies a VT-3 visual inspection for the clevis insert bolts, which may not be adequate to detect cracking before bolt failure occurs. In addition, since cracking of the clevis insert bolts was not addressed during the development of MRP-227-A, it is not clear to the staff whether this operating experience is applicable to the applicant and whether the Reactor Vessel Internals Program will need to be modified to account for this operating experience.

Request:

1. Specify the fabrication material, including any applicable heat treatment, for the clevis insert bolts at Units 1 and 2.

2. Discuss and justify whether the operating experience associated with cracking of the clevis insert bolts is applicable to Units 1 and 2.

a. If applicable, discuss and justify how the Reactor Vessel Internals Program will be augmented to require an inspection of the clevis insert bolts capable of detecting cracking. If the Reactor Vessel Internals Program will not be augmented, provide a technical justification for the adequacy of the existing VT-3 visual inspection to detect cracking before it results in clevis insert bolt failure.

Letter to J. Shea from R. Plasse dated September 26, 2013

SUBJECT:

REQUESTS FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE EQUOYAH NUCLEAR PLANT, UNITS 1 AND 2, LICENSE RENEWAL APPLICATION (TAC NOS. MF0481 AND MF0482)- SET 14.

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