Nuclear Regulatory Commission.
Notice, Director's decision regarding petition.
The U.S. Nuclear Regulatory Commission (NRC) has issued a director's decision with regard to a petition dated June 30, 2016, filed by Mr. David A. Lochbaum of the Union of Concerned Scientists (the petitioner), requesting that the NRC take enforcement action against Entergy Nuclear Operations, Inc., the licensee for Indian Point Nuclear Generating, Units No. 2 and 3 (Indian Point 2 and 3). The petitioner's requests and the director's decision are included in the SUPPLEMENTARY INFORMATION section of this document.
Please refer to Docket ID NRC-2017-0074 when contacting the NRC about the availability of information regarding this document. You may obtain publicly-available information related to this document using any of the following methods:
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Notice is hereby given that the Director, Office of Nuclear Reactor Regulation, has issued a director's decision DD-17-01 (ADAMS Accession No. ML17065A030) on a petition filed by the petitioner on June 30, 2016 (ADAMS Accession No. ML16187A186). The petition was supplemented by letter dated January 10, 2017 (ADAMS Accession No. ML17011A012).
In response to degradation of reactor vessel baffle-former bolts (BFBs) identified at Indian Point 2 during its spring 2016 refueling outage, the petitioner requested the NRC to:
(1) Issue an order requiring the licensee to inspect the reactor vessel BFBs and install the downflow to upflow modification at Indian Point 2 during its next refueling outage (i.e., spring 2018).
(2) Issue a demand for information requiring the licensee to submit an operability determination to the NRC regarding continued operation of Indian Point 3 until its reactor vessel BFBs can be inspected according to the Electric Power Research Institute Materials Reliability Program Topical Report MRP-227-A, “Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines” (ADAMS Package Accession No. ML120170453).
(3) Issue a demand for Information requiring the licensee to submit an evaluation of the performance, role, and operating experience of the Indian Point metal impact monitoring system in detecting and responding to indications of loose parts (such as broken baffle bolt heads and locking tab bars) within the reactor coolant system.
As the basis for this request, the petitioner cited Licensee Event Report 2016-004-00, “Unanalyzed Condition due to Degraded Reactor Baffle-Former Bolts,” submitted by the licensee on May 31, 2016 (ADAMS Accession No. ML16159A219), that describes an event where there was an unanalyzed condition due to degraded reactor vessel BFBs at Indian Point 2, which is reportable under § 50.73(a)(2)(ii)(B) of title 10 of the Code of Federal Regulations (10 CFR). The petitioner states that (1) an order is the proper means for ensuring that the bolts are inspected and that the downflow to upflow modification is installed during the next refueling outage at Indian Point 2, (2) Indian Point 3 is potentially operating with degraded BFBs and an operability determination is the mechanism established by the NRC to properly evaluate such situations, and (3) the metal impact monitoring system as described in the updated final safety analysis report has the potential to act as an alternate monitoring system to Start Printed Page 18678identify degraded BFBs, yet neither the NRC nor the licensee have referred to this system in publicly available documents relating to this issue.
On July 28, 2016, the petitioner and the licensee met with the NRC's Petition Review Board. The meeting provided the petitioner and the licensee an opportunity to provide additional information and to clarify issues cited in the petition. The transcript for that meeting is available in ADAMS under Accession No. ML16215A391.
In the supplemental letter dated January 10, 2017, the petitioner withdrew the first two requested enforcement actions, citing the plant shutdown agreement reached between the licensee and the State of New York, and documents released by the NRC in response to a Freedom of Information Act request (FOIA/PA-2016-0457).
The NRC sent a copy of the proposed director's decision to the petitioner and the licensee for comment on January 11, 2017 (ADAMS Accession Nos. ML16320A269 and ML16320A273, respectively). The petitioner and the licensee were asked to provide comments within 30 days on any part of the proposed director's decision that was considered to be erroneous or any issues in the petition that were not addressed. Comments were received from the petitioner and the licensee and are addressed in the final director's decision. In the licensee's response dated February 9, 2017 (ADAMS Accession No. ML17045A470), new information was provided to the NRC staff that was not available when the proposed director's decision was issued for comment. The licensee's response (1) provided detailed information on the enhanced BFB inspection plans for the remaining refueling outages, (2) provided the results of the BFB failure analysis performed at the Westinghouse hot lab testing facility, and (3) informed the NRC staff that the licensee had changed its commitment and would not perform the downflow to upflow modification at either of the Indian Point operating units.
Notwithstanding the fact that the petitioner withdrew the first two requested enforcement actions, the Director of the Office of Nuclear Reactor Regulation has determined that the petitioner's request to (1) issue an order requiring that Indian Point 2 inspect the BFBs during the spring 2018 refueling outage would have been denied because the licensee committed to take this action, and the staff retains the option to take enforcement actions if necessary, (2) issue a demand for information requiring Indian Point 3 to perform an operability determination was effectively met inasmuch as the licensee performed the evaluation and made it available to NRC inspectors as part of the NRC's reactor oversight program, and (3) issue a demand for information requiring the licensee to provide an evaluation of the operating history of the metal impact monitoring system be denied because the system has no operability or regulatory requirements, loose baffle-former bolt heads would be expected to remain in place due to the tight clearances between the baffle plate and fuel assemblies, thus making bolt failures very difficult to monitor using this system, and the staff finds no basis to require such information for a nonsafety system. The reasons for this decision are explained in the director's decision (DD-17-01) pursuant to 10 CFR 2.206 of the Commission's regulations.
The NRC will file a copy of the director's decision with the Secretary of the Commission for the Commission's review in accordance with 10 CFR 2.206. As provided by this regulation, the director's decision will constitute the final action of the Commission 25 days after the date of the decision unless the Commission, on its own motion, institutes a review of the director's decision in that time.
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Dated at Rockville, Maryland, this 13th day of April 2017.
For the Nuclear Regulatory Commission.
William M. Dean,
Director, Office of Nuclear Reactor Regulation.
[FR Doc. 2017-08015 Filed 4-19-17; 8:45 am]
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