ML20126K101

Sanitized Version of Investigation Rept Q-1-84-035 Re Alleged Failure to Make Significant Deficiency Rept Under 10CFR50.55(e)

ML20126K101
Person / Time
Site:	Millstone
Issue date:	11/25/1985
From:	Christopher R, Matakas R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20126J928	List: IA-92-163, Sanitized Version of Investigation Rept Q-1-85-017 Re Alleged Willful Violation of TS Concerning Rod Insertion Limits ML20126J925 ML20126J967 ML20126K007 ML20126K038 ML20126K101 ML20126K140 ML20126K178
References
FOIA-92-163 Q-1-84-035, Q-1-84-35, NUDOCS 9301070027
Download: ML20126K101 (3)

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Title:

Millstone Nuclear Generating Station, Unit 3 Alleged Failure to Make Significant Deficiency Report Under 10 CFR 50.55(e)

Licensee: Case Number:_ Q-1-84-035 Northeast Nuclear Energy Company Report Date: November 25,--_1985 P. O. Box 270 Hartford, Connecticut 06101 Control Office: 01:RI Status: Closed Inquiry Reported By: Approved By:

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[N N J* h . \ _, _ l?, ll > > V . ?l ~ ? AA' Rtch'a'rd A. Matakas, Investigtitor R.' Keith Christopher, Dirdct6r Office of Investigations Office of Investigations-Field Office, Region I Field Office, Region I l

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DETAILS OF INQUIRY Purpose of Inquiry This Inquiry was initiated on December 13, 1984, based on a request from the Executive Director for Operations, to investigate the licensee's (Northeast Nuclear Energy Company) failure to report a potential significant deficiency in the construction of the Millstone 3 Nuclear Generating Station relating to the design of the load path for the reactor coolant pump (P1) snubber support.

(Exhibit 1)

Background

Exhibit i reported that the licensee " failed to inform the Commission of a-design error discovered in 1981 during construction of Millstone 3" that "should have been reported as a deficiency under 10 CFR 50.55(e)." Exhibit.1 advised that the design error "could require extensive redesign or repair to establish the ad m m y of the snubber support to perform its intended safety function," and requested 01 to review the matter to, among other things, determine what justification, if any, was prepared for the licensee's decision not to report the error to the NRC.

Review of Correspondence Between the Licensee and NRR By letter dated October 18, 1984 (Exhibit 2), the licensee informed NRR that a recent analysis determined that the P1 snubber on the reactor coolant pumps in-the A and B cubicles, as currently designed and installed, were acceptable to meet LOCA load requirements.

By letter dated November 8, 1984 (Exhibit 3), NRR requested the licensee to, among other things, provide the Commission with t .' assurance the licensee had, which' determined that the design of the components and supports in question, were acceptable as installed.

By .55|letter dated December 7,1984]] (Exhibit 4), the licensee responded to NRR's-request, SUPRA, adding that the " error in the calculation was not considered to be a potential significant deficiency under 10 CFR 50.55(e)(1)(ii)" since-the calculation error was discovered while the design was still in the veri-fication stage and was not yet considered final. The licensee further stated.

that "such in-process issues are not considered to be reportable."

Contact with Robert J. BOSNAK, Acting Assistant Director for Components and Structures Engineering, Division of Engineering On October 4, 1985, the reporting Investigator met with B0SNAK to discuss information contained in Exhibits 1-4 He indicated that he was not familiar with Exhibit 4, SUPRA, wherein the licensee addressed the 50.55(e) issue.

Following discussions with James LIEBERMAN, Office of the Executive Legal Director (ELD), and Robert HEISHMAN, Reactor Construction Programs Branch (RCPB), the reporting Investigator was informed that, in view of the informa-l tion contained in Exhibit 4, additional IE review and/or inspection effort i would be needed to determine if a regulatory violation had occurred which-1

should be referred to 01,for investigation. BOSNAK indicated that the licensee's hardware design had been recently reviewed and accepted by NRR. He said, as it now stands, the Millstone 3 snubber load path is not a hardware safety problem.

Closure Information This Inquiry is closed pending the Staff's evaluation as to whether this issue constitutes violation of either 10 CFR Part 21 and/or 10 CFR Part 50.55(e).

(Exhibit 5 pertains)

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