ML19274D331

Forwards Rept GA-A15094:Effect of an Aluminum Alloy on Sanicro 31 at Elevated Temps.

ML19274D331
Person / Time
Site:	Fort Saint Vrain
Issue date:	01/17/1979
From:	Wessman G GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER
To:	Gammill W Office of Nuclear Reactor Regulation
Shared Package
ML19274D332	List: ML19274D331 ML19274D338
References
NUDOCS 7901230196
Download: ML19274D331 (2)

Text

NERAL ATOM So"INNe"?S"* **"""*

cm catmmru sms Q-0 January 17, 1979 Mr. William P. Gamill Assistant Director of Standardization and Advanced Reactors Division of Dreject Management U. S. Nuclear Regulatory Comission Washington, D.C. 20555

Subject:

FSV Steam Generator Leak of November 30, 1977.

References:

(1) GAC letter, G. L. Wessman to R. P. Denise, dated December 21, 1977.

(2) NRC letter, R. P. Denise to D. P. Lessig, dated January 6,1978.

(3) GAC letter, G. L. Wessman to R. P. Denise, dated January 30, 1978.

(4) NRC/PSC/GAC Meeting of April 19 and 20,1978, in Bethesda.

Dear Mr. Gammill:

Reference (1) notified the Commission that General Atomic Company was investigating the relationship of an aluminum chip which was dropped during fabrication in Module B-1-1 to the steam generator leak that. occurred in that same module on November 30, 1977. NRC asked a number of questions in Reference (2) which were responded to, for the most part, by Reference (3).

The status of the steam generator leak investigation was discussed in the Reference (4) meeting. At that time, it was stated the Commission would require a final report on the metallurgical examination to close out the records on the steam generator leak.

Enclosed are fifty copies of GA-A15094 which is the final report on the metallurgical examination of the potential for the aluminum chip caus-ing the steam generator leak. It should be noted that the conclusion of the investigation is that cracking as a result of liquid metal enbrittle-ment is highly unlikely at the environmental conditions that have or will exist in the future. Furthermore, it is concluded that it is unlikely that the aluminum chip was responsible for the tube leak experienced.

790123 O(96 f5$

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, 4 Mr. William P. Gammill . January 17, 1979 Reference (3) stated a review of the steam generator records for Module B-1-1 was in progress and at that time revealed nothing which would explain the cause of the leak. The records showed no weld repairs required in super-heater II section of subheader F of Module B-1-1, nor did they identify any weld defects which were accepted as is. This review of Module D-1-1 plus a review of the other eleven installed steem generator modules is now com-plete. This review indicated '. hat there was nothing unusual about Module B-1 subheader F of that uodule. Further, the review did not reveal anythir., in other ir. stalled . nodules which could be judged substandard work-manship or quality and, as such lead to steam generator tube leaks.

Based upon the above, we conclude that the investigation is complete and the steam generator leak of November 30, 1977, was a random failure which was postulated and allwed for in the Fort St. Vrain design (FSAR Sections 4.1.5 and 4.2.4.1).

We would be pleased to discuss any questions you may have regarding these conclusions.

Sincerely,

, .L W G. L. Wessman, Director Plant Licensing Division GLW:mk