ML20235U027

Part 21 Rept Re Westinghouse Plant Steam Generator Tubes Susceptible to High Cycle Fatigue Failure Similar to 870715 Event at North Anna Unit 1.Caused by Mean Stress in Tube & Superimposed Alternating Stress

ML20235U027
Person / Time
Site:	Point Beach, North Anna, 05000000
Issue date:	10/07/1987
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
CON-NRC-87-103, REF-PT21-87, REF-PT21-87-181-000 PT21-87-181, PT21-87-181-000, VPNPD-87-427, NUDOCS 8710130328
Download: ML20235U027 (2)

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Wisconsin Electnc rom come J 231 W. MICHIGAN,P.O. BOX 2046. MILWAUKEE W153201 (414)277-2345 VPNPD-87-427 NRC-87-103 October 7, 1987 U.S. NUCLEAR REGULATORY COMMISSION- i Document Control Desk l Washington, D.C. .

20555 j l

Gentlemen:

DOCKET 50-301 l SUSCEPTIBILITY TO HIGH CYCLE FATIGUE FAILURE l OF STEAM GENERATOR TUBES i POINT BEACH NUCLEAR PLANT, UNIT 2 l i

The purpose of this letter is to inform the Nuclear Regulatory Commission that Wisconsin Electric Power Company has been notified by Westinghouse Electric Corporation that Point Beach a Nuclear Plant Unit 2 may be potentially susceptible to high {

cycle fatigue of steam generator tubes.similar to that which is believed to have caused the steam generator. tube rupture event at North Anna Unit 1 on July 15, 1987. This determination is based on a conservative and preliminary evaluation conducted by' ,

Westinghouse.

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The loads causing high cycle fatigue are a combination of a mean stress in the tube and a superimposed alternating stress. The i i

mean stress is produced by denting of the tube at the top support plate and the alternating stress is due to out-of-plane deflection of the tube above the. top tube support caused by flow induced vibration. These loads are consistent with a lower bound fatigue curve for the tube material in an AVT water chemistry and are sufficient to produce fatigue. The most  !

significant contributor to the tube vibration is the reduction l in damping caused by denting at the tube-to-top support plate '

interface. The absence of antivibration bar (AVB) support is necessary for this vibration to occur. Conversely, the presence ,

of AVB support will restrict tube motion and thus preclude the deflection amplitude required for fatigue.

The notification that Point Beach Nuclear Plant Unit 2 may be susceptible is based on the belief that conditions in Unit 2 steam generators meet prerequisite criteria derived from the evaluations of the North Anna event. These criteria are:

(1) Tube denting at the top support plate; and .

(2) Tube bundle flow parameters which are greater than 90%

of those at North Anna.

8710130328 871007 'JN0 I EDR ADOCK 05000301 0 l 3 p. PDR .i

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Document Control Desk  !

October 7, 1987 Page 2 1

The tubes which may be susceptible are those located in Rows 8 through 12. Tubes located in Rows 1 through 7 are considered'to l be sufficiently rigid to preclude the need for additional 1 support. By design, tubes located in Rows 13 through 45 should l have AVB support and are not susceptible to vibration.

In order to better assess the potential for high cycle fatigue j of Point Beach Unit 2 steam generator tubes,.we are performing i an evaluation which goes beyond the preliminary assessment per- )

formed by Westinghouse. The scope of the eddy current testing program planned for the current Unit 2 refueling and maintenance outage has been expanded to identify tubes in Rows 8 through 12 which exhibit denting at the top tube support plate and to i identify the AVB insertion depths for each column in Rows 8 i through 12. Thermal-hydraulic, fatigue and vibration' analyses will also be performed for the Point Beach Unit 2 steam generators in conjunction with the denting and AVB insertion depth information. The combination of the data from the eddy current testing and these analyses will allow us to more l accurately assess which tubes, if any, in the Point Beach Unit 2 l

l steam generators may be susceptible to high cycle fatigue.

3 Preliminary information from both Westinghouse and Virginia Power Company indiates that low levels of primary-to-secondary leakage were detectable in the affected North Anna steam generator in order of one day prior to the event, confirming that the leak-before-break concept applies to tubes degraded by high cycle fatigue. As part of our evaluation, we are reviewing i our procedures for determining primary-to-secondary leakage to confirm our ability to detect and monitor tube leakage should a similar event occur at Point Beach.

We are not aware of any other failures of this type in Westinghouse steam generators and have not had any indications of tube distress due to fatigue in either Unit 1 or Unit 2 steam generetors. In any event, the rupture of a steam generator tube was considered in the design basis for both Units 1 and 2 and the North Anna event does not represent an unreviewed safety issue for Point Beach.

We will keep you informed of the results of our evaluation.

Please contact us if you have any questions in this regard.

Very truly yours,

/

i Chu C. W. Fay Vice President Nuclear Power

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