ML19260B223

LER 79-020/01T-0:on 791115,during Refueling Shutdown,Anchor in Steam Generator a Auxiliary Feedwater Line Contained Const Deficiency.Caused by Improper Installation of Three Bolts in Anchor Support Plate.Anchor Permanently Modified

ML19260B223
Person / Time
Site:	Point Beach
Issue date:	11/27/1979
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
Shared Package
ML19260B220	List: ML19260B219 ML19260B223
References
LER-79-020-01T, LER-79-20-1T, NUDOCS 7912070433
Download: ML19260B223 (2)

Text

NRC FORM 366 U. S. NUCLEAR REGULATORY COMMISSION (7 77)

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. LICENSEE EVENT REPORT (PLEASE PRINT OR TYPE ALL REQUIRED INFORMATION)

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ET NUMBER EVENT DESCRIPTION AND PROBABLE CONSEQUENCES h o 2 l During a refueling shutdown, an engineering investication and analvsis o 3 lin accordance with IE Bulletin 79-02 indicated that an anchor in the I o la l"A" SG auxiliary feedwater line contained a construction deficiency. An I o 3 las-built analysis of the anchor showed that it was capable of performingt a s lits support function, but not its restraint function. Failure of this I o17 l support could result in loss of auxiliary feed to the "A" SG, but would l o a Inot violate the steam generator pressure boundary. I 7 8 9 80 SYSTEV CAUSE CALSE COVP. VALVE COOE CODE SUBC0CE CCMPCNENT CCDE SLBCCOE SueCODE

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, no l 7 l 9l l-l l 0l2 l 0l l/] l Ol ll lTl l-l l0l 2' 22 23 24 .6 27 28 29 30 31 32 TAKE T ON ON PLANT T dOURS 22 S 8 IT D F R1 8. S PDL E YANUFACTLRER l3J F jg Z lg34 35 Zlg 36 Z[@ 31 0l0l0l [ Y lg 41 N lg l A [g B 1 3l0lg 40 42 43 44 47 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS i o lThree of the six bolts in the anchor succort plate were imorocerly 1 11 11 Ilinstalled, thus the anchor did not provide a desian safety factor of I

,,, l greater than 2.0. The anchor has been permanently modified such that I i , ] design safety factor is now greater than 4.0 and is now fully capable off i 4 [ performing its required function. I 7 a 9 90 ST S a6 POWER OTHER STATUS DIS V RV Of SCOVERY DESCRIPTION i s l H @ l 01 Ol 0l@lN/A l l C l@l IE Bulletin 79-02 Testing A TlVITY CO TENT RELEASED OF RELE ASE AVOUNT OF ACTIVITY LOCATION OF RELEASE l N/A

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ATTACHMENT TO LICENSEE EVENT REPORT NO. 79-020/0lT-0 Wisconsin Electric Power Company Foint Beach Nuclear Plant Unit 1 Docket No. 50-266 On November 15, 1979, with Unit 1 shut down for refueling, an engineering evaluation of an anchor on the "A" steam generator auxiliary feedwater line inside containment indicated that the anchor did not meet the design safety factor requirements. The analysis, performed on the as-built condition of the anchor, followed discovery of a ccustruction deficiency in that three of the six bolts in the anchor plate were improperly installed. The investigation and analysis was conducted in accordance with the requirements of IE Bulletin 79-02.

The analysis of the as-built condition of the anchor showed that it had a design safety factor of less than 2.0 and, thus, could not be assumed capable of fully providing its restraint function, although it would provide its support function. An engineering judgment, conservatively assuming that the restraint in question was not provided, predicted unacceptable piping stresses (i.e., not within yield).

The potential consequences of the postulated failure of :he " A" steam generator auxiliary feedwater line are reduced by the fact that the deficient anchor is upstream of two-line check valves, with one rigid and two spring hangers in between. Also, the anchor is located outside of the steam generator missile shield.

These two facts reduce the potential of violating the steam generator pressure boundary. Further, the line to the "A" steam generator can be isolated upstream of the anchor and all auxiliary feedwater flow directed to the "B" steam generator. No problems were identified in the "B" steam generator auxiliary feedwater line anchors.

The anchor has been permanently modified by providing additional support such that it is now fully capable of performing its required function; the safety factor is now greater than 4.0.

Because of differing equipment and piping layouts, the support in question is unique to Unit 1; a comparable support does not exist in Unit 2.

This discovery is reportable per Technical Specification .

15.6.9.2.A.9.

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