ML20010C611

Part 21 Rept Re Potential for Stress Corrosion Cracking of Anchor Bolts,Originally Reported on 810813.Utils That Have Plants for Which Vendor Has Supports & Restraints Design Responsibility Will Be Notified of Problem

ML20010C611
Person / Time
Site:	North Anna, Farley, 05000514, 05000515, 05000349
Issue date:	08/14/1981
From:	Taylor J BABCOCK & WILCOX CO.
To:	Stello V NRC OFFICE OF INSPECTION & ENFORCEMENT (IE)
References
REF-PT21-81-418-000 PT21-81-418, PT21-81-418-000, NUDOCS 8108200238
Download: ML20010C611 (4)

Text

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'** fo m y h ~594/M iBabcock & Wilcox- uuei..r Power Gener. tion Division ~

a McDermott company 3315 Old Forest Road P.O. Box 1260 Lynchburg, Virginia 24505 (804) 384-5111 August 14, 1981 us ,b.

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Mr. Victor Stello, Director k //

0ffice of Inspection and Enforcement j ,;

U. S. Nuclear Regulatory Commission -

4OC ]g// i Washington, DC 20555 C N

Subject:

10 CFR 21. Report

Dear Mr. Stello:

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Pursuant to the requirements of 10 CFR 21, this report in three copies is made concerning the potential for stress corrosion cracking of anchor bolts. This concern is judged reportable for the Bellefonte Units 1 and 2 of TVA and the Pebble Springs Unit of PGE.

i Mr. J. C. Deddens, acting for Mr. D. E. Guilbert, Vice President, Nuclear Power Generation Division, B&W, was informed of this reportable concern 1507 hours0.0174 days <br />0.419 hours <br />0.00249 weeks <br />5.734135e-4 months <br />, August 13, 1981. .

This letter confirms our telephonic report to Mr. W. R. Rut.herford of your office on August 13, 1981 at approximately 1523 hours0.0176 days <br />0.423 hours <br />0.00252 weeks <br />5.795015e-4 months <br />, The attachment presents the necessary information reintive to this CollCern.

Ver ly yours, V / f , O'L J. H. Taylor Manager, Licensing JHT/fw cc: Mr. R. B. Borsum - B&W Bethesda Office h

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)I 8108200238 810814 PDR ADOCK 05000348 AgW S PDR 7

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Nature of the Concern Stress Corrosien Cracking (SCC) is a potential failure rode for high yield strength

(>120 KSI) low alloy steel bolting material subjected to appreciable steady state loads (generally proload in the case of supports and restraints) and typical reac-tor contairrent corrosive environ:mnts (humid air or borated water fran spills) for extended pericds of tine. Generally, the bolting :raterid for the supports ard restraints is specified as SA 540 Grade B -22, -23, or ~-24 Class 1 or 2. SA 540 Class 1 bolts have a specified runu::un yield strength (YS) of 250 KSI an.1 speci-fled :rinir:um ultirrate tensile strencth (UIS) of 165 KSI at room te:rp&atare while Class 2 bolts have a specified minimu:n YS of 140 KSI and specified minimura UIS of 155 KSI. These raterials are therefore a concern with respect to SCC.

If significant stress corrosion cracking were to occur the anticipated node of failure could be the snaaen brittle fracture of individual bolts with or with-out the application of external loads to the support or restraint and/or the failure of the entire support or restraint due to failure of several or all of the cttaching bolts when external 1 mas are acplied. Structural failure of stww.Ls and restraints could invalidate assu:rptions made in the dynamic loading analysis of reactor coolant system ca:ponents and in the e:ergency core cooling analysis.

Plant Applicability

'Ihe applicability of the ST concern to B&W plants under construction or in opera-tion is in accordance with the following:

'IVA: (Bellefonte 1/2) ,

(a) B&W designed the supports and restraints.

(b) B&W specified a bolt preload of a minurun of 70% of ultimate tensile strength in rany cases.

(c) B&W provided the design of the non-eied bolts and a portion of the de-sign of the enbed bolts (raterial, diareter, pre-load) .

(d) Bolts are in process of installation.

VEPCo: (North Anna 3) 1 (a) B&W designed the supports and restraints.

(b) B&W has not specified a high bolt preload in cases where preloads are specified and in nest cases, preloads have not bm i specified.

(c) B&W has provided rany of the bolts.

(d) Bolt installation is not scheduled to occur for several years.

KE: (Pebble Springs)

(a) B&W designed the supports and restraints.

(b) B&W has specified a bolt preload by specifying " turn-of-the-nut" nethod (See Feference 1) in rany cases, which assures that.at least 70% of ultirate tensile strength is achieved.

(c) B&W has provided many of the bolts.

(d) Bolt installatica is not sc eduled to occur for several years.

For all these plants, the bolts of concern are both cieds (ie., erbodded in con-crete) and external joints.

Ecference 1 - Steel Structures, Design and Behavior, C. G. Salnen and J. E. Johnson, Harper and Few, 1971, pages 92-94.

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l WPPSS ~ (NNP 1/4)

.(a) B&W did not design the supports and restraints.

'(b)' D&W specified the raterial for anchor bolts and other bolts, and the bolt preload (1cw value) for the reactor vessel supports.

(c) B&W provided the bcir- for the RV supports. ~

(d) Bolt installation my ce cmpleted.

177 FA Plants In Coeraticn And Under Construction (a) B&W did not design the supports and restraints.

(b) B&W specified high preloads on Peactor Vessel support skirt embed bolts, but did not specify bolt ratenal and did not provide the bolts.

Peportability

'1he concern that anchor bolts that have been specified by B&W to be torqued to a high preload may be subject to stress corrosion cracking when in the reactor con-tainment building atrosphere is doanwl to be reportable to the NBC under 10CFR21 for the TVA and PG plants.

'Ihe concern is not reportable under Part 21 for VEPCo or hTPSS because any bolt' preloads that were specified by B&W were not of a sufficiently high value to cause SOC.

In the case of the remaining B&W plants, which are the 177 FA plants in operation or under construction, B&W is unable to make a determination as to whether a re-portable concern under Part 21 exists. B&W has no information on the design of ..

the supports or restraints, on the design of the bolts or the ma w inin used for -

the bolts, and is therefore unable to determine if the potenM al for SCC exists.

For these plants, B&W will advise these utilities of the concern for their eval-uation. ,

Corrective Iction Corrective action by B&W with respect to the various reactor plants varies because of the differing responsibility B&W has with respect to the design of the supports and restraints and the specificatica of bolt preloads.

1. For the three plants for which B&W has supports and ::estraints design respon-sibility, TVA, PGE, and VEPCo, the following actions have been or will be taken:

(a) 'IVA was advised by B&W cn January 26, 1981 that there is uncertainty as to the proper value to which the bolts should te tightened due to SCC potential; it was reccrimnded that the high torqueing should be stopped and the bolts tightened to " snug" only pending further advice frcm B&W.

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'IVA was given ndaiticnal advice about the SCC concern, including the po-l -

tential adverse safety consequences, on July 1,1981.

(b) PGE and VEPCo will also be advised cn the SCC concern.

2. B&W will inform WPPSS of the potential for SCC in high yield strength 1cw alloy steel bolts for their information.

3. B&W will advise the 177 FA plants in operation and under construction of the 4

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ocricern' for their evaluaticn and for their use in connection with their response to the NPC letter of my 19, 1980 to all Po w r-Reactor Licensees which aMmssed this issue (Referer.ce 2) .

Reference'2'- NPC letter to "All Power Reactor Licensees" dated May 19, 1980, Attachment 1, Part II, Page 3.

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