ML20091L715

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Advises That Westinghouse Method for Evaluation of Steam Generator Tube vibration-induced Fatigue Represents Sound Engineering Approach & If Properly Implemented Will Provide Reasonable Assurance That Event Will Not Recur
ML20091L715
Person / Time
Site: North Anna 
Issue date: 06/13/1988
From: Wambsganss M
ARGONNE NATIONAL LABORATORY
To: Murphy E
Office of Nuclear Reactor Regulation
Shared Package
ML20090B470 List:
References
FOIA-91-106 NUDOCS 9201280193
Download: ML20091L715 (9)
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Text

{{#Wiki_filter:. 1 l l ARCONNE NATIONAL LABORATORY 9K)0 Soua CAss Amu. Ancion.lilrds 60439 Telephone: 312/972-6144 June 13, 1988 \\ Mr. E. Murphy Of fice of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Murphy:

Subject:

Westinghouse Method for Evaluation of Steam Generator Tube Vibretion-Induced Fatigue Argonne National Lnboratory Ta8 bee n providing consultation to the U.S. NRC in the review of the North Anna Unit I steam generator tube failure assessment and the evaluation methodology developed by Westinghouse. The current status of the Westinghouse method is given in WCAP-ll?99 which documents the application of the methad to evaluate the steam generator tubes of Beaver Valley Unit I nuclear station. The review took place over a period of several months (September 1987 to date) that included the time period during which the Westinghouse method was under development. A'nong other things, as part of the review Argonne called attention to a number of uncertainties related to the inherent e mpicxit ies of the physical situation and to associated inadequacies in state-of-the-art modeling and analysis techniques. Upon review of the re ce n t assessment of the Beaver Valley Unit 1 ~ steam generator tubing (WCAP-ll799., it is concluded (based on documentation, laboratory visit, and correspondence) that Westinghouse has satisfactorily a identified and appropriately addressed the major uncertainties associated with the test and analysis program; as an example, See Sections 8. 2 - 8.7 of WCAP-11799. As a result, it is farther concluded that the Westinghouse method as presently developed represents a sound engineering approach that, if properly impleuented, will provide reasonable assurance that future steam generator tube failures of the type experienced at North Anna Unit 1 are not likely to occur. Since rely, M 7~ M. W. Wambsganss Materials and Components Technology Div. Mu': cml 9201280193 910621 PDR FOIA WILLIAM 91-106 PDR s Oprcmd b. Ik bv.nsn d Or,vp low h LWird Swts Dipwimm d Evro

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J/R/A ASSOCIATES Regulatory information & Support Systems 1407 Marco Drive Mitchellville MD 20721 301/249-9672 March 11,1991 OGW-062 Mr. Donnie H. Grimsley, Director TRE 0M OF iWORN00 Division of Freedom of Information and ,t.ri ki'UEST gp __ q/_ / Publications Services U.S. Nuclear Regulatory Commission // ~/ h/ g_ Washington, DC 20555 SUDJECT: FREEDOM OF INFORMA110N ACT REQUEST

Dear Mr. Grimsley:

Pursuant to the Freedom of Information Act, 5 U.S.C. 6552 and NRC's regulations, I am herein requesting that all documents, not previously released to the public via the Public Document Room, re the North Anna Unit 1 and Ginna steam generator tebe rupture events, be placed in the Public Document Room. I would appreciate your prompt response within ten working days of the receipt of this letter, as provided by the Code. I will be responsible for costs associated with processing this request, however, please provide an estimate if the fees exceed 550.00. I can be reached at 301/249-9672, if there are questions. Tnank you. 5 ncerely, fill lin A/.h6 w Ophe ia G. Williams n

t /j.%s. UNITED STATES f' - "f} NUCLE AR REGULATORY COMMISSION

c. ~,n, c e w gg'.,,,,;

c p AP. 4 1981 t MEMORANDUM FOR: Thomas H. Novak, Assistant Director for Operating Reactors Division of Licensing FROM: V. S. Neonan Assistant Director for Materials & Qualifications Engineering Division of Engineering GINNA STATION STEAM GENERATOR TUBES INSERVICE INSPECTION

SUBJECT:

PRO EAM FOR THE 1980-1989 INTERVAL (TAC #43196) Plant Name-R. E. Ginna Unit No. 1 Suppliers: Westinghouse; Gilbert Associates Docket Number: 50-E44 Responsible Branch and Project Manager:. ORB #5; R. P. Snaider Reviewer: D. T. Huang Description of Task: Review of Ginne Station's Steam Generator Tubes Inservice Inspection Program for tne 1980-1989 Interval Review Status: Additional Information Needed The Inservice Inspection Section of the Materials Engineering Branch, Division of Engineering has rev'.ewed that portion of Rochester Gas and Electric Corporation's submittal dated November 6,1980 regarding the Ginna Station Stean Generator Tubes Inservice Inspection Program for interval. We conclude on the basis of our review that the 1980-1989 the Ginna Station Steam Generator Tube Inservice Inspection Program for the 1980-1989 Interval is acceptsble if the followir.g two conditions are met:

1) Plugging limit for the ten test sleeves be established, r
2) Two typographical errors mentioned in our Safety Evaluatrion be corrected.

Our Safety Evaluation is attached. / I S. WJona irector MateriaT & Qualification s Engineering Division of Engineering

Contact:

D. T. Huang X27377 cc: R. H. Vollmer G. Johnson D. G. Eisenhut W. S. Hazelton R. A. Purple R. P. Snaider V. S. Noonan E. L. Murphy T. H. Novak D. T. Huang

5. S. Pawlicki

/ / V. Benaroya I _~))Ti (l [/ ydj i.6 }7 f b m,

I }< R. E. GINNA NUCLEAR POWER PLANT REVIEW OF THE STEAM GENERATOR TUBE INSERVICE INSPECTION PROGRAM FOR THE 1980 TO 1989 INTERVAL. SAFETY EVALUATION REPORT MATERIALS ENGINEERING BRANCH INSERVICE INSPECTION SECTION INTRODUCTION By letter dated November 6,1980, Rochester Gas and Electric Corporation (the licensee) submitted the "Ginna Station Inservice Inspection Program for the 1980-1989 Interval" for review. Changes.have been incorporated into Paragraph 5.7.1.1 and 6.5 of the Ginna Inservice Inspection Program in order to permit the installation of a maximum of ten test sleeves in steam generator tubes which would otherwise require plugging. These changes permit sleeving instead of plugging of tubes with unacceptable defects. EVALUATION We have reviewed the licensee's submittal dated November 6,1980 regarding the Ginna Inservice Inspection Program. Based upon our review, we conclude that this inspection program meets the recommendations of Re,gulatory Guide 1.83, " Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes," Revision 1 and the requirements of Secticn XI of ASME Code with respect to the inspection methods to be used, provisions ~ for a baseline inspections, selection and sampling of tubes, and inspection intervals. However, the Ginne Inservice Inspection Program is incomplete with respect to the installation of ten test sleeves in steam generator tubes, since it does not contain actions to be taken in the event defects are identified in the test sleeves, e.g. plugging limit for the ten test sleeves. Furthermore, two typographical errors should be revised to convey the same meaning regarding scope of l

( ( . n. A steam generators inservice inspection as originally intended. They are as follows: 1. The reference nunber in Paragrapn 3.5 should be changed to 5 in order to be ccnsistent with the reference list. 2. Paragraph 5.7.1.1 should be revised to indicate that the plant may resume operation only when both conditions (a) and (b) are met. In conclusion, we find that the steam generator tube inservice insp9ction portion of the "Ginna Stat' ion Inservice Inspection Program for the 1980-1989 Interval is acceptable with the condition that plugging limit for the , ten test sleeves be established and the two typographical errors mentioned above be corrected. 9 l.. f .}}

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