NRC-87-3280, Forwards Type I & II Ltrs Re Steam Generator Tube Rupture Event,In Response to Request for Info on Generic Implications of Tube Rupture for Use in Preparing Presentation to NRC on 871109.Part 21 Related

Forwards Type I & II Ltrs Re Steam Generator Tube Rupture Event,In Response to Request for Info on Generic Implications of Tube Rupture for Use in Preparing Presentation to NRC on 871109.Part 21 Related

ML20236E917
Person / Time
Site:	North Anna
Issue date:	10/28/1987
From:	Johnson W WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Cheng C Office of Nuclear Reactor Regulation
References
REF-PT21-87-182-000 NS-NRC-87-3280, PT21-87-182, PT21-87-182-000, NUDOCS 8710300063
Download: ML20236E917 (8)

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Westinghouse Power Systems Rn*s$'Igh u Pennsylvania 15230-0355 Electric Corporation NS-NRC-87-3280 l October 28, 1987 0.S. Nuclear Regulatory Commission C. Y. Cheng, Branch Chief Materials Engineering Branch, NRR Bethesda, MD 20814  !

SUBJECT:

Generic Implications of the North Anna Unit 1 Tube Rupture  !

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Dear Mr. Cheng:

Attached is the information requested by your staf f on the subject of the

" Generic Implications of the North Anna 1 Tube Rupture" for use in the preparation of your presentation to the Nuclear Regulatory Commission on November 9, 1987.

Specifically, a copy of the " Type I" and " Type II" letters transmitted by Westinghouse to utilities to advise them of the issue of fatigue cracking of <

dented steam generator tubes is provided. The " Type I" letter was sent to utilities with potentially affected plants; the " Type II" letter was sent to utilities which are not believed by Westinghouse to be affected. Also, figures which provide a comparison of stability ratios (with/without the consideration of void dependent damping) between North Anna Unit 1 and other domestic operating steam generators are included. Finally, a listing of all i Westinghouse communications to the industry regarding the tube rupture event at North Anna Unit 1 is furnished.

Should you require additional information, please contact Mr. Carl W. Hirst of 1 my staff at 412-374-4311.

l Very truly yours, i W 4 J hnson, Manager No ea- Safety Department cc: C. H. Berlinger, Chief Generic Communications Branch, NRR

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4 September 22,1987 L. .

1 Type i Letter -

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No l Steam GeneratorTube Rupture Event Westinghouse Preliminary Assessment Summary of North Anne Experience I On My 15,1987, a steam generator tube rupture event ooourred at North Anna Unit 1 shortly after reaching 100% power. For several days prior to the event, air ejector radiation monitor readings were erratic. However, grab samples were taken prior to the tube rupture for environmental release calculations. l t

Subsequent analysis of this data indicated that increasing primary to secondary leakage occurred over a N -

36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> period, which was below normal technical specification limits, prior to the tube rupture event. The

-l ruptured tube was located in Row 9 Column 51 in steam generator "C". The leakage location was found to

[ be at the top tube support plate on the cold leg side of the tube. The opening was circumferentialin i

orientation and extended 360' around the tube. - A preliminary assessment indicates that this event does

' not represent an unreviewed safety issue.

I Fallure Mechanism Explanation The cause of the tube ruptura has been determined to be high cycle fatigue, The source of the loads is believed to be a combination of a mean stress levelin the tube and a superimposed altemating stress. The mean stress is produced by denting of the tube at the top tube support plate and the aftemating stress is due to out of plane deflection of the tube above the top tube support caused by flow induced vibration.

Denting also shifts the maximum tube bending stress to the vicinity of the top tube support plate. These ,

loads are sufficient to produce fatigue and are consistent with a lower bound fatigue curve for the tube materialin an AVT water chemistry environment. The magnhude of the ahomating stress is consisterd whh a j fluidelastic tube vbration mechanism. f The most significant contributor to the occurrence of excessive vbration is the reduction in damping at the tube to tube support plate interface caused by the denting. The absence of antivlbration bar (AVB) aupport is necessary for requisite vibration to occur, together with the reduction in damping, The presence of AVB aupport will restrict tube motion and thus preclude the deflection amplitude required for fatigue.

The original design configuration required AVBs to be inserted to Row 11. Inspection data has shown that some AVBs in the North Anna Unit 1 steam generators penetrate to Row 8, exceeding the minimum AVB j insertion depth requirement. No AVB support was present for the Row g Column 51 tube that ruptured. )

' Also contributing to the level of vbration, and thus loading, is the local flow field associated whh the detailed geometry of the steam generator. The ruptured tube is considered to have a worst case combination of l loading conditions and fatigue properties.

The prerequisite conditions dortved from the evaluations were concluded to be:

Estlaup Spouframenta Prereculalle Conditions ,

Mean Stress

• Dentin 0 l Allemating Strees . Tube Vbration

• DentingattopTSP

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• HighI.ocalFluid Forces ..

• Absence of AVB oupport Material Fatigue Properties . AVT Environment

. l.ower range of properties i i

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September 22,1987 Type I!.etter l

Steam Generator Evaluation Criteria A general criteria for operating steam generators has been developed based on the North Anna experience. Any steam generators shich have flow conditions and tube support conditions which are less conservative than the modification criteria established for Row 8 through 12 at North Anna Unit i should be evaluated.

Yhe lolowing conditions make up this crteria: .

The stoam Generator should be evaluated ll the followin0 conditions (1) & (2) are met (1) Dentin 0 must be present at the top tube support plate,

.and. ,

l (2) the burde flow parameters must be hieher than 90% of those at North Anna prior to modification This second condition is satisfied if ather (a) the bundle flow is higher than 90% of North Anna, l

.or...

(b) the fluidelastic vbration stabi!!!y ratio is hi 0her than 90% of North Anna.

This criteria is preliminary and believed to be conservative. H is important to note that the North Anna unks have the highest bundle flow parameters of the 51 Series steam generators. If a steam 0enerator were found to exceed this criteria, h may still be less severely loaded than the North Anna units prior to j

modification; yet, further detalled evaluations would be necessary to determine the potential for a similar tube rupture event.

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Recommendations Based on the records available to Westinghouse it appears that your plant exceeds the criteria because (1) denting is believed to be present at the top tube support plate of at least one steam generator and (2) based on nominal power level data the bundle flow parameters are higher than 90% of the North Anna steam generator values prior to the recent field modification. 4 Evaluation Crheria at is recommended that an effort be made to evafuate the plant records describing the condtion of at tubes in Row 8 through 12 in all steam generators. Colled the foDowing data:

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(1) identify tubes wth any denting at the top tube support plate (either hot or cold leg );

(2) Quantify the AVB insertion depths for each column (from eddy current data);

i (3) Note any tube wear at any AVB or top tube support plate intersection..

l - Evaluate Plant i.eek Rate Measurement Systems I W the results of this data collection confirm that the above orkeria are exceeded, then it is also recommended I that the plant systems and practices for determining primary to secondary leak rates be evaluated. Such

) systems and practices should be capable of producing somrate leak rate data that would detect and classify l

• a tube rupture evenlike the one that occurred at North Anna.

A Technical Meeting la Planned ,

The Westinghouse Projects Office will be communicating whh you to assist you in this evaluation. A' customer meeting to discuss the technical detells related to this issue is scheduled for October 16,1987 in Pittsburgh. At that time a detailed presentation will be given describing both the North Anna situation and the potential actions to address and to resolve this lasue.

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September 22,1967 i L,, Type II Letter Steam GeneratorTube Rupture Event Westinghouse Preliminary Assessment

. 4 Summary of North Anna Experience On July 15,1987, a steam generator tube rupture event occurred at North Anna Unit i shortly after reaching 10g% power. For several days prior to the event, air ejector radiation monitor readings were erratic. However, grab samples were taken prior to the tube rupture for environmental release calculations.

Subsequent analysis of this data indecated that increasing primary to secondary leakage occurred over a 24 -

36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> period, which was below normal technical specification limits, prior to the tube rupture event. The {

fuptured tube was located in Row 9 Column 51 in stetm generator "C". The leakage location was found to be at the top tube support plate on the cold leg side of the tube. The opening was otrcumfetentialin orientation at d extended 360' around the tube. A preliminary assessment indicates that this event does not represent an unreviewed safety issue.

Failure Mechanism Explanation The cause of the tube rupture has been determined to be high cycle fatigue. The source of the loads is believed to be a combination of a mean stress levelin the tube and a superimposed altomating stress. The J mean stress is produced by denting of the tube at the top tube support plate and the attemating stress is due to out of plane deflection of the tube above the top tube support caused by flow induced vibration. ,

i Denting also shifts the maximum tube bending stress to the vicinity of the top tube support plate. These I

loads are sufficient to produce fatigue and are consistent with a lower bound fatigue curve for the tube materialin an AVT water chemistry environment. The magnitude of the themaling strate is consistent witifa' fluidefastic tube vbration mechanism.

The most sigrdficant contributor to the occurrence of excessive vbrallon is the reduction in damping at the tube to tube support plate interface caused by the denting. The absence of antivibration bar (AVB) support is necessary for requisite vibration to occur, together with the reduction in damping. The presence of AVB support wl:1 restrict tube motion and thus preclude the deflection amplitude required for fatigue.

The original deston configuration required AVBs to be inserted to Row 11. Inspection data has shown that some AVBs in the North Anna Unit 1 steam generators penetrate to Row 8, exceeding the minimum AVS insertion depth requirement. No AVB support was present for the Row 9 Column 51 tube that ruptured.

Also contributing to the level of vbration, and thus loading, is the local fbw field associated whh the detailed l

geometry of the steam generator. The ruptured tube is considered to have a worst case combination of I loading conditions and fatigue properties.

The prerequisite conditions derived from the evabat!ons were concluded to be:

Fallous Raoulramenta Prerecula'te Conditbat Mean Stress . Denting Artemating 8 tress . Tube Ybration

• Denting at topTSP .

• HighLocalFluidForces

• Absenceof AVBsupport

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Materta1 Fatigue Properties . AVT Erwlmnment l

. Lower range of properties ,

l I Steam Generator Evaluation critoria A general criteria for operating steam generators has been developed based on the North Anna experience. Any steam generators which have flow conditions and tube support conditions which are less conservative than the rnodifbstion criteria established for Row 8 through 12 at North Anna Unit 1 should be ,

evaluated.

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7, ) Septembe? 22,1987 l Type 11 Letter

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The following condtions make up this critoria: ,

l The steam generator should be evalosted ll the following condtions (1) & (2) are met (1) Denting must be present at the top tube support plate,

... and. '

(2) the bundle fbw parameters must be higher than 90% of those at North Anna prior to modificatbn This second condition is satisfied Il either I h) tw bundle flow is higher than 90% of North Anna, J 1

(b) the fluidelastic vbratbn stabaty ratio is higher than 90% of North Anna.

This orheria is preliminary and believed to be conservative, it is important to note that the North Anna unts  ;

have the highest bundle flow parameters of the 51 Series steam generators, if a steam generator were found to exceed thIs criterla, it may still be less severely loaded than the North Anna units prior to modification; yet, further detailed evaluations would be necessary to determine the potential for a similar tube rupture evert ,

Recommendations Based on the records available to Westinghouse it appears that your plant falls below the criteria either because (1) denting is not bell 6ved to be present at the top tube support plate in any of your steam generators or (2) the bundle fbw parameters in your stoam generators are less than 90% of the North Anna a sam generator values prior to the recent field modification.

Confirm That There le No Dentin 0'at the Top Tube Support Piste As a precaution, k is recommended that the oddy current inspection records for each steam generator be reviewed to confirm that no tube denting is present at the top tube support plate. Denting is believed to be j necessary to produce a relauvely high mean stress which reduces the tube fatigue endurance timt, to shift

. the maximum tube bending stress to the vicinity of the top tube support plate and for large amplitude tube vbration to occur.

Actions Recommended for Other Plante' '

Other plants, with steam generators which exceed the ortteria, have been notified to evaluate their plant records descrb!ng the condttion of all tubes in Row 8 through 12 in all steam generators. Collectbn of the following data has been recommended:

(1) Identification of all tubes with any denting at the top tube support plate (other hot or j cold leg);

(2) Quantification of the AVB insettlon depthe for each column (from eddy current data); )

(3) Notation of any tube wear at any AVB of top tube support plate Intersection.

If the resuts of this data collection confirm that the above crtteria are exceeded. then t is also being recommended that the plant systems and practices for determining primary to secondary leak rates be evaluated. Such systems and practices should be capable of producing accurate leak rate data that would detect and classify a tube rupture event ske the one that occurred at North Anna. ,

' A Technical Meeting le Planned ,

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The Westinghouse Projects Office wlH be communicating with you to keep you informed about any further l l I developments relating to your steam generators. A customer meeting to discuss the technical details

[ related to this lasue is scheduled for October 16,1987 in Pittsburgh. At that time a detailed presentation wit! l

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be given describing both the North Anna situation and the potential acDons to address and to resolve this I issue. l l

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1. Evaluation based upon nominal power capability parameters and plant data where available. ,

2. Denting status based upon Westinghouse information of known conditions.

3. Configuration based upon Westinghouse records and plant data where available. -

4. Revision from initial evaluation based on refined evaluation

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3. Configuration based upon Westinghouse records and plant data where available. ,

'4. Revision from initial evaluation based on refinad evaluation.

5. All type 3 8.G.'s have normalized stability statios less l than 0.7.

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WESTINGHOUSE COMMUNICATIONS j TO THE INDUSTRY REGARDING THE, S.G.T sR. AT NO. ANNA 1 j

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i e MULTIPLE NRC MEETINGS i (JULY - SEPTEMBER 1987) e PRESENTATION TO THE E OWNERS GROUP (SEPTEMBER 1987) e NOTIFICATION LETTER TO M PLANT OWNERS

_ . (SEPTEMBER 1987)

J e PRESENTATION TO EPRI SPONSORING UTILITIES l (OCTOBER 13, 1987) ,

l e DETAILED TECHNICAL MEETING FOR E PLANT OWNERS

_' - (OCTOBER 16, 1987) 6 e

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