ML20153F451

Responds to NRC Bulletin 88-002, Rapidly Propagating Fatigue Cracks in Steam Generator Tubes. Westinghouse Evaluation Indicates Very Low Potential for Fluidelastic Instability

ML20153F451
Person / Time
Site:	Yankee Rowe
Issue date:	09/01/1988
From:	Devincentis J YANKEE ATOMIC ELECTRIC CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
FYR-88-119, IEB-88-002, IEB-88-2, NUDOCS 8809070304
Download: ML20153F451 (5)

Text

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, ,' Tecphon)(617) 872 8100 TWX 710-380-1619 YANKEE ATOMIC ELECTRIC COMPANY y%

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1671 Worcester Road, Framingham, Afassachusetts 01701 September 1, 1988 FYR 88-119 United States Nuclear Regulatory Comission Document Control Desk Washington, DC 20555 References (a) License No. DPR-3 (Docket No. 50-29)

(b) Letter, FYR 88-40, YAEC to USNRC, dated March 25, 1988 (c) Westinghouse Report STD-7.2.3-8154, "Tube Vibration Induced Fatigue Evaluation for the Yankee Plant Steam Generators," dated August 25, 1988 (d) Letter, YAEC to USh1C, dated June 24, 1985

Subject:

Response to NRC Bulletin 88-02: Rapidly Propagating Fatigue Cracks in Steam Generator Tubes

Dear Sir:

Introduction h7C Bulletin 88-02 requested licensees utilizing Westinghouse designed steam generators to evaluate the potential for a steam generator tube rupture. On July 15, 1987, such an event occurred in a steam generator tube at North Anna. The cause was high cycle fatigue leading to rapidly propagating cracking.

Yankee Atomic Electric Company (YAEC) responded to NRC Bulletin 88-02 via Reference (b). In our letter, we comitted to implement an enhanced primary-to-secondary leak rate monitoring program ae an interim compensatory measure. We also stated that the detailed methodology and results of our steam generstor tube analysis would be submitted to the NRC staff at a later date. The additional information is submitted herewith.

Our evaluation of the Yankee Nuclear Power Station (YNPS) steam generators concludes that rapidly propagating tube rupture, similar to that which is described in the bulletin, is not a credible event. The reasons for this are presented in this letter.

Bulletin Action 0.2.(b): Anti-Vibration Bar Penetration Depth Assessment YNPS contains four Westinghouse Model 13 steam generators manufactured prior to 1960. Each steam generator consists of 1,620 U-bend type tubes made from 304 stainless steel. Each tube has an outside diameter of 0.750 inches and a nominal wall thickness of 0.072 inches. This tube thickness is approximately 40% greater than present day Inconel tubes. Each steam generator contains one antivibration bar (AVB).

8809070304 880901 PDR ADOCK 05000029 Q PNV r l g g

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United Statos Nucloce Rcguletory Connission 1

Page 2 FYR 88-119 [

The bulletin requested that the most recent steam generator inspection [

data be reviewed for evidence of denting at the uppermost tube support plate.

YAEC and Zetec personnel reviewed the eddy current data from the most recent

(1987) steam generator inspection. The 1987 Inspection Plan included a full '

length bobbin coil inspection (tube end-to-tube end) of 1.507 tubes in the .

i No. 2 steam generator.

i

] Eddy current data from a total of 380 tubes (Row 5 through Row 10) from '

the No. 2 steam generator were re-examined in 1988 for evidence of denting at the uppermost tube support plate. Several outer rows were also evelue vo to

, determine the AVB profile. The data was first normalized for distance. The.  !

distance of the AVB relative to the top support plate was then measured. The

[

data was reviewed from the fifth support plate cold leg to the fifth support l

, plate hot leg, with the AVB location measured and both support plate  !

j penetrations examined for denting.

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All locations examined at the upper tube support plate show evidence of  !

deposits on the outside of the tubest however, no physical distortion of the l tubes was found. This was expected because the stainless steel tube material used at YNPS has never exhibited physical distortion. Only limitad thinning ,

and pitting have been observed after 27 years of operation. L i

1 The AVB penetrates to Row 8 with complete coverage (see Figure 1).  ;

Approximately one half of Row 7 is supported by the AVB. The AVB profile is i unknown for Column 1 where no data is available due to the configuration of l the eddy current probe manipulator. In all cases the AVB is located on both i sides of center, indicating the AVB penetrates deeper than the tube. This  !

penetration depth exceeds the typical design depth (Row 9) found in most i i Westinghouse steam generators. Also, there is an 0.011 inch gap between the  !

j AVB and tube. This is notably smaller than other steam generator designs.

i This smaller gap further reduces tube deflection.

l t Bulletin Action C.2(a): Stability Ratio Assessment f

I

The bulletin requested that an analysis be performed to assess the  !

] potential for rapidly propagating fatigue cracks. The key contributor to this

, phenomenon is alternating stress produced by flow-induced vibration. Such vibration is caused by fluidelastic instability.

t Westinghouse performed an evaluation to determine if fluidelastic tube  !

vibration instability exists in the Yankee steam generators (Reference (c)). l The results of tt t analysis show that there is a very low potential for fluidelastic instabilit's, therefore, tube failure of the type described in the bulletin is improbable.

l The Westinghouse assessment made use of a simplified one-dimensional analysis technique which compares the calculated fluidelastic stability ratio  !

of selected YNPS tubes to that of the ruptured North Anna tube. It is defined I by the following equation:

SR 2 1/2 1/2 1/2 l olant = (E plant VRA n VRA VRA X y X SR VRA (pV ) VRA (m plant)" f (6 plant) n plant i

Unitcd Statsa Nucient R gulctory Commission Page 3

. FYR 88-119 While this simplified approach cannot account for the three-dimensional tube bundle flow distribution, it does consider the major factors affecting the stability ratio. Four components make up this ratiot a loading term based on the dynamic pressure (pV2 ), a tube incremental mass (m) term, the natural frequency of the tube (fn), and a damping ratio (6) term. The overall ratio is relative, in that each component is expressed as a ratio of the YNPS value to the value for North Anna 1.

The calculation of the relative stability ratio considers the overall steam generator operating conditions (steam flow, steam pressure, and circulation ratio), along with the geometry, to calculate an average secondary-side U-bend void fraction, density, and radial tube gap velocity.

Tne velocity and density are then used in forming a ratio based on dynamic pressure (pV 2 )l/2, which is one component of the relative stability ratio.

Another component is the tube damping which is calculated from an experimentally derived correlation dependent upon the void fraction. The particular correlation which is used for all relative stability ratio calculations is based on a conservative assumption that a dented condition at the top tube support plate exists (a clamped / clamped condition). The clamped condition is also assumed in calculating the tube natural frequency.

Justification for use of a simplified one-dimensional relative stability ratio is provided by comparing the results with those obtained from more detailed three-dimensional bundle flow / tube vibration calculations. These comparisons have been completed for a number of other larger size, feed-ring generator models. For these feed-ring steam generator assessments, the accuracy of the one-dimensional relative stability ratio results compared to the maximum ratios f rom the detailed analyses, three-dimensional was determined to be in the range of 110 percent for selected small radius tubes of interest.

It should be noted that the one-dimensional equation does not include local tube flow peaking factors which can result from local variations in AVB insertion depths. Inherent in applying the one-dimensional equation is the assumption that the flow peaking factor for the plant being evaluated is as high as the local flow peaking factor found for North Anna 1 Tube R9C51. To date, plant evaluations for tube fatigue have only identified a very few AVB positions leading to a peaking factor higher than North Anna R9C51.

The relative stability ratio for the fully supported YNPS Row 8 tube is a very low 0.45. With allowance for uncertainties in the one-dimensional method and the fact that the tubes do not exhibit any physical distortion, the stability ratio of 0.45 indicates that a low potential exists fer fluidelastic ,

instability and tube fatigue.

CONCLL'S IONS YNPS steam generator tubes have performed extremely well over the past j 27 years. We have experienced only 38 actual penetrat!ons over that period of time. All of these have demonstrated a consistent character and rate of leak

. t

' United States Nucicer Reguictory Comission Page 4 FYR 83-119 progression. Each followed the same pattern of gradual increase in leak rate over periods of many months. Their slow rate of progression suggests that they are initially pinhole in size and increase in diameter with time. They do not catastrophically fall or fail over a short period (days).

l Since 1977, Yankee has had a very aggressive policy toward edoy current inspections of steam generator tubes which has been responsible for our low leakage rates. A total of 309 tubes (encompassing cold and hot legs) have been plugged in the four steam generators. This excellent plugging record has resulted from a nearly 100 percent eddy current testing of all four generators. This places YNPS in the clean category (Reference (d)). The tube plugging criteria used at YNPS stipulates that a tube must be plugged when the wall thickness has been reduced by 40 percent. It is worth noting that for these tubes the remaining undegraded wall thickness is nearly equivalent to the total tube wall thickness of later generation facilities.

In addition, thw YNPS steam generator AVB penetrates to Row 8 which exceeds the typical Westinghouse design penetration, Row 9. We believe that thicker wall tubes coupled with adequate lateral support from the AVB results in tube stresses well below that which is required to cause high cycle fatigue therefore, rapidly propagating fatigue failure is not a credible event for YNPh.

If there are any questions regarding this matter, please contact us.

Very truly yours, YA.'KEE ATO. IC ELECTRIC COMPANY J. DeVincentis l

Vice President JDV/25.779 COM90NWEALTH OF MASSACHUSETTS)

)ss MIDDLESEX COUNTY )

Then pert anally appeared before me, J. DeVincentis, who, being duly sworn, did state that he is Vice President of Yankee Atomic Electric Company, that he is duly authorized to execute and file the foregoing document in the name and on the behalf of Yankee Atomic Electric Company and that the statements therein are true to the best of his knowledge and belief.

Robert H. Groce Notary Public My Comission Expires August 29, 1991 cet USNRC Region I USNRC Resident Inspector, YNPS W. T. Russell - Regional Administrator USNRC Region I

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