ML19221A031
| ML19221A031 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/31/1977 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19221A025 | List: |
| References | |
| SER-771231, NUDOCS 7905160508 | |
| Download: ML19221A031 (5) | |
Text
e SAFETY EVALUATION OF COM3USTION ENGINEERING OTSG TUEE SLEEVE QUALIFICATION PR0',RMI THR"E MILE ISLAND UNIT 2 December, 1977 n-7 1,20 al
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SAFETY EVALUATION 1.
INTRODUCTION 1.1.
Project Genesis.
Excessive OTSC tube vibration has been postulated as a contributing factor to several OTSG tube leaks at the OCONEE nuclear power station
'and to wear damage to OTSC tubes at Three Mile Island Unit 1.
In an effort to assess t'ie potential for future damage and to prepare for possible remedial action, General Public Utilities Service Corporation has elected to pursue two separate, but parallel, courses of action.
One OTSG at TMI-2 is to be instrumented with acceleremeters and strain gauges to determine tube v;bration characteristics.
In parallel with this instrumentation program, General Public Utilities Service Corpora-tion is developing a tube stif fening device.
In June, 1977, CPUSC contracted with Combustion Engineering to perform single tube vibration testing of sleeved and unsleeved steam generator tubes. The purpose of this testing was to determine if sleeves would be ef fective in reducing dynamic strain.
The results of the testing are contained in Attachment 1 to this Safety Evaluation. The conclusion from the testing was tha t a minimum stress reduction of 30% at the upper tube cheet is expected to result from a similar sleeving operation of a once-through steam generator.
In additi6n, a comparison of mode shape pa terns, shown on Figure 12 of Attachment 1, indicates that dynamic strain at other support plates is not increased to a level greater than that tha t occurs at the upper tube sheet.
The conclusion is that the installation of sleeves in a steam generator tube at the upper tube
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e sheet and 15th support plate l o ca t io n s does not cause an unacceptable increase in strain at other aupport elevations.
As a result of these encouraging vibration test results, GPUSC con-tracted with Combustion Engineering in August, 1977, to develop the techniques, tools, and materials for installing demonstration repair sleeves in a Three Mile Island Nuclear Power Plant Unit 2 steam generator.
1.2.
Purpose The purpose of the installation of demonstration repair sleeves in a TMI Unit 2 steam generator is to allow verification of the perfor-manc2 of the sleeves under operating conditions. The performance of the sleeves will be monitored using eddy current methods during refueling outages. These subsequent eddy current examinations will involve the use of an axially wound eddy current test probe provided by Combus tion Engineering.
In addition to eddy current examina-tions, it is planned that one tube containing two sleeves will be removed at the end of the first fuel cycle for a destructive metal-lurgical examination.
2.
SAFETY EVALUATION A modification of TMI Unit 2 OTSG "B" is required to accomodate the installation of demonstration repair sleeves as described in Attach-ment 2.
In this regard, the structural modifications have been reviewed to assure that the s tructural integrity of the OTSG, in its final configuration, will be as good as, or better than, that pre-vious ly described in the FSAR.
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The conse'quences o r an accident can be regarded to e increased by (1) failure of an additional structure due to loads induced by the initial accident ccedition and/or (2) failure of cafety related equipment to perform as expected.
With regard to item (1), above, the sleeves do not constitute a primary or secondary side pressure retaining boundary.
In addi* ion, the.004 inch to.006 inch diametral expansion of the parent tube results in a hoop strain that does not exceed 10.
A hoop strain of less than 1% is considered insignificant and does not degrade the capability of the parent tube to act as a piimary pressure retaining boundary.
Alto, the expansion from sleeve installation is less than the roller expansion that occurs at the tube sheets during steam generator construction.
This information pro-vides sufficient assurance that additional structures will not fail and that the consequences of such an ac_ident will not be increased by this mechanism. With regard to item (2), there will not ce a piece of hard-ware directly related to safety which will be = edified.
documents that an axial load in excess of 3,000 pounds would be required to separate the sleeve from the parent tube, Since the flow loading on the sleeve is less than 20 pounds, there is sufficient margin to assure that the sleeve will not be ejected as a loose part from the parent tube.
Neither the structural integrity, nor the operational charac-teristics of any " safety related equipment" is af fected.
Consideraticn of the preceding information demonstrates that tne structural integrity of the CTSG has not been degraded, and the probability of an accident, previously evaluated, is not increased.
In addition, the probability of an accident occurring, which has not been previously evaluated or described in the FSAR, is not increased.
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ATT.\\cus y 57g L.
Test report, single tube vibration testing and evaluation associated with the Three "ile Island Nuclear Power Plant steam generator.
2.
Procedure for installation of sleeves at Three Mile Island Unit 2 steam generators.
3.
Full test results for sleeve tubes.
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