ML20086H835
| ML20086H835 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 07/12/1995 |
| From: | Shell R TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9507180293 | |
| Download: ML20086H835 (4) | |
Text
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W IUA Tennessee Vasey Aufncoff Post Othce Bau 2TO Soddy Da,sy Tennessee 3/379 July 12,1995 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:
In the Matter of
)
Docket Nos. 50-327 Tennessee Valley Authority
)
50-328 SEQUOYAH NUCLEAR PLANT (SON)- 10 CFR 50.46 ANNUAL REPORT
References:
- 1. TVA letter to NRC dated July 6,1994
- 2. TVA letter to NRC dated November 18,1994, "Sequoyah Nuclear.
Plant (SON)- 10 CFR 50.46 Updated Annual Report" 10 CFR 50.46 requires reporting, at least on an annual basis, each change to or error discovered in an acceptable loss-of-coc! ant accident (LOCA) evaluation model or in the l
application of such a model that affects tae peak clad temperature (PCT) calculation.
The purpose of this letter is to provide the annual report.
In 1993, Westinghouse Electric Corporation performed a structural analysis of the steam generator (S/G) tubes to confirm tube integrity under combined seismic and l
LOCA loading conditions. Westinghouse conservatively established a five percent S/G flow area reduction due to the combined seismic /LOCA loads for SON. This was reported to NRC in the 1993 SON Annual Report as a 20 degree Fahrenheit (F)large break LOCA penality. Westinghouse has now calculated plant specific seismic and LOCA loadings for the SON S/G tube support plates. Based on the new calculations, the penality is only two degrees F. This is reflected in the onclosed report.
Reference 1 supplied SON's 199410 CFR 50.46 Annual Report. Reference 2 provided an updated annual report as a result of a significant change identified in the PCT analytical results. The 1995 report is being provided at this time to remain on the current SON 10 CFR 50.46 report update frequency.
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U.S. Nuclear Regulatory Commission Page 2 July 12,1995 Additionally, potentialissues are under investigation by Westinghouse that may impact the PCT for both large and small break LOCA. The potentialissues have had PCT margin temporarily allocated to ensure that the cumulative efforts are tracked such that the 10 CFR 50.46 PCT limit of 2200 degrees F is not exceeded. Upon their resolution, these issues will continue to be reported as appropriate.
Please direct questions concerning this issue to W. C. Ludwig at (615) 843-7460.
Sincerely,
/
R. H. Shell Manager SON Site Licensing Enclosure cc (Enclosure):
Mr. D. E. LaBarge, Project Manager Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 NRC Resident inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy-Daisy, Tcnnesseo 37370-3624 i
Regional Administrator U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 J
Atlanta, Georgia 30323-2711 l
6 ENCLOSURE 10 CFR 50.46 REPORT DOCUMENTATION l
Larae Break Loss-of-Coolant Accident (LOCA)
PCT Attachment i
Previous Licensing Basis Peak 2174 F Cladding Temperature (PCT)
(July 6,1994)
- Steam Generator Tube Seismic /LOCA
-18 F 1
Area Reduction Plant Specific Analysis N
Updated Licensing Basis PCT 21560F Net Change 18oF Small Break LOCA PCT Previous Licensing Basis PCT 1716oF (November 18,1994)
No Change Identified 0F Updated 'pensing Basis PCT 1716oF Net Change 0F A detailed discussion of the change outlined above is included in the indicated attachment.
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D STEAM GENERATOR FLOW AREA REDUCTION - SEISMIC /LOCA ASSUMPTIONS
Background
1 In the 199310CFR50.46 Annual Report for Sequoyah, a modification to the large break LOCA evaluation model was identified which addressed a potentialloss of core cooling resulting from steam generator tube deformation. In 1993, Westinghouse performed a structural analysis of steam generator tubes to confirm tube integrity under combined seismic and LOCA loading conditions. (The loads were combined to demonstrate structural integrity margins in accordance with Regulatory Guide 1.121.) The analysis found that while tube integrity was maintained under the combined loading conditions, some tube deformation could occur.
Westinghouse conservatively established a 5% steam generator flow area reduction due to the combined seismic /LOCA loads for Sequoyah. This was a bounding value established for all Model 51 steam generators.
a Steam generator tube flow area reduction under combined seismic /LOCA loading is a function of three principal parameters. These include the tube support plate LOCA loads, tube support plate seismic loads and tube support plate load / deflection characteristics. When the original evaluation was performed, actual Model 51 steam generator values were not available for these parameters. Conservative extrapolations of values applicable to the Westinghouse Model D steam generator were used in the evaluation.
Since the original evaluation was performed, Westinghouse has calculated plant specific seismic and LOCA loadings for the Sequoyah steam generator tube support plates. These loadings were based upon Sequoyah plant specific seismic response spectra and the limiting Sequoyah primary system pipe break (as " leak before-break" methodology has been incorporated into the Sequoyah design basis for primary loop piping, small break LOCA pipe loads are limiting for Sequoyah). Additionally, a plate crush test specific to the Model 51 steam generator configuration was performed to establish support plate deflection characteristics. Using the results of these calculations and tests, Westinghouse concluded that the actual steam generator tube flow area reduction for Sequoyah is 0.36% instead of the generic 5% value previously assumed.
Estimated Effect Steam generator tube deformation increases the resistance to steam flow from the core during a LOCA. Based upon sensitivity studies performed by Westinghouse, calculated peak clad temperature will increase 4 F for every 1% reduction in steam generator tube flow area. A 5%
decrease in the flow area results in a 20 F increase in calculated peak clad temperature. A 0.36%
reduction conservatively results in a 2*F peak clad temperature increase. As such, the net result of the Sequoyah plant specific flow area reduction is an 18 F peak clad temperature reduction.
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