ML19347D989
| ML19347D989 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 04/07/1981 |
| From: | Mills L TENNESSEE VALLEY AUTHORITY |
| To: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| 10CFR-050.55E, 10CFR-50.55E, NCR-SWP-8023, NUDOCS 8104140582 | |
| Download: ML19347D989 (2) | |
Text
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4-TENNE 39EE VALLEY AUT:--
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April 7, 1981 h
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,,qh Mr. J P. O'Reilly, Director Office of pection and Enforcement P
1 U.S. Nuclear egulatory Cournission Region II - Su e 3100 Me e
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Atlanta, Georgia 303
Dear Mr. O'Reilly:
SEQUOIAH NUCLEAR PLANT UNIT 2 - SEISMIC ANALYSIS FOR AS-BUILT SAFETY-RELATED PIPING SYSTEMS - NCR SWP 8023 - FINAL REPORT The subject deficiency was initially r1rported to NRC-0IE Inspector R. W. Wright on November 28, 1980, in accordance with 10 CFR 50.55(e).
Interim reports were submitted on December 29, 1980, and March 2, 1981.
Enclosed is our final report.
If you have any questions, please get in touch with D. L. Lambert at FTS 857-2581.
Very truly yours, TENNESSEE VALLEY AUTHORITY L. M. Mills, Manager Nuclear Regulation and Safety Enclosure cc:
Mr. Victor Stello, Director (Enclosure)
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Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, DC 20555
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ENCLOSURE SEQUOYAH NUCLEAR PLANT UNIT 2 3
SEISMIC ANALYSIS FOR AS-BUILT SAFETY-RELATED PIPING SYSTEMS NCR SWP 8023 10 CFR 50.55(e)
FINAL REPORT Description of Deficiency This NCR was generated as a result of IE Bulletin 79-14.
The as-built
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configurations of various safety-related piping systems have revealed a number.of nonconformances to design drawings. These discrepancies between as-built and as-designed pf ping and hangers could affect the validity of the seismic analyses. Design specifications and drawings are used to obtain input information for seismic analysis of safety-related systems. As a result, various safety-related systems may not be seismically qualified.
Safety Implications Failure of the piping during a seismic event could render various safety-related systems inoperable which could adversely affect the safety of the plant. For instance, less coolant could be delivered te the reactor core during an LOCA than is claimed in the plant safety analysis.
Corrective Action 4
TVA has determined that approximately 12-1/2 percent of the supports and 40 percent of the piping inspected will require craft rework. The piping rework for the most part is to provide proper clearance for pipe movement. Of the piping packages inspected, approximately 7 percent required reanalysis to determine if the seismic analysis is still valid. The complete field walkdown, reanalysis, and rework of piping and supports will be completed by fuel loading. Any rework not complete by fuel loading will be justified on a case-by-case basis.
Final assurance that the requirements of IE Bulletin 79-14 have been satisfied will be demonstrated by a detailed inspection and evaluation of a representative sampling of rigorously analyzed, sa.fety-related piping. This program is ongoing and will also be completed before fuel loading.
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