ML20125C093
| ML20125C093 | |
| Person / Time | |
|---|---|
| Issue date: | 07/11/1979 |
| From: | Zwolinski J Office of Nuclear Reactor Regulation |
| To: | Strosnider J Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-03, REF-GTECI-A-04, REF-GTECI-A-05, REF-GTECI-SG, TASK-A-03, TASK-A-04, TASK-A-05, TASK-A-3, TASK-A-4, TASK-A-5, TASK-OR NUDOCS 8001030475 | |
| Download: ML20125C093 (2) | |
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liEMORANDUM FOR:
J. Strosnider, Task l'anager, A-3, A-4 and A-5, D0R FROM:
J. Zwolinski, Containment Systems Branch, DSS
'BJECT:
CONTAINMENT RESPONSE TO MSLB 'SSUMING STEAM GENERATOR TUBE FAILURES Follening intc1nal staff discussions and receipt of the subsequent memo dated June 29, 1979 on TAP A-3, "Uestinghouse Steam Generator Tube Leakage" in which a request was made to investigate the consequcaces of steam gcnera tor tube failures coincident with a main steam line break, the follo'.;ing r.'a ta hwe bcen generated.
Westinghouse PWR With 0.908 Square Foot MSI.B TNcher of Tubes Ruptured Resulting Peak Assuming 125/ gal / min P (psia)
P(psia)
Flow p r Tube 0 625 sec 0 1800 sec 1
50 29 10 52 32 20
/
54 35 50 58 45 Generally, based on the study performed, one can interpret the results as showing that in all cases the containment pressure response does increase as the nu.ter of tube failures increases.
A prior analysis of a related problem had bcon perfou.ed by the Contain:;ent Systems Branch.
This prior analysis, " Acceptability of I'on-Safety Grade Equipmnt in Mitigating a Main Steam Line Break Accident Inside Containment,"
from F. Eltewila to R. Tedesco, dated January 12, 1977, was found to provide additional insight.
Specifically, it was reported that a main steam line break analysis had been performed assuming that the entire energy inventorv of the primary and secondary systems was available for release to the containm2nt.
The principal purpose of that analysis was to bound the containment pressure response to a main steam line break accident.
It was l
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m J. Strosnider :JL 11,.79 reported that for all the cases considered; i.e., h'estinghouse 3 & 4 loop plants and a typical CE 2 loop plant, the containmant design piessure Uculd be excceded by a factor of less than two.
The raximum allowable contain int pressure, for either a steel vessel or a reinforced concrete structure, based on the ultimate strength of steel is approxinately 2.75 times the contairuent design pressure.
It was, therefore, concluded that the consequences of such a hypothetical accident would not include the loss of containoant structural integrity.
h'e conclude, therefore, that steam generator tube ruptures coincid:nt with a MSLB acciderit would not lead to a loss of contair.... ant integrity.
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Jol.nq,,olinski Contancent Systcms Branch Division of Systcms Safety cc:
S. lbnwer R. Denise D. Eiscnhut M. Aycock W. Butler J. 'ri c k V.
','o; na n Z. Resztoczy P. Ch ck D. Pickett
- t. Sorrer 90012244 K. Parczewski F. Odar F. Al p ter E. Adensam