ML20147D638
| ML20147D638 | |
| Person / Time | |
|---|---|
| Site: | Yellow Creek  |
| Issue date: | 09/29/1978 |
| From: | Ketchen E NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | Kornblith L, Paris O, Smith I Atomic Safety and Licensing Board Panel |
| References | |
| NUDOCS 7810140152 | |
| Download: ML20147D638 (6) | |
Text
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4 UNITED STATES G
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NUCLEAR REGULATORY COMMISslON g
j WASHINGTON, D. C. 20556 k*****
September 29, 1978 NRC PUBLIC DOCUMENT ROOM Ivan W. Smith, Esq., Chairman Dr. Oscar H. Paris Atomic Safety and Licensing Board Atomic Safety and oard
~ U.S. Nuclear Regulatory Commission U.S. Nuclear Re ion Washington, D. C.
20555 Washington, D.
0555 f.h, f Mr. Lester Kornblith Atomic Safety and Licensing Board
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U.S. Nuclear Regulatory Comission Washington, D. C.
20555' fas#/
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6 In the Matter of TENNESSEE VALLEY AUTHORITY co (Yellow Creek Nuclear Plant, Units 1 & 2)
Docket Nos. 50-566 & 50-567 Gentlemen:
Recently during the course of certain'semiscale experiments, conducted at Idaho Nuclear Engineering Laboratory to model various aspects of ECCS performance, one test run exhibited behavior unanticipated by current' ECCS performance models.
Semiscale experiment Mod-3, S-A7-6 was run on September 12, 1978.
It was intended to model' an integral blowdown-refill-reflood scenario for a double-ended cold-leg break. Some of the detailed results were un-anticipated.
For example, the heated core simulator was predicted by the RELAP code to quench at 110 seconds.
Instead, it dried out again and went through several cycles of dryout and rewet (see enclosed Figure 1). Other portions of the cladding temperature profile also y
showed discrepancies in that test temperatures in some instances were somewhat above those predicted and in some instances were somewhat below those predicted (see Figure 2 and 3).
During_ the test, the downcomer voided several times in the 100 to 400 seconds period of time. This
/;
also was not predicted by RELAP (Figure 4 shows one such void).
During the periods of downcomer voiding there was also negative (downward) flow from the heater to the lower plenum.
As indicated above, nearly complete downcomer voiding occurred after downcomer fill. This is not predicted by the ECCS evaluation models (for PWR's) used in connection with 10 CFR 550.46 and Appendix K applications. Also, typical Appendix K calculations do not show successive i
dryout and rewets over the extended reflood cycle. A quick-look report on this experiment will be published by INEL about October 1,1978.
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2 The present judgment of INEL is that these unanticipated results are atypical and have been produced by certain characteristics of the experiment which are not typical of reactor systems, particularly the stored heat in the downcomer pipe and'in the one-dimensional arrangement of the downcomer.
These matters are under further study by the NRC Staff and INEL. When further information or conclusions concerning this matter become available, we will inform the Boards.
Sincerely,
)
Edward G. Ketchen Counsel for NRC Staff Enclosure as Stated cc (w/ encl.):
I ra L. Mye rs, M. D.
Atomic Safety and Licensing Board Panel Atomic Safety and Licensing Appeal Panel Docketing and Service Section Herbert S. Sanger, Jr., Esq.
Honorable A. F. Summer Alton B. Cobb, M.D.
William B. Hubbard, Esq.
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Figure 1 COMPARISON OF ROD CLADDING TEMPERATURES AT CORE HIGH POWER ZONE WITH RELAP4 FOR TEST S-A7-6 4'
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COMPARISON OF MEASURED AND CALCULATED CLADDING TEMPERAUTRE IN LOWER PART OF CORE FOR TEST S-A7-6 If I
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COMPARISON OF MEASURED AND CALCULATED CLADDING TEMPERATURE IN UPPER PART OF CORE FOR TEST S-A7-6 1050 gr 1000 TH-C2-277 950 n
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.. i CALCULATED COLLAPSED DOWNCOMER LIQUID LEVEL FOR TEST S-A7-6 I,
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