ML20147D638

During Semiscale Experiments Conducted to Model Aspects of ECCS Performance,One Test Run Showed Behavior Unanticipated by ECCS Models.Believed That Results Were Atypical.Matter Is Under Further Study.Bds Will Be Informed of Addl Info

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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. UNITED STATES g

~\j NUCLEAR REGULATORY COMMISslON WASHINGTON, D. C. 20556 September 29, 1978 k***** 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 Mr. Lester Kornblith f.h, f Atomic Safety and Licensing Board U.S. Nuclear Regulatory Comission ( q, g-r$

Washington, D. C. 20555' "

fas#/

&/ 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 1 (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.

3 % %4-h n5't w-,e-- ,

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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400

-100 0 100 -200 300 400 500 . -

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Figure 2 '

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COMPARISON OF MEASURED AND CALCULATED CLADDING TEMPERAUTRE IN LOWER PART OF CORE FOR TEST S-A7-6 I

If I l l l l l l l 1000 - '

n Y

u -

, 900 -

L 3 '

r 800 - "

A < TH-C3-115 -

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1 500 -

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-50 0 50 100 150 200 250 300 350 400 450 500 .

Twn. Af ter Rup t ure C=3 l

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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 -

6 900 -

!= 850 -

j 800 -

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e 550 -

500 -

450 -

I I I 400 I I I I I I

-50 0 50 100 150 200 250 300 350 400 450 500 .

T.m. Avi. no so . c.2

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, .. i CALCULATED COLLAPSED DOWNCOMER LIQUID LEVEL FOR TEST S-A7-6 ,, .,

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'a li lf ' RELAP4 CALCULATION 5 -

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DATA

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0 l I I I I I I I I I I I I I 50 60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 '

Tim. A e s.r nu,*ue. c.s

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