ML20027D543
| ML20027D543 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 10/19/1982 |
| From: | Mattson R Office of Nuclear Reactor Regulation |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20027D544 | List: |
| References | |
| NUDOCS 8211060583 | |
| Download: ML20027D543 (3) | |
Text
/
e UNITED STATES o,dg
! %.. # o NUCLEAR REGULATORY COMMISSION I
WASHINGT ON, D. C. 20$55
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i MEMORANDUM FOR:
Darrell Eisenhut, Director, Division of Licensing FROM:
Roger J. Mattson, Director, Division of Systems Integration
SUBJECT:
BOARD NOTIFICATION CONCERNING RECENT SEMISCALE TEST RESULTS In my memorandum dated August 30, 1982, I reported the results of a recent Semiscale " feed and ble ed" test (S-SR-2).
In that memorandum, I stated that there was insufficient information available at that time to draw any con-clusions from the results and we would pursue resolution of this issue and inform the Boards of our conclusions following completion of our evaluation.
The purpose of this memorandum is to request that you inform the Boards that we have completed our evaluation and have concluded that Semiscale test S-SR-2 does not exhibit any new phenomena and can be adequately predicted by our computer codes.
Semiscale test S-SR-2 simulated a loss of all feedwater which resulted in a-complete dryout of the secondary side of the steam generators.
This resulted in a pressurization of the primary system.
The scaled PORY was opened to de-pressurize the primary system to below the HPI pump shutoff head. This action, referred to as primary feed and bleed, was done in an attempt to reach an equilibrium thermal-hydraulic condition for core cooling. This involves re-lieving the primary pressure increase due to core decay heat through the PORY and replacing the primary coolant inventory lost out of the PORV with HPI coolant. The high head charging pump was assumed to be inoperable for this test.
Prior to achieving this equilibrium thermal hydraulic condition, the core simulator rods began to heat up excessively.
Our evaluation and conclusions are based on RELAP-5 analyses conducted by EG&G, Idaho, for the Office of Huclear Regulatory Research of both the Semi-scale S-SR-2 test and a corresponding feed and bleed mode of operation for a' typical Westinghouse 4-loop plant (RESAR plant). These analyses have shown the following:
- 1) A RELAP-5 analysis of the Semiscale test SR-2 demonstrates the code's ability to accurately calculate both the overall system response and local responses.
The RELAP-5 results show good quantitative agree-
. ment with the test data.
- 2) RELAP-5 feed and bleed analyses for a Westinghouse RESAR plant de-sign both with and without full ECCS (charging and HPI) agree with the general behavior seen in the Semiscale experiments.
(The RESAR calculation with ECCS results in steady state core cooling being achieved and an eventual restoration of subcooling in the primary system.}*
- The RESAR calculation assumed that ECCS charging flow was available and a decay heat curve was used while Semiscale used only HPI pumps and a constant power level.
7 CONTACT:
M. Keane, 4gfd3 x28957 6,2 / / U
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Darrell Eisenhut -
00T 18198Zt Additionally, the following conclusions are reached about the Semiscale test results.
After voiding of the hot leg, the ability to maintain steady state core cooling is considered within the experimental uncertainties of the values of core decay heat, PORV characteristics and HPI pumping capa-bility. A small reduction in core power or PORV. mass flowrate or an in-crease in HPI flowrate would probably have resulted in steady-state core cooling before core uncovery.
The RELAP-5 calculation of the test results in either steady-state cooling before core uncovery or just as core un-covery begins depending on how uncertainties in steam generator heat trans-fer are treated in the code input.
The test analysis confirms our previous conclusion that the viability of feed and bleed as well as the system response is a function of many para-meters, including core decay heat level, PORV characteristics, ECCS pump-ing capacity and operator action time.
In summary, our review of the test data and the RELAP-5 analyses performed.to predict this test data lead us to conclude that no new phenomena are exhibited by Semiscale test S-SR-2 and that the staff's RELAP-5 analysis code, adequately predicts the test data and associated thermal hydraulic phenomena.
We point out that regardless of the conclusions that may have been reached from this test regarding viability of feed and bleed, feed and bleed cooling is not a design basis reouirement considered necessary to meet the Commis-sion's Regulations for any LWRs currently licensed or being considered for a license. We are providing this follow-up evaluation due to the interest in feed and bleed cooling expressed in recent licensing proceedings and be-cause we believe it is in 'the best interest of the regulatory process to keep the licensing boards informed of these recent test results.
These analyses and this test are not a generic indication of the ability of PWRs to feed and bleed.
The detailed capability of a PWR to feed and bleed must be determined by individual analyses, e
bi
,h Roger J. Mattson, Director Division of Systems Integration ec: H.'Denton E. Case S. Hanauer
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- 0. Bassett, RES M. Cutchin, OELD
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BOARD NOTIFICATION 82-107 DATED Semiscale Test Results To Commission Summer
' Floating Nuclear. Plants
'Diablo Canyon To ASLB To ASLAB Callaway Diablo Canyon Comanche Peak Summer Palo Verde Midl Floating Nuclear Plants
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. Midland Waterford
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DISTRIBUTIDN
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NRC PDR L PDR PRC System NSIC DL:L Reading E. Christenbury,0 ELD T. Novak/L. Berry J. Sci nto, OELD S. Black A. Bennett, OELD.
J. Youngblood M. Rushbrook E. Edison S. Burwell W. Kane
- 10. l'J1iChtCP.
J. Lee E. Licitra B. Buclkey E. Adensam M. Duncan D. Hood R. Hernan C. Stahle D. Eisenhut/R. Purple M. Williams H. Denton/E. Case PPAS
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R. Vollmer R. Mattson S. Hanauer H. Thompson Attorney, OELD (for appropr. ate plant)
E. L. Jordan, I&E J. M. Taylor, I&E W. J. Dircks, EDO (3)
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