ML19350A965
| ML19350A965 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, Oconee, Arkansas Nuclear, Crystal River, Rancho Seco, Crane  |
| Issue date: | 03/04/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19350A902 | List: |
| References | |
| NUDOCS 8103170671 | |
| Download: ML19350A965 (2) | |
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' " UNITED STATES NUCLEAR REGULATORY COMMISSION
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SAFETY EVALUATIDH BY THE OFFICE OF NUCLEAR P.EACTOR REGULATION CONCERNING ITEM 11.K.2.19 "BEHCHPARK ANALYSIS OF SEOUENTIAL AUXILIARY FEEDRATER FLOW" FOR BABC0CK & WILCOX REACTOR PLANTS D3CKETS NOS. 50-269, 50-270, 50-287, 50-289, 50-302,50-312, 50-313. AND 50-346 Introduction At a meetino in Bethesda, April 25, 1979, with the ceners of Babcock and Wilcox (S&W) reactor plants, we requested a benchnark analysis of sequen.
tial auxiliary feedwater flow to the steam generators following a loss of main feedaater.
This analysis was provided in a let er from J. Taylor (BLU) to R. l'attson (HRC) deted June 15, 1979. However, in this analysis the TPAP-2 Code with 6 node steam generator model was utilized.
All small break analysis presented to the NRC have been perfomed using the CRAFT-2 Code with a 3 node steam generator model. We require a benchmark analysis for sequential auxiliary feedwater flow also be performed using CRAFT-2 with a 3 mode steam generator representation. By le-ter dated August 21; 1979 we requested such analysis.
Each licensee of ELW reactor plants
, responded with a report which presented analysis of sequential auxiliary feedwater flow to the steam generators for a loss of main feedwater trans-
. ients using the CRAFT-2 Code.
This issue was later identified as Item II.K.2.19 of the TMI Action Plan requirements.
Discussion & Conclusions EEN utilizes the CRAFT-2 computer progran in perforcing loss of coolant eccident (LOCA) licensing evaluations for their nuclear steam supply systems (NSSS).
i Subsequent to the Tlil-2 accident, this computer progra n was used to confirm
- " 'cmergency operator guidelines for all power plants with NSS5s designed by
$&W. Our review of these confirmatory analyses have led to questions re-garding the ability of the CRAFT-2 program to adequately predict steam generator perforrance anc its influence on the prim.ry system thermal-hydraulic behavior.
In particular, we noted that the CRAFT-2 steam generator nodel did not contain the same degree of cetail as the model used with the TPAP-2 Code.
TPAP-2 is a computer code prirarily used for non-LOCA transients by D&9.
In order to validate the TF.'F-2 transient code
> ith actual plant data, an asyrmetric cooldown test was incorporated into the Crystal River Unit 3 po.ter ascension prograr..
Escause cf the simplified steam generator model in the CRAFT-2 Code, we also re:;uested that the CRAFT-2
- n be assessed against the Crystal River Unit 3 asy.:.e ric coolde.en data.
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l P00R ORIGINAli The comparative analyses of the startup test de:nonstrated that the simplified steam generator model used in the licensing code (CPAFT-2) predicted thermal-hydraulic behavior similar to the more detailed steam generator model utilized in the TPAP-2 Code.
However, ro:parisons with data for both codes i
were poor.
Further exa=inaticn of the. Crystal River Unit 3 asymnetric startup test has indicated the test to be inappropriate for assessing com-
- uter codes.
This is attributed to inadequate instruz.entation whereby key data required for code assessment were not obtained.
Revitas conducted by our E&O Task Force, folloaing the TMI-2 accident, have concluded that further assessment of the CPAFT-2 Code would be required.
The r.ajority of the concerns identified are documented in NUREG-0565.
In particular, the neglect of a ne:hanistic, regine-dependent heat transfer model and the use of a constant, steam generator heat transfer coefficient throughout the transier.: have been identified as requiring either revision or ft,rther justificatien.
This requirement for further justification and/or revision of -he stall treak ECC5 nctels is being perforced under Till Action Plan Ite. ?!.K.3.33.
We believe that satisfactory resolution of code modeling cent. erns as par cf the Ac-ion Iten )).K.3.30 will resolve the tedelinr; cer.: erns cf II.K.2.19.
The conIlusions of our revie. cf Action Item II.r. 2.19 are as follows:
(a) The intent of Ite: II.K.2.19 was acconplished, (b) Results provided by CTAFT-2 wer2 similar to these provided by the more detailed T?AP-2 program.
However, both codes showed poor agreement when cc: pared with the test data.
-(c) The poor agreeme.: of the code prediction with test data has been attributed to the fact that the Crystal River ascension test data was
.not adequate for essessing thernal-hydraulic codes, and (d) A more rigorous assessment of the SLW sr.all break LOCA nodel is being
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perforced under TEI Action Item 11.K.3.30.
Further code assessment i
under TM1 Action I em II.K.2.19 is therefere unnecessary.
i Sased on the above conclusions, we consider Item 11.K.2.19 completed by all licensees with ElW NSS5s by issuance of this Safety Evaluation Report.
l Itrecycr. we do not believe it necessary for Item ]I.K.2.19 to be addressed any further.
bated: March 4, 1981 E l e
e