ML20008F124
| ML20008F124 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, Oconee, Arkansas Nuclear, Crystal River, Rancho Seco, Crane  |
| Issue date: | 02/27/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20008F115 | List: |
| References | |
| NUDOCS 8103120308 | |
| Download: ML20008F124 (2) | |
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4 UNITED STATES i
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NUCLEAR REGULATORY COMMISSION U - A '.*
t WASHINGTON, D. C. 205S5 -
h SAFETY EVALUATION BY THE OFFICE OF NUCLEAR F.EACTOR REGULATION CONCERNING ITEM II.K.2.19 "BENCHPARK ANALYSIS OF SEOUENTI AL AUXILIARY FEEDWATER FL OW" 4
FOR j
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 meetine in Bethesda, April 25, 1979, with the craners of Babcock and 4
Wilcox (B&W) reactor plants, we requested a benchnark analysis of sequen-tial auxiliary feedwater flow to the steam generators following a less of nain feedwater. This analysis was provided in a letter from J. Taylor (B&W) to R. Mattson (NRC) dated 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 performed using the CRAFT-2 Code with a 3 node steam generator model. We require a benchmark analysis for sequential auxiliary feedwater flow also be perforced using CRAFT-2 with a 3 mode steam generator representation. By letter 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-i ients using the CRAFT-2 Code.
This issue was later identified as Item II.K.2.19 of the TMI Action Plan req ui rements.
Discussion & Conclusions l
S&W utilizes the CRAFT-2 computer progran in performing loss of coolant -
l accident (LOCA) licensing evaluations for their nuclear steam supply systems (NSSS).
Subsequent to the illI-2 accident, this conputer program was used to confirm
- emergency operator guidelines -for all power plants with USSSs-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 performance anc its influence on the primsry system thermal-
- hydraulic behavior.
In particular, we noted that the CFAFT-2 steam I
l generator nodel did not contain the same degree of cetail as the model.uted v;ith the TPAP-2 Code.
TPAF-2 is a computer code primarily used for non-LOCA transients by D&W. 'In order to validate ~ the.TFAF-2 transient code seith actual. plant data, an asycretric cooldown test sas incorporated into the Crystal River Unit 3 poder ascer.sion progran.- Escause cf the simplified
' steam centrator model in the CRAFT-2 Code, we also re:;uested that the CRAFT-2 ti:-be assessed against the Crystai River Unit 3 asy-etric cooldown date..
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The comparative analyses of the startup test demonstrated that the simplified steam generator model used in the licensing code (CRAFT-2) predicted thermal-hydraulic behavior similar to the more detailed steam generator model utilized in the TPAP-2 Code.
However, comparisons with data for both codes were poor.
Further examination of the Crystal River Unit 3 asymmetric startup test has indicated the test to be inappropriate for assessing com-puter codes. This is attributed to inadequate instrumentation whereby key data required for code assessment were not obtained.
Revieas conducted by our B&O Task Force, following the TMI-2 accident, have concluded that further assessment of the CPAFT-2 Code would be required.
The najority of the concerns icentified are documented in NUREG-0565.
In particular, the neglect of a mechanistic, regime-dependent heat transfer model and the use of a constant, steam generator heat transfer coefficient throughout the transient have been identified as requiring either revision or further justification.
This requirement for further justification and/or revision of the small break ECCS nedels is being performed under Till Action Plan Iten II.K.3.30.
We believe that satisfactory resolution of code modeling ccncerns as part of the Action Iten II.K.3.30 will resolve the modeling concerns of II.K.2.19.
The conclusions of our review of Action Item II.K.2.19 are as follows:
(a) The intent of Item II.K.2.19 was accomplished, (b) Results provided by CRAFT-2 were similar to those provided by the more detailed TRAP-2 program.
However, both codes showed poor agreement when compared with the test data, (c) The poor agreement of the code prediction with test data has been attributed to the fact that the Crystal River ascension test data was
- not adequate for assessing thermal-hydraulic codes, and (d) A more rigorous assessment of the B&W small break LOCA model is being performed under TMI Action item II.K.3.30.
Further code assessment under Bil Action Item II.K.2.19 is therefore unnecessary.
Based on the above conclus3cns, we consider Item II.K.2.19 completed by all licensees with B&W NSSSs by issuance of this Safety Evaluation Report.
Moreover, we do not believe it necessary for Item II.K.2.19 to be addressed any further.
Dated:
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