LWR core thermal-hydraulic analysis : assessment and comparison of the range of applicability of the codes COBRA IIIC/MIT and COBRA IV-I
Author(s)Kelly, J. E.; Loomis, James N.; Wolf, Lothar
This report summarizes the result of studies concerning the range of applicability of two subchannel codes for a variety of thermal-hydraulic analyses. The subchannel codes used include COBRA IIIC/MIT and the newly developed code, COBRA IV-I which is considered the benchmark code for the purpose of this report. Hence, through the comparisons of the two codes, the applicability of COBRA IIIC/MIT is assessed with respect to COBRA IV-I. A variety of LWR thermal-hydraulic analyses are examined. Results of both codes for steady-state and transient analyses are compared. The types of analysis include BWR bundle-wide analysis, a simulated rod ejection and loss of flow transients for a PWR. The system parameters were changed drastically to reach extreme coolant conditions, thereby establishing upper limits. In addition to these cases, both codes are compared to experimental data including measured coolant exit temperatures in a core, interbundle mixing for inlet flow upset cases and two-subchannel flow blockage measurements. The comparisons showed that, overall, COBRA IIIC/MIT predicts most thermal-hydraulic parameters quite satisfactorily. However, the clad temperature predictions differ from those calculated by COBRA IV-I and appear to be in error. These incorrect predictions are caused by the discontinuity in the heat transfer coefficient at the start of boiling. Hence, if the heat transfer package is corrected, then COBRA IIIC/MIT should be just as applicable as the implicit option of COBRA IV-I.
Based on the M.S. thesis of the first author in the M.I.T. Dept. of Nuclear Engineering, 1978.
MIT Energy Laboratory
Nuclear fuel elements |x Computer programs., Boiling water reactors.
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