Improvements to the modelling of in a transient nuclear two-phase flow and heat transfer reactor analysis code

Accurate two-phase flow heat transfer prediction is of great importance in the analysis of reactor safety and TRAC (transient reactor analysis code) is a best estimate (BE) system code developed for such analysis. The work described here forms part of a research project that aims to evaluate and improve the TRAC code behavior by comparing code predictions with a range of "single effect" experiments. It has been shown that the necessary friction factor can be obtained from a voidage correlation and the drift flux parameters and the wall friction coefficient can be further derived from the annular flow model of Owen and Hewitt. These modifications have been combined to give what is regarded as an optimum code for vertical pipe flows. The work has been extended to make further improvements to the heat transfer correlations of the code. The performance of both the original and the improved codes has been tested against experiments on both evaporating and condensing flows. It is found that the improved code, combining the strongest parts of the best available correlations, gives better predictions in every case. (c) 2007 Elsevier Ltd. All rights reserved.