Overview on critical heat flux experiment for the reactor fuel assemblies

Abstract

The critical heat flux (CHF) experiment for the reactor fuel assemblies (FAs) is the most difficult and important one in the reactor thermo-hydraulic study. The FA CHF experiments are performed on the large scale thermo-hydraulic test facilities with large DC power suppliers, complicated test sections and rigid test process. The results are analyzed with subchannel codes to obtain the parameters at the CHF point. Based on the experimental data, the CHF correlation is derived as the tool for the reactor safety analysis and design. After many years’ work, the common understanding of the FA CHF was established and the method of the FA CHF experiment becomes similar. This paper reviewed the progress of the FA CHF experiments, analyzed the modeling of the test sections, introduced requirement of the test facilities, and discussed the hypothesis in the modelling. The benchmark qualification test adopted to verify the test method and test section used in the experiment are discussed in the paper, too.