Experimental study of critical heat flux in a narrow vertical rectangular channel. Comparison between uniform heat flux case and nonuniform heat flux case.

Abstract

Quantitative understanding of critical heat flux (CHF) in a narrow vertical rectangular channel is required for the thermohydraulic design and safety analysis of research nuclear reactors in which flat-plate-type fuel is employed. Experiments were carried out on CHF for both upward flow and downward flow, with nonuniform heat flux simulating a subchannel in the fuel element of the research nuclear reactor JRR-3. Investigation of experimental data showed that the CHF scheme proposed by the authors for upward flow and downward flow is applicable, not only to the uniform heat flux case, but also to the nonuniform heat flux case.