Experimental investigation and prediction of CHF in wire-wrapped rod bundles

Abstract

The wire-wrapped tight lattice fuel assembly is one of the options for small modular reactors. To clarify the characteristics of critical heat transfer for departure from nucleate boiling, experiments were conducted to investigate the critical heat flux in a wire-wrapped 19-rod bundle. The critical heat flux coincident with associated thermal–hydraulic parameters was obtained under typical pressurized water reactor conditions. Validation of data accuracy was performed to provide confidence in the experiment results. It was found that the critical heat flux in the wire-wrapped rod bundle is sensitive to the change in steam quality at some steam quality regions. Besides, the critical heat flux in wire-wrapped rod bundles has a similar reflection under the effects of local pressure, local mass flux, and steam quality. However, the current correlations tend to overestimate the critical heat flux for a wire-wrapped rod bundle. Hence, further development of KfK-3 correlations was proposed for better prediction of the critical heat flux in wire-wrapped bundles. This work is helpful for rod bundle designs of advanced pressurized water reactors.