Numerical Study on Subcooled Boiling and CHF Phenomenon in Eccentric Annular Channels

Abstract

In the period of the increasing energy crisis, nuclear energy is of particularly importantance. It is critical to understand the core behaviors of fuel assembly and ultimately increase the CHF by accurately predicting flow and heat transfer phenomena inside the rod bundle. In the reactor core, the flow is quite complicated, including single phase flow, subcooled two-phase flow and Critical Heat Flux(CHF). The precise computational analysis of how to enhance the safety of nuclear reactors and thereby reduce costs becomes especially significant. Using computational Fluid Dynamics (CFD), it is possible to extract three-dimensional characteristics such as pressure, velocity, and temperature, providing theoretical guidance for nuclear reactor design and operation. By comparing Bartolomei's experimental data to the validated two-phase boiling CFD models, the CFD codes with RPI boiling models are applied to the internal and external flow phenomena of the pipe. The subcooled boiling two phase flow cases are compared. The void fraction and heated wall temperature distribution along the flow direction are discussed and compared under internal and external flow and a comparative analysis is conducted for CHF. It was found that the cold wall of the annular pipe has a significant effect on boiling and CHF. The study further extends the comparison to eccentric annular pipe configurations, exploring the performances of subcooled boiling and CHF. The results offer valuable insights into the understanding of the flow and heat transfer in the rod bundle of fuel assembly.