Experimental visualization of flow structure inside subchannels of a 4 × 6 rod-bundle

Abstract

A fuel element geometry frequently used in nuclear reactors is the rod-bundle. In it, coolant flows axially through subchannels formed between the rods. The mixing of cooling fluid in a rod-bundle reduces the temperature differences in the coolant and along the perimeter of the rods. To ensure thermal performance of a nuclear reactor, detailed information about heat transfer and turbulent mixing taking place within subchannels is required. To improve the prediction ability of the subchannel analysis using computer code, a considerable amount of experimental data should be utilized for developing a new model for subchannel analysis. Experimental work has been conducted in the PRIUS-I (in-PWR Rod-bundle Investigation of Undeveloped mixing flow across Subchannel) test facility, which has adopted the matched index-of-refraction (MIR) technique to visualize the multi-dimensional velocity and turbulence fields in an unheated 4 × 6 rod-bundle geometry as well as to provide benchmark data for computer code validation. The 4 × 6 rod-bundle array has the same pitch-to-diameter (P/D) as the prototype fuel assembly. PRIUS-I test facility can simulate the uniform and non-uniform inlet flow conditions in order to validate the effect in crossflow between the fuel assemblies of PWR. In this study, the crossflow and mixing phenomena that exist between two adjacent subchannels through the gaps are visualized in any cross section of the rod-bundle geometry using a MIR technique. For non-uniform inlet velocity condition, the flow characteristics near the inlet where the flow mixing occurs actively is compared with computational fluid dynamics (CFD) calculation results. The experimental data can be used as benchmark data for the system safety analysis code or CFD code validation.