Abstract
The helical cruciform fuel rod is a new fuel design. Its advantages include a large surface area–to-volume ratio, short thermal conductivity distance, and no need for grid spacers. This new fuel rod can effectively improve the hydraulic performance of nuclear reactors. To study the performance of the helical cruciform fuel assembly, the subcooled boiling flow and heat transfer characteristics of this assembly are analyzed in the present work based on computational fluid dynamics. The results indicate that the temperature distribution of the central rod wall surface in the circumferential direction has inhomogeneity and periodicity. The fluid’s temperature and velocity distribution in the cross section are high in the center and low elsewhere, and the fuel rod’s torsional orientation is compatible with the velocity vector’s direction. The vapor volume fraction on the wall of the center rod of the fuel assembly is the highest, and the vapor volume fraction in the mainstream area is relatively low. This work provides a reference for further research on helical cruciform fuel assemblies in the thermal analysis of nuclear reactors.
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