The basic relations used in the HYDRA-IBRAE/LM/V1 code for calculating the fl ow and heat-exchange of sodium coolant in fuel-rod assemblies in regimes with and without boiling and in the presence of a crisis of heat exchange are presented. The experimental data suitable for verifying the thermohydraulic code for calculating friction in one- and two-phase regimes as well as heat exchange in regimes with and without boiling are picked on the basis of a review of the literature. The computational error of different parameters is obtained on the basis of the verifi cation results. It is shown that the thermohydraulic code HYDRAIBRAE/LV/V1 can be used to calculate correctly the primary processes in design basis and beyond-design basis accidents in sodium-cooled reactor facilities.
