The basic relations used in the HYDRA-IBRAE/LM/V1 code for calculating the flow 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 verification results. It is shown that the thermohydraulic code HYDRA-IBRAE/LV/V1 can be used to calculate correctly the primary processes in design basis and beyond-design basis accidents in sodium-cooled reactor facilities.
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Translated from Atomnaya Énergiya, Vol. 118, No. 6, pp. 309–313, June, 2015.
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Usov, E.V., Pribaturin, N.A., Kudashov, I.G. et al. A Step in the Verification of the Hydra-Ibrae/LM/V1 Thermohydraulic Code for Calculating Sodium Coolant Flow in Fuel-Rod Assemblies. At Energy 118, 382–388 (2015). https://doi.org/10.1007/s10512-015-0012-8
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DOI: https://doi.org/10.1007/s10512-015-0012-8