A Step in the Verification of the Hydra-Ibrae/LM/V1 Thermohydraulic Code for Calculating Sodium Coolant Flow in Fuel-Rod Assemblies

• Published in: Atomic energy (New York, N.Y.) Vol. 118; no. 6; pp. 382 - 388
• Main Authors: Usov, E. V., Pribaturin, N. A., Kudashov, I. G., Butov, A. A., Dugarov, G. A., Mosunova, N. A., Strizhov, V. F., Ivanov, E. N.
• Format: Journal Article
• Language: English
• Published: New York Springer US 01.10.2015; Springer; Springer Nature B.V
• Subjects: Analysis; Approximation; Assemblies; Boiling; Control theory; Coolants; Experiments; Friction; Hadrons; Heat conductivity; Heat exchange; Heat exchangers; Heat transfer; Heavy Ions; Lattice theory; Mathematical analysis; Nuclear Chemistry; Nuclear Energy; Nuclear fuels; Nuclear Physics; Nuclear reactors; Physics; Physics and Astronomy; Reynolds number; Sodium; Studies; Thermohydraulics; Fuel Element; Fuel Assembly; Heat Exchange; Liquid Metal; Nusselt Number
• ISSN: 1063-4258; 1573-8205
• DOI: 10.1007/s10512-015-0012-8

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.