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

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 thermohydrau...

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Published inAtomic 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
LanguageEnglish
Published New York Springer US 01.10.2015
Springer
Springer Nature B.V
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Summary: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.
Bibliography:ObjectType-Article-1
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ISSN:1063-4258
1573-8205
DOI:10.1007/s10512-015-0012-8