Abstract
The model of a heat-exchange crisis used in the PUChOK BM-DF code, intended for calculating by means of elementary cells the local thermohydraulic parameters of water and a steam–water mixture during stationary flow in fuel rod assemblies, was verified. The verification was done on experimental data obtained on the PSB RBMK thermohydraulic stand (Electrogorsk Research Center for Nuclear Power Plant Safety), which models one circuit of the multiple forced circulation loop in RBMK and includes a model of all main elements of the loop. The experimental investigation was performed in order to determine the effect of changes in the geometry of RBMK channels on the conditions for the appearance of a heat-exchange crisis. It was shown that in terms of the local parameters the computed critical power agrees with the experimental data.
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