Search - Number 22 - THE PROBLEMS OF CALCULATION OF HEAT TRANSFER CRISIS IN FUEL ASSEMBLIES OF PW REACTORS BASED ON MODERN VERSIONS OF THERMOHYDRAULIC CODES

THE PROBLEMS OF CALCULATION OF HEAT TRANSFER CRISIS IN FUEL ASSEMBLIES OF PW REACTORS BASED ON MODERN VERSIONS OF THERMOHYDRAULIC CODES
Cover not present THE PROBLEMS OF CALCULATION OF HEAT TRANSFER CRISIS IN FUEL ASSEMBLIES  OF PW REACTORS BASED ON MODERN VERSIONS OF THERMOHYDRAULIC CODES
Category: Number 22
Publication: 22
Summary

This article gives an analysis of the adequacy of computer software systems FASCICLE BM-DF and COBRA, which are designed to calculate the main parameters of the safety of water-cooled nuclear reactors. This calculation is based on determining the local thermal-hydraulic parameters of the flow of coolant in the fuel rod assembled elements. In this article introduced the results of the comparison of experiments performed to determine the distribution of the main thermal-hydraulic flow parameters characteristic of subchannels of fuel rod assembled elements with the data for calculating these parameters on the basis of declared computer codes. Particular attention is paid to the analysis of experimental and calculated data, by definition, burnout in rod fuel assembled elements. It the article is shown the possibility of a reliable determination of this important parameter of a nuclear reactor safety through the use of two-dimensional computer code FASCICLE BM-DF and COBRA.

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