Modeling the core of a lead-cooled reactor

Abstract

Abstract This paper presents the results of thermohydraulic modeling of the reactor core with a lead coolant, which is based on the calculation scheme for the ATHLET code, obtained on the basis of open information on the BREST-OD-300 reactor plant. Only the reactor core and methods of its modeling are considered. The work also presents the results on the development of the division of the in-reactor space of the BREST-OD-300 reactor into a system of parallel channels. The subdivision of the in-core space is based on the type and number of different core elements. This way of modeling the core allows you to see changes in different parts of the reactor when calculating transients. The developed model of the partitioning of the in-core space will be further used in the calculation of various transient modes (MCP shutdown, steam generator tube rupture, etc.). Neutron physics is not considered in this paper. There are plans to carry out joint neutron-physical and thermohydraulic calculations using the model from this paper.