Comparison of Simple Design of Sodium and Lead Cooled Fast Reactor Cores

Piotr Mazgaj, Piotr Darnowski, Sebastian Gurgacz, Maciej Lipka, Karolina Dziubanii

Abstract


Abstract The purpose of this report was to present the results of a numerical simulation of thermal hydraulics processes in the liquid metal cooled fast reactor core, combined with the simple neutron population computing for infinite pin cell lattice. Two types of the coolant have been studied: liquid sodium and the liquid lead, with all requirements regarded to safety conditions. Temperature distributions along the cooling channel and distributions in radial direction have been prepared and in the next step the criticality calculations using MCNP Monte Carlo code for MOX fuel have been conducted.

Keywords


lead fast reactor; sodium fast reactor; pin design; MCNP

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References


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