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Comparison of simple design of sodium and lead cooled fast reactor cores

Mazgaj P., Darnowski P., Gurgacz S., Lipka M., Dziubanii K.

Journal of Power Technologies

2014

Vol. 94, nr 5

16--25

This report presents the results of a numerical simulation of thermal hydraulics processes in a liquid metal cooled fast reactor core, combined with simple neutron population computing for an infinite pin cell lattice. Two types of coolant were studied: liquid sodium and liquid lead, with all requirements regarding safety conditions observed. Temperature distributions along the cooling channel and distributions in the radial direction were prepared, then criticality calculations were performed for MOX fuel using MCNP Monte Carlo code.