Neutronics Analysis for MSR Cell with Different Fuel Salt Channel Geometry.

The neutronic properties of Molten Salt Reactor are different from that of traditional solid-fuel reactors due to its nuclear fuel particularity. Based upon MCNP code, the influence of the size and shape of fuel salt channel on neutron physics of MSR cell was studied systematically in this work. The results show that the infinite multiplication factors increases first and then decreases with the change of graphite cell size under the condition of given fuel volume fraction. In the case of the same FVF and average chord length, when the average chord length is relatively small, the k values with different fuel salt channel shapes are in good agreement; when the average chord length is relatively large, the k values with different fuel salt channel shapes are greatly different. In addition, some examples of practical application of this work are illustrated in the end, including cell selection for the core and thermal expansion displacement analysis of the cell.