MCNP Based Calculation of Reactivity Loss due to Fuel Circulation in Molten Salt Reactors

In molten salt reactors, the delayed neutron precursors continuously change their position in the reactor and primary loop due to the circulation of the fuel. Therefore, the influence of the delayed neutrons on the reactivity differs from that of solid fuel reactors. The paper describes a method, which is developed in order to modify the program MCNP so that it can take into account the transport of the delayed neutron precursors, and therefore it is capable to calculate the reactivity loss of reactors with circulating fuel. With the aid of the modified version of MCNP, calculations were performed on a simple homogeneous reactor with cylindrical core. The method was also applied to perform calculations on the reactivity loss of the MSRE (Molten-Salt Reactor Experiment, Oak Ridge National Laboratories, 1960s). The results obtained are in good agreement with the theoretical predictions and the measured reactivity loss values