New burnup calculation system for fusion–fission hybrid reactor

Abstract

Fusion–fission hybrid reactor is a new energy system based on a fusion reactor. Recently, it is thought to be able to apply to nuclear waste incineration. For the neutronics analysis, the author's group developed their own neutronics and burnup calculation system, which consists of a general-purpose Monte Carlo code MCNP-4B and a point burnup code ORIGEN2. Since the fusion–fission hybrid reactor is operated in a sub-critical state, the cross-section library attached in ORIGEN2 cannot be utilized as well known. One-group cross-sections for burnup calculation had to, thus, be evaluated by a tally function of MCNP in our calculation system. However, all the necessary nuclides could not be dealt with appropriately by the above procedure because of several limitations of MCNP. In the present study, a new procedure has thus been developed to collapse cross-sections by directly using tracking information of neutron accumulated in MCNP calculation. From the test calculation, it was found that the new collapsing procedure successfully removed approximation in the previous evaluation method of one-group cross-sections, which was employed in the calculation system the authors developed previously. Also, ongoing feasibility studies for speedup of the system were briefly discussed, because this kind of cross-section evaluation procedure requires a long computation time at present. The present procedure would be a general method to evaluate collapsed cross-section for burnup with a transport calculation result of continuous energy Monte Carlo code.