This paper presents a methodology for 3D neutronic calculations suitable for complex and extensive geometries. The geometry of the system design is first fully modeled with a CAD program; this modeling is then processed - requiring few simplifications - with MCNP-CAD interface in order to generate a MCNP geometry file. Neutronic irradiation results are finally achieved running the MCNP program, where the geometry input card used is directly the MCNP-CAD interface output. This methodology enables accurate neutronic calculations for complex geometries characterized by high detail levels, such as ITER or other fusion facilities (IFMIF), in which we are presently involved.

This procedure has been applied to the Fast Ignition Fusion Reactor KOYO-F. We have determined the neutron fluxes and energy deposition in the reactor blanket, and obtained the front panel damage and activation for several alternative front panel materials. To carry out this calculation, KOYO-F blanket design is modeled using CATIA V5, and the selected CAD-MCNP interface is MCAM, developed by the FDS Team (China). The activation of the front panel material is finally evaluated with our code ACAB, based on the neutronic irradiation results provided by MCNP.