FreeCAD based Monte Carlo modeling approach for fusion reactor facilities

Abstract

Abstract MCNPX is a general purpose multi-particle Monte Carlo transport code with board ranges of energies, which has already been widely employed in the fields of radiotherapy, nuclear engineering, and nuclear fusion research. The neutronics design and analysis of fusion reactor facilities using MCNPX have been considered as the time consuming and error prone tasks. For example, the ITER-type models include a lot of nested structures and complicated high-order curve surfaces, which can not be easily described in the mathematics expressions, and the spline surfaces are also not supported by the native syntax of MCNPX. In this context, a code system called CAD-PSMC (FreeCAD based parsing script for MCNPX) has been developed in IMPCAS (Institute of Modern Physics, Chinese academy of Sciences) for the study of modeling conversion problem in ITER-type neutron calculations. Two kinds of methods have been provided and discussed in the framework of this code, including the establishing of the mapping relationship between the specification tree and the build-in boolean operations in consideration of the complicated hierarchical structures, and the demonstration of fast ray-casting technology based iteration method coupling with the mesh-oriented optimized CAD geometries in solving the high-order curve surfaces related problems. Finally, the stability and feasibility of the CAD-PSMC have been verified and validated through the comparison with the selected reference models and the neutronic simulation results of the ITER-type benchmark model.