OpenMC is an open source Monte Carlo code, and was developed based on state-of-the-art informatics and with the general aim of performing novel reactor physics analyses. It is capable of storing computed observables on Python-oriented databases, and is also capable of performing macroscopic cross-section homogenization which in turn, can be employed in deterministic models of core simulators.
Up to recently, the in core fuel management and irradiation studies of the accelerator driven system knwon as MYRRHA, has been based on the best estimate MCNP code. Nevertheless, since this is a general purpose particle transport code, it cannot create database conformed with the spatial distribution of the many multi-group and homogenized macroscopic cross-sections of the core. Thus, the creation of a database for few-group trasnport or diffusion core simulators, is not so straightforward with MCNP. For this reason, the usage of academic codes such as OpenMC that are reactor physics oriented, is intended in this project for the creation of a state-of-the-art database cross-section of the MYRRHA core.