Neutronic analysis of the fusion reactor ARC: Monte Carlo simulations with the Serpent code (dataset)

The input files of the spatially-uniform and non-uniform neutron sources used for the Monte Carlo simulations with the Serpent code and the CAD files for the geometry definition of the fusion reactor ARC used in the framework of the work published in the paper "Neutronic analysis of the fusion reactor ARC: Monte Carlo simulations with the Serpent code" (in Fusion Science and Technology). In the repository there are also the Python files used for the post processing of the results and the output files of Serpent with the main results of the neutronic simulations