Neutron reactions on fusion reactor materials are key phenomena to be understood to enable fusion as a feasible energy source for future reactors. We present significant improvements and validations of NEUTRO, a deterministic neutron transport code dedicated to solving the Boltzmann Transport Equation. The code is integrated as a module in the Alya system developed by the Barcelona Supercomputing Center and uses the Discrete Ordinates Method over an angular portion, multigroup for energy discretization and the Finite Element Method over unstructured meshes to treat special complex domains. Material anisotropy of the scattering medium is introduced into the scattering kernel using real base expressions for spherical harmonics. In order to build the total cross-section and the respective group matrix for the elastic cross-section, we use the NJOY code. We test the solver using different geometries and materials with varying levels of scattering properties. We compare our results on classic tests, benchmarks obtained from a Nuclear Energy Agency database and test cases using other neutron transport codes.