This is the investigation into The generation of high-fidelity multigroup multiband cross sections from Monte Carlo neutron transport simulations. Previous methods for generating multigroup multiband (MGMB) cross sections, and multigroup cross sections, assume an approximate shape for the scalar flux. This approximate flux shape is the product of an energy-dependent spectrum and a cross-section-dependent self-shielding factor for the material of interest. Since these methods assume a scalar flux a priori, the cross sections generated may not be sufficiently consistent with the actual scalar flux of any particular problem. Instead, in our approach we tally both the reaction rate and the flux for each energy group and cross section magnitude band using the Mercury Monte Carlo particle transport code and compute the group-band cross sections from these tallies. This approach eliminates the need for assuming a flux by using a Monte Carlo simulation to calculate these cross sections.