MC2-2. Fast Neutron Spectra & Multigroup X-Sections

MC**2-2 solves the neutron slowing-down equations using basic neutron data derived from ENDF/B data files to determine spectra for use in generating multigroup neutron cross sections. The current edition includes the ability to treat all ENDF/B-V representations, high-order PL scattering representations, a free-format input processor, isotope mixing, delayed neutron processing, and flexibility in output data selection.