In the multigroup transport scheme, the production of secondary neutrons via (n,xn) reactions is taken into account implicitly by the so-called "non-absorption probability", a group-dependent factor by which the weight of a neutron is multiplied after exiting a collision. If the only possible reactions are capture and scattering, the non-absorption probability is < 1, but at energies above the threshold for (n,2n) reaction it can take values larger than 1. Fission neutrons, however, are treated separately and created explicitly using a group-dependent fission probability. They are assumed to be emitted isotropically and their energy is sampled from the fission spectrum appropriate for the relevant isotope and neutron energy. The fission neutron multiplicity is obtained separately from data extracted from European, American and Japanese databases.