Fission product

Since not all the relevant cross-sections for the production and disappearance of fission products are included in the MCNP data base (ENDF-VI), one has to use a kind of ‘mean fission product’; the cross-sections for all fission fragments have to be analysed, together with the mass distribution of fission fragments for each of the fissile nuclei relevant to the study to evaluate the mean fission product and its cross-section (a linear combination of the data available in ENDF-VI is used to reproduce the mean values).[36] A time evolution study of these average cross-sections for the fissile nuclei has to be done to modify, if necessary, the linear combination after exposure to the neutron flux. In fast spectra, these cross-sections are generally more or less stable, but in a thermal case, the variation of these cross-sections being much faster, one has to modify the linear combination.