Hi! In this .inp file I set in MATERIAL card atomic number equal to 63 and charge number 29 and density equal to 8.92 g/cm^3, i. e. it is Cu63.
Isotope yields.inp (2.4 KB)
It’s irradiated by neutron beam of 1.5 MeV energy. In sum.lis file there are obtained some quantity of Cu65.
Isotope yields_21_sum.lis (2.9 KB)
How it’s possible if target is clean Cu63 ? If no, what should I do to have clean Cu63 target ?
For low energy neutrons, library data are used and these refer only to natural copper (see the material list in the manual, Chapter 10), independently of your monoisotopic material definition.
If you are interested in retrieving cross sections, look for instance at ENDF.