kbat
commented
6 years ago OK, now I have mapped the standard low energy FLUKA names for all available nuclides in FLUKA (based on nuclide atomic number), so now neutrons below 20 MeV can be transported. Note that I use the standard FLUKA names only. As far as I understand the manual, this assigns the standard FLUKA cross sections to materials, i.e. the behaviour is as if we used the standard FLUKA names for elements without specifying the LOW-MAT cards. This should be fine for most of the cases.
However, Table 10.3 in the FLUKA manual suggests more cross sections, but it means that more information about a nuclide must be used in order to assign the correct cross section (i.e. not only the Z-number, but also mass number, temperature and neutron table identifier).
This detailed mapping is not yet implemented because I am not exactly sure how to do it. The problem is that MCNP provides wider choice of cross sections and it's not obvious for me how to map them in the elegant way to cross sections available in FLUKA.
Elements defined by the MATERIAL cards use non-standard names (starting with E). This means that FLUKA does not assign any low energy neutron libraries related to these materials unless the appropriate LOW-MAT cards are defined. Currently these cards are not (yet) put in the input deck, therefore neutron-related calculations will be wrong. As a safety measure, the defaults are set to EM-CASCADE (i.e. only e+- and gammas are transported). I plan to add the LOW-MAT cards in the near future.