Hi all,
There are some specifications of tally in openmc user’s guide. For example, the “flux” tally has the unit of “neutron-cm per source particle”. However, the correct unit of “flux” is neutron/(m-2s-1). So, how to transform the unit of tally so that we can get the real “flux”？
Similarly, how can we get the total neutron production rate: < musigma*phi >
Best regards
Benjamin