when we use the mutigroup decay-rate tally, we only need to give the number of the groups we want, but how the delayed neutron groups were divided?
it seems depends on the cross section library（endf or jeff）
when we use the mutigroup decay-rate tally, we only need to give the number of the groups we want, but how the delayed neutron groups were divided?
it seems depends on the cross section library（endf or jeff）