Hello everyone, I am running a coupled neutron-gamma problem, with a point source emitting neutrons, which subsequently interact in materials through inelastic or thermal neutron capture reactions. I am trying to tally the gammas emitted by those reactions, when they reach a certain surface in the geometry (“current” tally), which is in a different cell than the production site. I can successfully get the total gamma current, but how can I tag the gammas according to the nuclide that produced them? I have tried to use the nuclide filtering option for the tally, but it doesn’t seem to make any difference in the result. Am I doing something wrong (I am a new user), or is this capability not present? Thank you.