## MCNP - Measuring Neutron Absorption in a Moderator

Hello all. I'm am a first time poster but a long time visitor. I am having a little trouble that I was hoping someone far wiser and more knowledgeable than myself might be able to help with.

I've been using MCNP to investigate criticality in a simple geometry consisting of a central natural uranium sphere in a cylindrical container (i.e. a barrel). Inside the container and surrounding the sphere is a moderator, either water of graphite. I am attempting to investigate the neutron economy of the moderators. I have been told that the best way to do this is to have MCNP record how many neutrons are absorbed in the moderator. I have also been told that there should be some sort of tally to do this.

However, despite scouring over various primers and manuals until my eyes bled, I have been unable to find any such tally (or at least I have not found any tally I have thought capable of making this measurement). I would appreciate it if anyone could let me know what the code would be? I would also welcome any suggestions of better ways to investigate neutron economy of moderators, should there be one.

Thank you very much for your assistance.
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 Hi... one way to do this is to calculate flux (f4) in the cell you are interested: for example: f4:n 101 (were 101 is the cell number where you want to calculate flux) If you would like to know the absorption reaction rates in that cell you then write the following tally multiplier, for example: f4m:n -1 10 (-2:-6) this way you will calculate total absorption reaction rates (absorption+fission, since absorption (-2) in MCNP is just the capture+fissin (-6)), 10 is the number of the material in that cell and -1 is the multiplier (atom density of thet material)...you can use -1 or enter the atom density of that material... be careful using tallys, when you calculate flux there must always be 4 the last number... you can use the following numbers... f04, f14, f24, f34,... I hope this will help... best regards