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MCNP F6 tally

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  1. Aug 8, 2016 #1
    Hi,

    I'm working on a MCNP simulation where I have to use F6 tallies. According to the manual: "In the F6 and +F6 tallies, material density is available for the chosen cells, and normalization is MeV/gm/source-particle."

    To which source-particles is this value normalized: the source-particles in the chosen cell or the source-particles of the whole simulation?

    (Example: Lets say I have two cells (cell1 and cell2) in the simulation, both tallied. If I would like to get cell1's MeV/gm energy density with which amount of source-particles do I have to multiply it with: those of cell1 or those of cell1+cell2?)
     
  2. jcsd
  3. Aug 12, 2016 #2
    it is per simulation source particle

    You should multiply it by the real world source strength.

    For example in a fusion reactor you might get 1e21 neutrons per second , so you would take the tally results and multiply it be 1e21 neutrons to get using of MeV per gram

    I typically go a little further and include volume, density and convert to joules in my calculations (I wish MCNP offered some choices about the units)
     
  4. Aug 13, 2016 #3
    Thank you very much!
     
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