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MCNP tally help

  1. Aug 24, 2016 #1
    How can i use F4 tally in hexagonal lattice geometry which composed of circular fuel assembly? For example one hexagonal lattice includes circle which is cell number 5 and another hexagonal lattice includes circle which is cell number 7 etc. I wanna use all these cells in one f4 tally.
    Please help
    Many thanks.....
     
  2. jcsd
  3. Aug 25, 2016 #2
    Do you want a separate tally bin for each cell or an average f4 tally for the same cell over all the lattice locations?
     
  4. Aug 25, 2016 #3
    I wanna average f4 tally for the same cell
     
  5. Aug 26, 2016 #4
    In that case its as straightforward as:
    F4:N (5<L1) (7<L2)
    Where L1 and L2 are the lattice numbers where cells 5 and 7 are located.​

    If you want an average flux for all cells 5 and 7 all over the geometry it'll be:
    F4:N (5 7)​
     
  6. Aug 26, 2016 #5
    i understand many many thanks.....:thumbup::thumbup::thumbup:
     
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