I am trying to calculate power density in the reactor core using mesh tally. I have used the FM card & got the results. Instead of explicitly setting material number on FM card (61, in my case), I want MCNP to track the material itself based upon the mesh being sampled! Is that possible somehow?
snippet of my input file is, C TALLY FOR POWER DENSITY FMESH14:N GEOM=CYL ORIGIN=0 0 -351.818 IMESH=6.5 13 19 25 39 50 60 70.5 83.5 90 IINTS= 1 1 1 1 1 1 1 1 1 1 JMESH=7.5 25.5 43.5 61.5 79.5 97.5 115.5 133.5 151.5 169.5 187.5 JINTS= 1 1 1 1 1 1 1 1 1 1 1 KMESH=1 KINTS=1 AXS=0 0 1 VEC=1 0 0 FM14 (1 61 -6)
Hi Sohail Ahmad Raza,
you can also try replacing the material number with zero; i.e. FM14(1 0 -6). By doing so, MCNP will use the reaction cross
sections for the material in which the particle is traveling.
You will also need to add -8 in the FM card; this will give you the equivalent of tally F7. See "2.5.4.1 Equivalence of F4, F6, and F7 Tallies" in the MCNP6.3 manual that is available online.
Regards,
dimitris
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