- #1
marco_nuc
- 1
- 0
Hello there,
I am using mcnp6 to simulate a tokamak. I am interested in the energy deposition in the blanket and I am using a fmesh4 and the tally multiplier fm4 as follow:
fmesh4:n ORIGIN=0 -24.2 -50 OUT=CF
imesh=35.2 iints=352
jmesh=24.2 jints=484
kmesh=50 kints=1
fm4 (-1 1 1 -4)
Now, MCNP allows you to use only one material in the fm card, in my case material 1. Unfortunately, I have six different materials (in several different cells) in the mesh domain and I am wondering if there's a way to make mcnp aware of that such that it'll automatically consider different materials of different cells in the mesh domain. I could use six different fm4, one fore each material, and merge the six outputs in a single matrix representing my blanket but it would be really tedious. I hope someone can help. I have read about a wild-card material 0 but I can't figure out how that works :(
I am using mcnp6 to simulate a tokamak. I am interested in the energy deposition in the blanket and I am using a fmesh4 and the tally multiplier fm4 as follow:
fmesh4:n ORIGIN=0 -24.2 -50 OUT=CF
imesh=35.2 iints=352
jmesh=24.2 jints=484
kmesh=50 kints=1
fm4 (-1 1 1 -4)
Now, MCNP allows you to use only one material in the fm card, in my case material 1. Unfortunately, I have six different materials (in several different cells) in the mesh domain and I am wondering if there's a way to make mcnp aware of that such that it'll automatically consider different materials of different cells in the mesh domain. I could use six different fm4, one fore each material, and merge the six outputs in a single matrix representing my blanket but it would be really tedious. I hope someone can help. I have read about a wild-card material 0 but I can't figure out how that works :(