MCNP Simulation of Tokamak Energy Deposition

[marco_nuc]

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 :(

 

[Achy47]

Hello,

Thank you for your post. It seems like you are facing a common issue in MCNP simulations, where you have multiple materials in your mesh domain and want to accurately calculate the energy deposition in each material. One potential solution to this problem is to use the FMESH4 card with the MATERIAL modifier, which allows you to specify different materials for each mesh cell.

The syntax for this card is:

FMESH4:n ORIGIN=ox oy oz OUT=CF
imesh=ix iints=nix
jmesh=jy jints=niy
kmesh=kz kints=niz
MATERIAL=(mat1,mat2,mat3,...)

This will assign mat1 to the first mesh cell, mat2 to the second mesh cell, and so on. You can also use the WILD card material 0 to assign the default material to all other mesh cells.

Additionally, you can use the FMESH:R and FMESH:RZ cards to define a region or region-z mesh, respectively, which allows you to define different materials for each region.

I hope this helps you in your simulation. If you have any further questions, please feel free to ask. Best of luck with your research!