Calculation of activation with mesh tally in MCNPX 2.6

[Gogy]

Hi, All!
Can anyone help me to construct a card for calculation of a spatial activation? Reaction is 27Al(n,p). I used a tmesh card for calculation of neutron flux as:

tmesh
rmesh21:n flux
cora21 190 9i 210
corb21 -10 9i 10
corc21 -10 9i 10
ergsh21 0 20
endmd

Somewhere in a problem I had signed
F14:n 1
FM14 1 99 103
M99 13027.24y 1

A manual teaches me that it is real to use my FM14 coupled with tmesh.
I tried:

tmesh
rmesh21:n mfact 14 1 1 1
cora21 190 9i 210
corb21 -10 9i 10
corc21 -10 9i 10
ergsh21 0 20
endmd

But my reward was... zeroes, zeroes, zeroes.
Another way is to set a function like
mshmf21 E1 F1 E2 F2........
An easy way. But MCNP doesn't alow more then 80 characters per string. Inside tmesh block symbol "&" doesn't work, so I'm unable to set all the Ei Fi pairs.
How can Fm marry tmesh? They are waiting for each other.

Please! Help them.

 

[Alex A]

According to the manual, this should work. FM is an allowed card for a type 1 mesh tally, so long as it is outside the mesh data block.

I have mixed information on if it should start "FMn -1 ..." or not. It is suggested that forces MCNP to use the cell density. Either way it should not produce a zero answer.

That is a very old library, but that is the right way to use it. 103 is the correct MT for (n,p), it means something different in newer libs, but that should work.

Can you post the input file you tried, or if you can't give it out, a minimal input file that demonstrates the problem? If you rename to make it have .txt you can add as an attachment.

 

[Gogy]

Thank you, Alex, for your trying to help!
Here is an example. It can show a principal problem.
Of course, *****.24y is an old library, but it's structure is the same that newer libs have.
24y here is only for example.

 

[Alex A]

I was getting a lot of weirdness, x was spinning forever on the mesh and getting nothing done. I hacked off the mesh and a lot of invisible chars like spaces and tabs.
As a sanity check, I ran mcnp ixz, and did
"xs=13027.24y mt=103"
Which suggests the cross section for the reaction may not be tabulated or be zero at 1 MeV.

So there is that.

I will have another go later.

 

[Alex A]

I cannot vouch for any of the numbers being meaningful, but they are non zero. This is a 2x2x2 mesh that is successfully modified by the fm card, as well as a regular tally modified in the same way by another fm card. The source neutron energy is set to 10 MeV because these have a tabulated non zero (n,p) cross section in the lib.

 

[Gogy]

I cannot vouch for any of the numbers being meaningful, but they are non zero. This is a 2x2x2 mesh that is successfully modified by the fm card, as well as a regular tally modified in the same way by another fm card. The source neutron energy is set to 10 MeV because these have a tabulated non zero (n,p) cross section in the lib.

Alex, it was great!
I had splitted your mesh for 10x10x10. Then in a void geometry for elementary cell of the mesh at the 90 cm distance from the point isotropic source a flux was determined (9.70884E-6 vs 9.824E-6 for 1/R^2 dependence). Accepted. Inspecting of 13027.24y reveals a meaning of 0.104 for mt=103 cross section (for 10 MeV). After that FM spell was casted. A miracle! A meaning of activation for this elementary cell became 1.00972E-6.
1.00972E-6 = 9.70884E-6*0.104 exactly.
A MESH SECRET is out!
Alex, thank you very much!!!

P.S. A lot of invisible chars could originate for ctrl_A + ctrl_C from VISED "Input file" window.