How to use mesh tally in MCNPX to calculate dose?

In summary, the conversation discusses using mesh tally to calculate dose in MCNPX. The speaker sets a cylinder with a specified material and then wants to divide it into smaller cylinders. They also mention the need to record the flux of each mesh and use a dedf card to obtain the dose. The dimensions of the cylinders are also mentioned. The conversation ends with the speaker asking for more details on how to do this in MCNPX and being directed to specific sections in the manual.
  • #1
LeeGru
6
0
Hello,guys,
I wonder how to use mesh tally to calculate dose.I set a cylinder，and set the material of the cylinder.Then I want to divide a cylinder into smaller cylinders in a direction perpendicular to the z-axis.And I need to record the flux of each mesh,use dedf card to get dose.
I have tried to write a little bit, please help to see if it is correct and what needs to be added?
The radius of big cylinder is 7cm,length of it is 10cm,the radius of little cylinder is 7cm,too,the length of little cylinder is 1mm.
TMESH
CMESH1:n
CORA1 0.0 7 99r
CORB1 0 100i 10
CORC1 0 179r 360
ENDMD
 
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  • #2
Hi,
after "CMESH1:n" you must put the keyword DOSE and the options (int iu and fact)
see 3.3.5.24.2 of the MCNP6 version 1.0 manual,
 
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Likes Astronuc
  • #3
PSRB191921 said:
Hi,
after "CMESH1:n" you must put the keyword DOSE and the options (int iu and fact)
see 3.3.5.24.2 of the MCNP6 version 1.0 manual,
Thank you!However,I don't have MCNP6.I'm using MCNPX,and could you tell me more details?
Thank you very much!
 
  • #4
Hi
in MCNPX 2.5.0 manual (LA-CP-05-0369) see 5.6.22.2
For an another version of MCNPX try to search "TRACK-AVERAGED MESH TALLY" in the manual
 

1. What is a mesh tally in MCNPX?

A mesh tally in MCNPX is a feature that allows for the calculation of dose or flux on a user-defined mesh grid. It can be used to obtain more detailed information about the radiation distribution in a specific area of interest.

2. How do I define a mesh tally in MCNPX?

To define a mesh tally in MCNPX, you need to specify the mesh grid size, position, and orientation in the input file. You also need to define the type of tally (e.g. dose, flux, or track-length) and the particle type to be scored.

3. What is the difference between a surface tally and a mesh tally in MCNPX?

A surface tally in MCNPX calculates the dose or flux on a specific surface, while a mesh tally calculates it on a user-defined mesh grid. This allows for more detailed and accurate results, as the mesh grid can be tailored to the specific geometry and materials of the problem.

4. How can I visualize the results from a mesh tally in MCNPX?

You can visualize the results from a mesh tally in MCNPX by using the built-in graphics capabilities or by exporting the data to a post-processing software such as VisIt or ParaView. These tools allow for the creation of 3D plots and visualizations of the dose or flux distribution on the mesh grid.

5. Can I use a mesh tally in MCNPX for any type of radiation?

Yes, a mesh tally in MCNPX can be used for any type of radiation, as long as the particle type is defined correctly in the input file. This includes photons, neutrons, electrons, and other particles commonly used in radiation transport simulations.

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