# About distribution flux

**URL:** <https://openmc.discourse.group/t/about-distribution-flux/6366>\
**Category:** User Support\
**Created:** [June 16, 2026, 9:50pm UTC](https://openmc.discourse.group/t/about-distribution-flux/6366 "2026-06-16T21:50:24Z")\
**Posts on this page:** 3\
**Page:** 1

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**Author:** ![DinaEl](https://avatars.discourse-cdn.com/v4/letter/d/77aa72/32.png) [@DinaEl](https://openmc.discourse.group/u/DinaEl)\
**Post date:** [June 16, 2026, 9:50pm UTC](https://openmc.discourse.group/t/about-distribution-flux/6366/1 "2026-06-16T21:50:24Z")

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Hello everyone !  
would like to calculate and plot the neutron flux distribution in both the axial and radial directions.

What is the recommended way to define the mesh tallies for:

1. Axial flux distribution (flux versus z)
2. Radial flux distribution (flux versus r)

Should I use a RegularMesh tally, a CylindricalMesh tally, or another approach?

If possible, could someone provide a simple example of the tally definition and the Python plotting procedure from the statepoint file?

Thanks

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**Author:** ![tjlaboss](https://yyz2.discourse-cdn.com/free1/user_avatar/openmc.discourse.group/tjlaboss/32/1390_2.png) [@tjlaboss](https://openmc.discourse.group/u/tjlaboss)\
**Post date:** [June 17, 2026, 10:27pm UTC](https://openmc.discourse.group/t/about-distribution-flux/6366/2 "2026-06-17T22:27:43Z")

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A cylindrical mesh tally ([`MeshFilter`](https://docs.openmc.org/en/stable/pythonapi/generated/openmc.MeshFilter.html) on [`CylindricalMesh`](https://docs.openmc.org/en/stable/pythonapi/generated/openmc.CylindricalMesh.html)) is indeed what you want. All you will need to do is create the `CylindricalMesh` with an origin at the bottom of your region of interest, and grid points in _r_ and _z_ (and \varphi if you want). The documentation is excellent and should make it intuitive.

The plotting procedure depends on how you want to visualize it. One way is to call [`Tally.get_pandas_dataframe()`](https://docs.openmc.org/en/stable/pythonapi/generated/openmc.Tally.html#openmc.Tally.get_pandas_dataframe) (or your favorite tally result getter) and plot it manually using `matplotlib` or (your favorite plotting library). Another is to use [openmc-plotter](https://github.com/openmc-dev/plotter#features)’s tally visualization. Yet another is to export it to VTK and visualize it in Paraview ([workshop example](https://github.com/fusion-energy/neutronics-workshop/blob/main/tasks/task_08_CSG_mesh_tally/3_example_cylinder_mesh_tallies.ipynb)).

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**Author:** ![DinaEl](https://avatars.discourse-cdn.com/v4/letter/d/77aa72/32.png) [@DinaEl](https://openmc.discourse.group/u/DinaEl)\
**Post date:** [June 21, 2026, 12:44pm UTC](https://openmc.discourse.group/t/about-distribution-flux/6366/3 "2026-06-21T12:44:52Z")

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Hello!  
Thank you for your explanation.

I am able to create a CylindricalMesh tally and obtain the flux values in the mesh. However, after extracting the tally results from the statepoint file, I am not sure how to process the data correctly to obtain and plot the radial flux (or power) distribution.

Could you please explain the steps after loading the statepoint file? In particular, how do I go from the mesh tally values to a radial profile similar to the plots usually shown for reactor flux or power distributions?