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:
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
A cylindrical mesh tally (MeshFilter on CylindricalMesh) 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() (or your favorite tally result getter) and plot it manually using matplotlib or (your favorite plotting library). Another is to use openmc-plotter’s tally visualization. Yet another is to export it to VTK and visualize it in Paraview (workshop example).
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?