Hello everyone,
I’m working on calculating the distributed neutron flux in my OpenMC model, but my results are not coming out as expected. There seems to be an issue somewhere, but I’m not sure what exactly is causing it.
Could anyone please guide me on how to correctly calculate and visualize the distributed (spatial) flux in OpenMC? I want to confirm if I should use MeshFilter or another approach, and how to properly define the en
ergy bins or normalization for accurate results.
If needed, I upload my input files and images of the results I’m getting. Any advice or example code would be greatly appreciated.
flux.ipynb (155.3 KB)