Obtaining nuclide concentrations from depletion results file

[wahidluthfi]

Try to use results.export_to_materials(burnup_index=)
and I think this discussion topic could help you

Data extraction of the OpenMC depletion results

Right now the Results class doesn’t provide any methods that give that information in an easy way. What I would recommend is to export your results to a new Materials object and then use methods on the Material class. This would look something like: # Export materials from the last timestep materials = results.export_to_materials(-1) # materials is a list of Material objects, so you'll need to get the right index # for the material of interest material = materials[index] # Get number of at…