Is there a way to retrieve depleted data at a specific burnup step and generate a new materials.xml file with the appropriate units?
I came across a suggestion in another post which involved using the following code:
results = openmc.deplete.Results(‘depletion_results.h5’)
materials_dep = results.export_to_materials(burnupstep) # ‘burnupstep’ is an integer value
However, upon inspecting the new materials.xml file, I noticed that the nuclides didn’t have the correct units, despite the xml file indicating that the values were in ‘ao’. What are the units in this file and is there a way to just export them in ‘ao’ at a particular burnup step?