Depletion/burn-out of a material at a specific point?

User Support
1

Hi all,
I am trying to calculate the material composition within a box at a certain point in the vessel of a simulated reactor after the time reactor runs using the depletion tool. However, I am having trouble figuring out how I might isolate the point I want to look at. I don’t care about the fuel depletion or anything else. If the depletion functionality is not the best for this, would it be better to simply create a tally for that geometry at that point specifically and calculate the reaction rate and then calculate the material change by hand?
Thank you for any help,
Michael

2

Hi Michael,

Any material that contains a fissionable nuclide is automatically marked as “depletable”, but you can also manually mark other materials as being depletable. For example:

vessel = openmc.Material()
…
vessel.depletable = True

Then when you run a depletion calculation, you will be able to obtain the depleted composition for the vessel material at whatever times were specified in the calculation.

Best,
Paul