Geometry can change during depletion calculation?

Hi all,

I’d like to simulate a core depletion using OpenMC. I have created all files and have set depletion parameters (thank you for the job: depletion setting is easy :slight_smile: ).
However, I didn’t see anywhere in the documentation the possibility to change the geometry during depletion calculation. Indeed parts of my geometry change (absorbers for example) to keep critical condition. Without this possibility, my results will be wrong. Did I miss something in the manual ? Or, is this possibility not currently implemented in OpenMC ?

Thx for your answers,

Regards

Julien

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Hello Julien,

Thank you for the kind words. Currently we do not support changing the geometry through depletion in a simple manner. There have been some discussions in https://github.com/openmc-dev/openmc/issues/1519 and in change the material in individual cells during a depletion calculation that may be of interest

Regards,

Andrew

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Hello,

Thank you for your answer. I’ll read the posts you gave me.

I was thinking about this problem, and I’m going to test the following possibility using current OPENMC capacities. Not an ideal possibility but it may enableto have first results :

  1. Creation of all geometry objects : one geometry per time-step (called for example geom1, geom2, geom3, …). Each geometry is different (absorbers move) to keep the core critical. I imagine I know the critical parameters (critical absorbers positions for example).

  2. Creation of first materials object: fresh one.

  3. First depletion with both initial geometry and fresh compositions.

  4. Recovery of depleted materials after 1st step, and conversion to a new materials object for 2nd step (materials2 object).

  5. Creation of new depletion Operator composed by both geom₂ object and loading of materials2 object

  6. Second depletion…

  7. And iteration until end of depletion timesteps.

All this encapsulated in an iterator loop, it enables to manage all this “small depletions” in a big depletion calculation.

The main difficulty I see here is the conversion of depleted material to new XML : I imagine OPENMC has some python methods to get depleted materials.

Finally, it can be imagined in a future OPENMC version to give to depletion operator a list/np_array of geometry objects with a geometry by time-step. Then OPENMC can change depletion operator with various geometries alone.

I imagine people have already thought and tried this solution : is there any feedback about something like that ?

Regards,

Julien

Julien,

That seems to be a pretty sensible approach. I think we (the core developers) are trying to figure out the best way to support these types of changes through depletion. Both for geometry and possibly material edits. There isn’t a great timeline at this point, but please let us know how your experience goes! Any insight would be great as we sort through this.

With respect to extracting the depleted material compositions, we don’t have a straight forward way for doing this. Yet! There are two active and similar pull requests that are angling to add this functionality. Could be a good inspiration point for your own work.

Regards,

Andrew