Getting number densities and microscopic cross sections on homogenized mesh cells

Hi there,

Can OpenMC provide homogenized number densities and microscopic cross sections on mesh cells (each homogenized cell contains heterogeneous geometry and materials)?

I attempted to generate the homogenized microscopic cross sections using both openmc.mgxs and openmc.mgxs.library on RegularMesh, but it seems not working because it only gives the following error.

Unable to get micro xs for mesh domain since the mesh cells do not know the nuclide densities in each mesh cell.

I would appreciate your any comments and advices if I am making any mistakes.

Thank you.

Unfortunately as the error message indicates, getting microscopic cross sections on a mesh is not currently supported. Iā€™m not aware of anyone working on adding support for this either.

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Hi @paulromano, Okay, I see. Thank you for your reply.

Is it possible to tally siga*flux and flux, then generate macro cross section manually, and then obtain micro cross section by divided with number density?

Hi @jerrylizy,

Thanks for your concern. That one you mentioned works on single material or geometry.
What I tried to do was to get the homogenized quantities on the mesh grid, seems not working for this.


Yes, my method may only work for single material compositon.

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