How to obtained the multi-group homogenized cross-sections using OpenMC

Hi all,

I know that the OpenMC can obtained the multi-group cross-section for each cell in heterogeneous pin or assembly. However, I don’t know whether it can give the homogenized cross-section of the heterogeneous pin or assembly without any other external procedures. Could you give me some advice? Thank you.

Hey Yahui,

You should be able to set the multigroup cross section tally up like any other tally. You should specify the geometric and material delineations with a list of filters.

Good luck!

Just to add in here – OpenMC allows you to calculate multigroup cross sections in a specific universe, or a cell that contains a universe or a lattice.

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