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Add the ability to tally microscopic cross sections in void materials#3771

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GuySten wants to merge 3 commits intoopenmc-dev:developfrom
GuySten:void-micro
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Add the ability to tally microscopic cross sections in void materials#3771
GuySten wants to merge 3 commits intoopenmc-dev:developfrom
GuySten:void-micro

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@GuySten
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@GuySten GuySten commented Feb 4, 2026

Description

This PR add the ability to tally microscopic cross sections in void materials.

Fixes #3412.

Checklist

  • I have performed a self-review of my own code
  • I have run clang-format (version 15) on any C++ source files (if applicable)
  • I have followed the style guidelines for Python source files (if applicable)
  • I have made corresponding changes to the documentation (if applicable)
  • I have added tests that prove my fix is effective or that my feature works (if applicable)

@lewisgross1296
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Just for my own curiosity, what is the application/use case of this?

@GuySten
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GuySten commented Feb 4, 2026

It lets you calculate reaction rates in void regions without touching the model.
For example it can be used for calculation of equivalent thermal neutron flux (equivalent with respect to some thermal microscopic cross section).

@GuySten GuySten marked this pull request as draft February 4, 2026 21:15
@GuySten GuySten marked this pull request as ready for review February 5, 2026 04:13
@GuySten GuySten added the Tallies label Feb 5, 2026
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Calculation of micro-cross section in void for nuclides specified on tally

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