Could openmc not generate multi-group cross sections in fixed source mode?

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Dear all ：
I have successfully run the example of generating multi-group cross-sections in k eigenvalue simulation. And I tried to generate multi-group cross-sections in fixed source mode, program code is shown in the attachment. But the program reports an error directly, and the error is reported as shown in the screenshot below.
Could openmc not generate multi-group cross-sections in fixed source mode? @paulromano
Best wishes,
William Xia
MUSE4_RY_mgxs_Library_external_source_material.py (7.8 KB)

image
image966×405 43.6 KB

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@William_Xia This does appear to be a bug. In particular, if you try to score the fission reaction rate with an outgoing energy filter in a fixed source simulation, it crashes. I’m going to look into a fix for this and will report back once I have something working.

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Bug fix has been submitted: Two small bug fixes (fixed source nu-fission tally, stochastic volume calc) by paulromano · Pull Request #1828 · openmc-dev/openmc · GitHub