Tallying arbitrary MT number #179
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ChasingNeutrons
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Oct 24, 2025
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Looks good, just the handful of aesthetic things to fix.
NuclearData/ceNeutronData/aceDatabase/aceNeutronNuclide_class.f90
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| !! NOTE: despite being in the interface, this function only makes sense | ||
| !! for CE. The MG extension returns a fatalError if called | ||
| !! | ||
| function getMTxs(self, MT, p) result(xs) |
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Why can this not work for MG? Can't certain MT numbers be included/excluded?
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In a MG material you can only look up the cross sections you provide. Those are legit tally responses, but are not provided by getMTxs (they are looked up from the xss Packages directly). Otherwise, the ENDF MT numbers don't exist in MG.
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Adding the capability to micro and macro Responses to get a larger set of MT numbers as input.