OpenMC mat fix #1076
OpenMC mat fix #1076
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OpenMC does need you to expand elements before writing XML. The only cases where it accepts elemental data is if the nuclear data library itself uses elemental data. The most common example is carbon which has historically been given as an elemental evaluation (although evaluations for C12 and C13 have now become available in ENDF/B-VIII.0). Other than carbon, everything else should probably be split out into nuclides. The way we handle this in OpenMC's Python API is to look at the user's OPENMC_CROSS_SECTIONS environment variable to figure out what data they actually have available. Not sure you'd really want to have that kind of logic here though...
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Welp. That was rather silly of me. I ran into that carbon issue when converting a DAGMC model and generalized it without looking more closely. I'll include a corner case for carbon for now. |
Ignoring Carbon for expansion in the OpenMC material writer.
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I've updated the writer and added the option of passing a set of nucids to ignore in |
pyne/material.pyx
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| Parameters | ||
| ---------- | ||
| nucset : set, optional |
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Indentation looks like it got messed up here (and below as well)
pyne/material.pyx
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| """expand_elements(self) | ||
| Exapnds the elements ('U', 'C', etc) in the material by replacing them | ||
| with their natural isotopic distributions. This function returns a copy. | ||
| Exapnds the elements ('U', 'C', etc) in the material by |
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Any further comments @gonuke? |
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Already approved... |
This PR updates OpenMC material writing in the following ways:
It adds tests which: