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Designed to be subclassed by helpers that aim to compute effective fission yields with tallied data. Constructs a tally in all burnable materials, scoring the fission rate in all nuclides that have non-zero density (as reported by the Operator) and have multiple sets of yields. For nuclides with non-zero densities and a single set of yields, those yields are assumed to be constant across incident neutron energies. The unpack method is now abstract, as each subclass should have its own data layout, some with different energy structures perhaps?? Maybe some weighting functions? |
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| .. | ||
| base.rst | ||
| capi.rst | ||
| data.rst | ||
| deplete.rst | ||
| examples.rst | ||
| index.rst | ||
| mgxs.rst | ||
| model.rst | ||
| openmoc.rst | ||
| stats.rst | ||