What’s New in 0.12.0¶
This release of OpenMC includes an assortment of new features and many bug fixes.
In particular, the
openmc.deplete module has been heavily tested which
has resulted in a number of usability improvements, bug fixes, and other
enhancements. Energy deposition calculations, particularly for coupled
neutron-photon simulations, have been improved as well.
Improvements in modeling capabilities continue to be added to the code,
including the ability to rotate surfaces in the Python API, several new
“composite” surfaces, a variety of new methods on
unstructured mesh tallies that leverage the existing DAGMC infrastructure,
effective dose coefficients from ICRP-116, and a new cell instance tally
All surfaces now have a rotate method that allows them to be rotated.
Several “composite” surfaces, which are actually composed of multiple surfaces but can be treated as a normal surface through the -/+ unary operators, have been added. These include:
Various improvements related to depletion:
The matrix exponential solver can now be configured through the solver argument on depletion integrator classes.
openmc.deplete.Chain.reduce()method can automatically reduce the number of nuclides in a depletion chain.
Depletion integrator classes now allow a user to specify timesteps in several units (s, min, h, d, MWd/kg).
openmc.deplete.ResultsList.get_atoms()now allows a user to obtain depleted material compositions in atom/b-cm.
Several new methods on
openmc.Material.add_elements_from_formula()method allows a user to create a material based on a chemical formula.
openmc.Material.add_element()now supports the enrichment argument for non-uranium elements when only two isotopes are naturally occurring.
openmc.Material.add_element()now supports adding elements by name rather than by symbol.
openmc.Material.get_elements()method returns a list of elements within a material.
openmc.Material.mix_materials()method allows multiple materials to be mixed together based on atom, weight, or volume fractions.
The acceptable number of lost particles can now be configured through
Delayed photons produced from fission are now accounted for by default by scaling the yield of prompt fission photons. This behavior can be modified through the
A trigger can now be specified for a volume calculation via the
Custom external source distributions can be used via the
Unstructured mesh class,
openmc.UnstructuredMesh, that can be used in tallies.
This release contains new contributions from the following people: