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Add useful warnings for add_nuclide/add_element

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Paul Romano 2018-03-02 10:36:08 -06:00
parent 395a5a53d0
commit 0508274a3d
6 changed files with 38 additions and 42 deletions
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14
docs/source/usersguide/materials.rst
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@ -43,14 +43,22 @@ of an element, you specify the element itself. For example,
Internally, OpenMC stores data on the atomic masses and natural abundances of
all known isotopes and then uses this data to determine what isotopes should be
added to the material. When the material is later exported to XML for use by the
:ref:`scripts_openmc` executable, you'll see that any natural elements are
:ref:`scripts_openmc` executable, you'll see that any natural elements were
expanded to the naturally-occurring isotopes.
The :meth:`Material.add_element` method can also be used to add uranium at a
specified enrichment through the `enrichment` argument. For example, the
following would add 3.2% enriched uranium to a material::
mat.add_element('U', 1.0, enrichment=3.2)
In addition to U235 and U238, concentrations of U234 and U236 will be present
and are determined through a correlation based on measured data.
Often, cross section libraries don't actually have all naturally-occurring
isotopes for a given element. For example, in ENDF/B-VII.1, cross section
evaluations are given for O16 and O17 but not for O18. If OpenMC is aware of
what cross sections you will be using (either through the
:attr:`Materials.cross_sections` attribute or the
what cross sections you will be using (through the
:envvar:`OPENMC_CROSS_SECTIONS` environment variable), it will attempt to only
put isotopes in your model for which you have cross section data. In the case of
oxygen in ENDF/B-VII.1, the abundance of O18 would end up being lumped with O16.