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updated documentation and comments in Python API
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Sam Shaner
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7 changed files with 80 additions and 174 deletions
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@ -105,19 +105,6 @@ standard deviation.
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*Default*: false
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.. _cross_sections:
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``<cross_sections>`` Element
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----------------------------
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The ``<cross_sections>`` element has no attributes and simply indicates the path
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to an XML cross section listing file (usually named cross_sections.xml). If this
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element is absent from the settings.xml file, the
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:envvar:`OPENMC_CROSS_SECTIONS` environment variable will be used to find the
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path to the XML cross section listing when in continuous-energy mode, and the
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:envvar:`OPENMC_MG_CROSS_SECTIONS` environment variable will be used in
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multi-group mode.
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``<cutoff>`` Element
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--------------------
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@ -289,20 +276,6 @@ based on the recommended value in LA-UR-14-24530_.
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.. note:: This element is not used in the multi-group :ref:`energy_mode`.
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.. _multipole_library:
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``<multipole_library>`` Element
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-------------------------------
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The ``<multipole_library>`` element indicates the directory containing a
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windowed multipole library. If a windowed multipole library is available,
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OpenMC can use it for on-the-fly Doppler-broadening of resolved resonance range
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cross sections. If this element is absent from the settings.xml file, the
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:envvar:`OPENMC_MULTIPOLE_LIBRARY` environment variable will be used.
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.. note:: The <temperature_multipole> element must also be set to "true" for
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windowed multipole functionality.
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``<max_order>`` Element
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---------------------------
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@ -316,29 +289,6 @@ then, OpenMC will only use up to the :math:`P_1` data.
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.. note:: This element is not used in the continuous-energy
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:ref:`energy_mode`.
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.. _natural_elements:
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``<natural_elements>`` Element
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------------------------------
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The ``<natural_elements>`` element indicates to OpenMC what nuclides are
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available in the cross section library when expanding an ``<element>`` into
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separate isotopes (see :ref:`material`). The accepted values are:
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- ENDF/B-VII.0
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- ENDF/B-VII.1
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- JEFF-3.1.1
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- JEFF-3.1.2
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- JEFF-3.2
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- JENDL-3.2
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- JENDL-3.3
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- JENDL-4.0
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Note that the value is case-insensitive, so "ENDF/B-VII.1" is equivalent to
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"endf/b-vii.1".
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*Default*: ENDF/B-VII.1
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``<no_reduce>`` Element
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-----------------------
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@ -1368,6 +1318,33 @@ Here is an example of a properly defined 2d hexagonal lattice:
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Materials Specification -- materials.xml
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----------------------------------------
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.. _cross_sections:
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``<cross_sections>`` Element
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----------------------------
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The ``<cross_sections>`` element has no attributes and simply indicates the path
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to an XML cross section listing file (usually named cross_sections.xml). If this
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element is absent from the settings.xml file, the
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:envvar:`OPENMC_CROSS_SECTIONS` environment variable will be used to find the
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path to the XML cross section listing when in continuous-energy mode, and the
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:envvar:`OPENMC_MG_CROSS_SECTIONS` environment variable will be used in
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multi-group mode.
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.. _multipole_library:
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``<multipole_library>`` Element
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-------------------------------
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The ``<multipole_library>`` element indicates the directory containing a
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windowed multipole library. If a windowed multipole library is available,
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OpenMC can use it for on-the-fly Doppler-broadening of resolved resonance range
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cross sections. If this element is absent from the settings.xml file, the
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:envvar:`OPENMC_MULTIPOLE_LIBRARY` environment variable will be used.
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.. note:: The <temperature_multipole> element must also be set to "true" for
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windowed multipole functionality.
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.. _material:
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``<material>`` Element
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@ -1440,43 +1417,6 @@ Each ``material`` element can have the following attributes or sub-elements:
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.. note:: The ``scattering`` attribute/sub-element is not used in the
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multi-group :ref:`energy_mode`.
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:element:
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Specifies that a natural element is present in the material. The natural
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element is split up into individual isotopes based on `IUPAC Isotopic
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Compositions of the Elements 2009`_. This element has
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attributes/sub-elements called ``name``, and ``ao``. The ``name``
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attribute is the atomic symbol of the element. Finally, the ``ao``
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attribute specifies the atom percent of the element within the material,
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respectively. One example would be as follows:
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.. code-block:: xml
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<element name="Al" ao="8.7115e-03" />
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<element name="Mg" ao="1.5498e-04" />
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<element name="Mn" ao="2.7426e-05" />
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<element name="Cu" ao="1.6993e-04" />
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In some cross section libraries, certain naturally occurring isotopes do not
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have cross sections. The :ref:`natural_elements` option determines how a
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natural element is split into isotopes in these cases.
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*Default*: None
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An optional attribute/sub-element for each element is ``scattering``. This
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attribute may be set to "data" to use the scattering laws specified by the
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cross section library (default). Alternatively, when set to "iso-in-lab",
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the scattering laws are used to sample the outgoing energy but an
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isotropic-in-lab distribution is used to sample the outgoing angle at each
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scattering interaction. The ``scattering`` attribute may be most useful
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when using OpenMC to compute multi-group cross-sections for deterministic
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transport codes and to quantify the effects of anisotropic scattering.
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*Default*: None
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.. note:: The ``scattering`` attribute/sub-element is not used in the
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multi-group :ref:`energy_mode`.
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:sab:
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Associates an S(a,b) table with the material. This element has one
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attribute/sub-element called ``name``. The ``name`` attribute
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@ -1503,9 +1443,6 @@ Each ``material`` element can have the following attributes or sub-elements:
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*Default*: None
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.. _IUPAC Isotopic Compositions of the Elements 2009:
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http://pac.iupac.org/publications/pac/pdf/2011/pdf/8302x0397.pdf
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------------------------------------
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Tallies Specification -- tallies.xml
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------------------------------------
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