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126 lines
4.8 KiB
ReStructuredText
126 lines
4.8 KiB
ReStructuredText
.. _io_materials:
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========================================
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Materials Specification -- materials.xml
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========================================
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.. _cross_sections:
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----------------------------
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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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.. _material:
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----------------------
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``<material>`` Element
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----------------------
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Each ``material`` element can have the following attributes or sub-elements:
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:id:
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A unique integer that can be used to identify the material.
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:name:
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An optional string name to identify the material in summary output
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files.
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*Default*: ""
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:depletable:
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Boolean value indicating whether the material is depletable.
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:volume:
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Volume of the material in cm^3.
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:temperature:
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Temperature of the material in Kelvin.
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*Default*: If a material default temperature is not given and a cell
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temperature is not specified, the :ref:`global default temperature
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<temperature_default>` is used.
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:density:
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An element with attributes/sub-elements called ``value`` and ``units``. The
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``value`` attribute is the numeric value of the density while the ``units``
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can be "g/cm3", "kg/m3", "atom/b-cm", "atom/cm3", or "sum". The "sum" unit
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indicates that values appearing in ``ao`` or ``wo`` attributes for ``<nuclide>``
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and ``<element>`` sub-elements are to be interpreted as absolute nuclide/element
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densities in atom/b-cm or g/cm3, and the total density of the material is
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taken as the sum of all nuclides/elements. The "macro" unit is used with
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a ``macroscopic`` quantity to indicate that the density is already included
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in the library and thus not needed here. However, if a value is provided
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for the ``value``, then this is treated as a number density multiplier on
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the macroscopic cross sections in the multi-group data. This can be used,
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for example, when perturbing the density slightly.
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*Default*: None
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.. note:: A ``macroscopic`` quantity can not be used in conjunction with a
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``nuclide``, ``element``, or ``sab`` quantity.
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:nuclide:
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An element with attributes/sub-elements called ``name``, and ``ao``
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or ``wo``. The ``name`` attribute is the name of the cross-section for a
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desired nuclide. Finally, the ``ao`` and ``wo`` attributes specify the atom or
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weight percent of that nuclide within the material, respectively. One
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example would be as follows:
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.. code-block:: xml
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<nuclide name="H1" ao="2.0" />
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<nuclide name="O16" ao="1.0" />
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.. note:: If one nuclide is specified in atom percent, all others must also
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be given in atom percent. The same applies for weight percentages.
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*Default*: None
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:sab:
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Associates an S(a,b) table with the material. This element has an
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attribute/sub-element called ``name``. The ``name`` attribute
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is the name of the S(a,b) table that should be associated with the material.
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There is also an optional ``fraction`` element which indicates what fraction
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of the relevant nuclides will be affected by the S(a,b) table (e.g. which
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fraction of a material is crystalline versus amorphous). ``fraction``
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defaults to unity.
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*Default*: None
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.. note:: This element is not used in the multi-group :ref:`energy_mode`.
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:isotropic:
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The ``isotropic`` element indicates a list of nuclides for which elastic
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scattering should be treated as though it were isotropic in the laboratory
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system. This element may be most useful when using OpenMC to compute
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multi-group cross-sections for deterministic transport codes and to quantify
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the effects of anisotropic scattering.
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*Default*: No nuclides are treated as have isotropic elastic scattering.
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.. note:: This element is not used in the multi-group :ref:`energy_mode`.
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:macroscopic:
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The ``macroscopic`` element is similar to the ``nuclide`` element, but,
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recognizes that some multi-group libraries may be providing material
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specific macroscopic cross sections instead of always providing nuclide
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specific data like in the continuous-energy case. To that end, the
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macroscopic element has one attribute/sub-element called ``name``.
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The ``name`` attribute is the name of the cross-section for a
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desired nuclide. One example would be as follows:
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.. code-block:: xml
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<macroscopic name="UO2" />
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.. note:: This element is only used in the multi-group :ref:`energy_mode`.
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*Default*: None
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