2. Materials Specification – materials.xml

2.1. <cross_sections> Element

The <cross_sections> element has no attributes and simply indicates the path to an XML cross section listing file (usually named cross_sections.xml). If this element is absent from the settings.xml file, the OPENMC_CROSS_SECTIONS environment variable will be used to find the path to the XML cross section listing when in continuous-energy mode, and the OPENMC_MG_CROSS_SECTIONS environment variable will be used in multi-group mode.

2.2. <multipole_library> Element

The <multipole_library> element indicates the directory containing a windowed multipole library. If a windowed multipole library is available, OpenMC can use it for on-the-fly Doppler-broadening of resolved resonance range cross sections. If this element is absent from the settings.xml file, the OPENMC_MULTIPOLE_LIBRARY environment variable will be used.

Note

The <temperature_multipole> element must also be set to “true” for windowed multipole functionality.

2.3. <material> Element

Each material element can have the following attributes or sub-elements:

id:

A unique integer that can be used to identify the material.
name:

An optional string name to identify the material in summary output files. This string is limited to 52 characters for formatting purposes.

Default: “”
temperature:

An element with no attributes which is used to set the default temperature of the material in Kelvin.

Default: If a material default temperature is not given and a cell temperature is not specified, the global default temperature is used.
density:

An element with attributes/sub-elements called value and units. The value attribute is the numeric value of the density while the units can be “g/cm3”, “kg/m3”, “atom/b-cm”, “atom/cm3”, or “sum”. The “sum” unit indicates that values appearing in ao or wo attributes for <nuclide> and <element> sub-elements are to be interpreted as absolute nuclide/element densities in atom/b-cm or g/cm3, and the total density of the material is taken as the sum of all nuclides/elements. The “macro” unit is used with a macroscopic quantity to indicate that the density is already included in the library and thus not needed here. However, if a value is provided for the value, then this is treated as a number density multiplier on the macroscopic cross sections in the multi-group data. This can be used, for example, when perturbing the density slightly.

Default: None

Note

A macroscopic quantity can not be used in conjunction with a nuclide, element, or sab quantity.
nuclide:

An element with attributes/sub-elements called name, and ao or wo. The name attribute is the name of the cross-section for a desired nuclide. Finally, the ao and wo attributes specify the atom or weight percent of that nuclide within the material, respectively. One example would be as follows:

<nuclide name="H1" ao="2.0" />
<nuclide name="O16" ao="1.0" />

Note

If one nuclide is specified in atom percent, all others must also be given in atom percent. The same applies for weight percentages.

An optional attribute/sub-element for each nuclide is scattering. This attribute may be set to “data” to use the scattering laws specified by the cross section library (default). Alternatively, when set to “iso-in-lab”, the scattering laws are used to sample the outgoing energy but an isotropic-in-lab distribution is used to sample the outgoing angle at each scattering interaction. The scattering attribute may be most useful when using OpenMC to compute multi-group cross-sections for deterministic transport codes and to quantify the effects of anisotropic scattering.

Default: None

Note

The scattering attribute/sub-element is not used in the multi-group <energy_mode> Element.
sab:

Associates an S(a,b) table with the material. This element has one attribute/sub-element called name. The name attribute is the name of the S(a,b) table that should be associated with the material.

Default: None

Note

This element is not used in the multi-group <energy_mode> Element.
macroscopic:

The macroscopic element is similar to the nuclide element, but, recognizes that some multi-group libraries may be providing material specific macroscopic cross sections instead of always providing nuclide specific data like in the continuous-energy case. To that end, the macroscopic element has one attribute/sub-element called name. The name attribute is the name of the cross-section for a desired nuclide. One example would be as follows:

<macroscopic name="UO2" />

Note

This element is only used in the multi-group <energy_mode> Element.

Default: None