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.

Default: None
sab:

Associates an S(a,b) table with the material. This element has an attribute/sub-element called name. The name attribute is the name of the S(a,b) table that should be associated with the material. There is also an optional fraction element which indicates what fraction of the relevant nuclides will be affected by the S(a,b) table (e.g. which fraction of a material is crystalline versus amorphous). fraction defaults to unity.

Default: None

Note

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

The isotropic element indicates a list of nuclides for which elastic scattering should be treated as though it were isotropic in the laboratory system. This element 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: No nuclides are treated as have isotropic elastic scattering.

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