openmc.Materials

class openmc.Materials(materials=None)[source]

Collection of Materials used for an OpenMC simulation.

This class corresponds directly to the materials.xml input file. It can be thought of as a normal Python list where each member is a Material. It behaves like a list as the following example demonstrates:

>>> fuel = openmc.Material()
>>> clad = openmc.Material()
>>> water = openmc.Material()
>>> m = openmc.Materials([fuel])
>>> m.append(water)
>>> m += [clad]
Parameters

materials (Iterable of openmc.Material) – Materials to add to the collection

Variables

cross_sections (str or path-like) – Indicates the path to an XML cross section listing file (usually named cross_sections.xml). If it is not set, the OPENMC_CROSS_SECTIONS environment variable will be used for continuous-energy calculations and OPENMC_MG_CROSS_SECTIONS will be used for multi-group calculations to find the path to the HDF5 cross section file.

append(material)[source]

Append material to collection

Parameters

material (openmc.Material) – Material to append

export_to_xml(path: Union[str, PathLike] = 'materials.xml', nuclides_to_ignore: Optional[Iterable[str]] = None)[source]

Export material collection to an XML file.

Parameters
  • path (str) – Path to file to write. Defaults to ‘materials.xml’.

  • nuclides_to_ignore (list of str) – Nuclides to ignore when exporting to XML.

classmethod from_xml(path: Union[str, PathLike] = 'materials.xml') Material[source]

Generate materials collection from XML file

Parameters

path (str) – Path to materials XML file

Returns

Materials collection

Return type

openmc.Materials

classmethod from_xml_element(elem) Material[source]

Generate materials collection from XML file

Parameters

elem (lxml.etree._Element) – XML element

Returns

Materials collection

Return type

openmc.Materials

insert(index: int, material)[source]

Insert material before index

Parameters