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]
  • materials (Iterable of openmc.Material) – Materials to add to the collection
  • cross_sections (str) – 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 to collection

Parameters:material (openmc.Material) – Material to append

Export material collection to an XML file.

Parameters:path (str) – Path to file to write. Defaults to ‘materials.xml’.
classmethod from_xml(path='materials.xml')[source]

Generate materials collection from XML file

Parameters:path (str, optional) – Path to materials XML file
Returns:Materials collection
Return type:openmc.Materials
insert(index, material)[source]

Insert material before index