Supported Neutron Transport Codes

Exporting a neutron source

tokamak_neutron_source is designed to be able to create neutron sources for use in any neutron transport code. At present, interfaces for two codes are supported:

OpenMC

tokamak_neutron_source is able to export a neutron source in native OpenMC format, via the specification of a list of openmc.IndependentSource. When creating the source, the default OpenMC units are used (eV and cm). Exercise caution if you have configured OpenMC to operate with a different set of units. This functionality requires you to have OpenMC installed in the same environment.

If multiple neutronic fusion reactions are specified, the energy distributions are combined into a single source at each point.

MCNP-6

TODO: Add documentation when this functionality exists.