# openmc.SourceParticle¶

class openmc.SourceParticle(r=(0.0, 0.0, 0.0), u=(0.0, 0.0, 1.0), E=1000000.0, wgt=1.0, delayed_group=0, surf_id=0, particle=<ParticleType.NEUTRON: 0>)[source]

Source particle

This class can be used to create source particles that can be written to a file and used by OpenMC

Parameters: r (iterable of float) – Position of particle in Cartesian coordinates u (iterable of float) – Directional cosines E (float) – Energy of particle in [eV] wgt (float) – Weight of the particle delayed_group (int) – Delayed group particle was created in (neutrons only) surf_id (int) – Surface ID where particle is at, if any. particle (ParticleType) – Type of the particle
to_tuple()[source]

Return source particle attributes as a tuple

Returns: Source particle attributes tuple