It would be great to add an example of irradiation with a known/pre-calculated flux and deplete/activate/transmute a material.

This example would be similar to inventory codes (ALARA, ORIGEN and FISPACT) and would not perform neutron transport.

While OpenMC can do more advanced deplete/activate/transmute with neutron transport these inventory codes don't account for transport.

So if we want a simplified deplete/activate/transmute example to compare OpenMC with inventory codes we can "downgrade" the deplete/activate/transmute methods available in OpenMC to make this simplified inventory simulation.