OpenMC is a Monte Carlo particle transport simulation code focused on neutron criticality calculations. It is capable of simulating 3D models based on constructive solid geometry with second-order surfaces. The particle interaction data is based on ACE format cross sections, also used in the MCNP and Serpent Monte Carlo codes. It simulates 3D models based on constructive solid geometry with second-order surfaces.
This example is a hybrid example job from the test suite using OpenMC and MPICH on Rescale.
|Simulation Code||OpenMC 0.6.0|
|Analysis Type||Monte Carlo particle transport simulation|
|Description||An example based on the test “test_source_energy_maxwell”.|
|Suggested Hardware||Titanium / 20 cores|
mpirun -f $HOME/mpd.hosts -n 2 -ppn 1 openmc -s 8 $(pwd)
|Estimated Run Time||1 minute|