OpenMC Example

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.

Import Job Setup
Get Job Results
Simulation CodeOpenMC 0.6.0
Analysis TypeMonte Carlo particle transport simulation
DescriptionAn example based on the test “test_source_energy_maxwell”.
Suggested HardwareTitanium / 20 cores
Command 
mpirun -f $HOME/mpd.hosts -n 2 -ppn 1 openmc -s 8 $(pwd)
Estimated Run Time1 minute