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.


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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
Command
mpirun -f $HOME/mpd.hosts -n 2 -ppn 1 openmc -s 8 $(pwd)
Estimated Run Time 1 minute