byADIT Free software for engineers and students
EngiFree
OpenMC — The Complete Guide: Free Monte Carlo Neutron Transport for Nuclear Engineering

OpenMC — The Complete Guide: Free Monte Carlo Neutron Transport for Nuclear Engineering

OpenMC is an open-source Monte Carlo code for neutron and photon transport, started at MIT and developed today also at Argonne National Laboratory. It is used in nuclear engineering for criticality, neutron flux, fuel depletion and radiation shielding — a field where the traditional codes were until recently restricted by licensing and export control.

What you get

How to use it

The whole model — materials, geometry, settings and tallies — is written in Python, so parametric studies and automation are relatively simple. You need to download a standard cross-section library in the format the code uses. It runs on Linux and macOS; on Windows use WSL or Docker.

Related tools

To view flux fields and 3D results — ParaView. For coupled multiphysics, for example with thermal hydraulics — MOOSE. For data analysis and plots — JupyterLab.

The bottom line

OpenMC is the open, modern alternative to nuclear Monte Carlo codes — completely free, with an active community and good documentation.

Further reading