github.com/openmc-dev/openmc ↗

About openmc-dev/openmc

OpenMC is a modern Monte Carlo particle transport code designed for simulating the movement and interaction of neutrons, photons, and other particles through complex materials and geometries. Built with contemporary computational methods, it uses constructive solid geometry for model definition and continuous-energy transport physics with cross-section data stored in HDF5 format. The project originated at MIT's Computational Reactor Physics Group and serves researchers and engineers in nuclear energy, fusion, and radiation transport applications.

The codebase provides a comprehensive simulation environment for nuclear engineering analysis, including capabilities for neutronics calculations, radiation shielding design, and reactor physics studies. OpenMC is implemented primarily in Python with high-performance computing capabilities, allowing users to model detailed geometric configurations and run particle transport simulations efficiently. The project maintains extensive documentation, offers installation through Conda, and provides Docker containers for easy deployment across different computing environments.

The OpenMC community is actively maintained through a discussion forum where users can seek help, report issues, and contribute to development. The code is open-source under the MIT license and has been the subject of peer-reviewed research, making it a widely recognized tool in the computational nuclear engineering field. It supports both development workflows and production research applications, with continuous integration testing and regular updates to the codebase.