OpenMC and Open-Source Neutronics
OpenMC is an open-source Monte-Carlo particle transport code widely used for fusion neutronics, with a scriptable interface and modern geometry support.
An open transport code
OpenMC is a community-developed, open-source Monte-Carlo neutron and photon transport code. Being open, it can be inspected, modified, and integrated into automated workflows without licensing barriers, which makes it attractive for reproducible fusion neutronics studies.
Geometry and materials
Models are built from either constructive solid geometry, defining regions by combining surfaces, or from tessellated CAD-derived meshes for complex engineering shapes. Materials are specified by nuclide composition and density, and the code reads continuous-energy cross-section data in standard evaluated-data formats.
Scriptable workflow
- A Python interface builds geometry, materials, sources, and tallies programmatically
- Tallies request fluxes, reaction rates, heating, and breeding on cells or meshes
- Depletion and activation follow material composition changes under irradiation
- Results are structured data suited to automated post-processing
Fusion-relevant capabilities
For fusion, OpenMC supports fusion source definitions, mesh tallies for spatial maps of heating and breeding, and coupling to activation calculations. Its scriptability makes parameter scans, such as varying blanket composition to study breeding, straightforward to automate and repeat.
Verification
Like any transport code, OpenMC results depend on the cross-section library used and on adequate statistics. Sound practice states the data library, reports statistical uncertainties, and benchmarks against reference problems or an independent code before trusting a new model.
An open, scriptable neutronics code fits naturally into a reproducible analysis suite, where a breeding or shielding result can be regenerated from versioned inputs by anyone with the same data library.