openmc-dev/openmc

Health v2 · maintenance only

Flags: no_license

Adoption not part of the score

1093 stars · 676 forks observed · 2026-09-03

What it is AI-extracted, prompt v1, taxonomy v1, 2026-08-30, confidence not recorded

OpenMC is a Monte Carlo particle transport code for simulating neutron and photon transport in constructive solid geometry using continuous-energy HDF5 cross sections. It combines a C++ simulation kernel with a Python API for building models, running simulations, and analyzing results.

Use cases

• simulate neutron transport in a nuclear reactor core
• run Monte Carlo radiation transport calculations
• compute k-eigenvalues and reaction rates for reactor physics
• model shielding and photon transport problems
• perform neutronics analysis for fusion blankets
• build parametric reactor models with the Python API
• use continuous-energy nuclear data in HDF5 format for simulations

When to choose

• you need an open-source Monte Carlo code for neutron or photon transport
• you want a scriptable Python API for building and running particle transport models
• you need continuous-energy transport with constructive solid geometry
• you are doing reactor physics, fusion neutronics, or shielding research
• you want a code that scales to high-performance computing clusters

When to avoid

• you need deterministic transport methods rather than Monte Carlo
• you need coupled multiphysics simulation out of the box
• you want a GUI-driven tool with no programming
• your application is unrelated to radiation or particle transport

Facets

library · maturity active

simulation math sdk developer-tools simulation windows python cpp cross-platform monte-carlo particle-transport neutronics radiation-transport nuclear-engineering reactor-physics hdf5 constructive-solid-geometry hpc physics computational-physics high-performance-computing linux macos

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