Particle transport · Nuclear data · Activation & inventory
FusionFissionGeneral nuclear

OpenMC.

Open Monte Carlo Particle Transport Code

OpenMC is an open-source Monte Carlo code that tracks neutrons and photons through models built from constructive solid geometry or CAD (through DAGMC), using continuous-energy or multigroup nuclear data, in fixed-source, eigenvalue and subcritical modes with MPI and OpenMP parallelism. Its Python API drives model building, depletion and activation workflows (including R2S and D1S shutdown dose rate), nuclear-data processing and tally analysis.

Runs in the HubMorton reproducedSources checked 2026-10-10
Engineering questions it helps answer
  • What neutron flux, nuclear heating, tritium breeding ratio or damage-driving reaction rates does a blanket, divertor or shield see for a given fusion source?
  • What is the shutdown dose rate around an activated component at a chosen cooling time?
  • How much do shielding or geometry changes alter dose and flux at a sensitive location, using weight windows to converge deep-penetration problems?
  • What is k-effective, and how does fuel composition evolve with burnup, for a fission lattice or core?
Physics scope
  • Continuous-energy and multigroup neutron transport
  • Photon transport, coupled neutron-photon runs, pulse-height tallies
  • Fixed-source, k-eigenvalue and subcritical-multiplication modes
  • CSG geometry with lattices; CAD geometry via DAGMC
  • Depletion, transmutation, activation and shutdown dose rate (R2S, D1S)
  • Cell, regular/rectilinear/cylindrical/spherical and unstructured mesh tallies
  • Weight-window variance reduction; random-ray deterministic solver
  • Nuclear-data processing in Python (ENDF, ACE, HDF5)

Limitations

Transports neutrons and photons only; electrons and positrons are not tracked. It has no thermal-hydraulics, structural or electromagnetic physics of its own, so multiphysics coupling is done by external drivers through the Python and C/C++ APIs. Nuclear data must be supplied in OpenMC's HDF5 format, and CAD geometry needs a DAGMC-enabled build. Monte Carlo results carry statistical uncertainty; deep-penetration shielding needs weight windows to converge.

Fusion relevance

A general neutron and photon transport code with a fission-physics origin, now widely used for fusion neutronics because it carries what fusion work needs: CAD geometry through DAGMC, a documented decay-source chapter for R2S and D1S shutdown-dose workflows, weight-window variance reduction, and since v0.16.0 a native parametric tokamak source and a fusion neutron spectrum helper. Fission use is just as well supported.

Upstream examples

Published by the project itself. These are the project's own tutorials, not Morton Labs runs.

Reference workflows on this site
Upstream example

Cardinal LWR pincell: OpenMC transport coupled to MOOSE heat conduction

For one pressurised-water-reactor fuel pin, what converged fission heat-source distribution and solid temperature field result when OpenMC eigenvalue transport with temperature feedback is iterated against a MOOSE heat-conduction solve?

Upstream exampleCardinalOpenMCMOOSE
Upstream example

Tritium breeding ratio of a lithium-lead spherical blanket with a 14 MeV point source

How much tritium does a simple spherical lithium-lead breeder blanket produce per 14 MeV source neutron, and which lithium isotope produces most of it?

Morton reproducedOpenMC
People
Editorial spotlight

Paul Romano

Computational scientist at Argonne National Laboratory; project lead of OpenMC; curator of Awesome Nuclear

Computational Scientist · Argonne National Laboratory

Awesome NuclearCardinalendf-pythonOpenMC
Paul RomanoOriginal authorHistorical · 2011–

OpenMC was begun at the MIT Computational Reactor Physics Group in 2011; he is first author of the recommended 2015 citation and the first listed software author in the repository's CITATION.cff.

Evidence: docs.openmc.org · checked 2026-10-10
Paul RomanoProject leadCurrent

The OpenMC developer guide names him Project Lead and a Technical Committee member. He also published every OpenMC release from v0.15.0 (2024) to v0.16.0 (August 2026).

Evidence: docs.openmc.org · checked 2026-10-10

Roles name a specific relationship with evidence. Authorship of a paper is not maintainership, and a listed person is not a Morton Labs partner or endorser.

Source collections

This entry was written by Morton Labs from public sources and last checked on 2026-10-10. Spotted an error, or want something changed or removed? Email zach@mortonlabs.ai with the page address.

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