OpenMC
MIT / Argonne National Laboratory
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openmc.org
A Monte Carlo code for neutron and photon transport, used for criticality calculations, neutron flux, fuel depletion and radiation shielding in nuclear engineering. It is defined from Python, which makes parametric studies straightforward, and it uses standard cross-section libraries. This is the open alternative to codes that until recently were restricted by licensing and export control.
דירוג וביקורות
| Licence type | Completely Free, Open Source |
|---|---|
| Licence terms | Open-source licence. Free for any purpose including commercial work, with no time limit, no registration and no licence key. |
| Platforms | Linux, macOS |
| Engineering field | Mechanical Engineering |
| Tool type | Simulation & FEA |
| Vendor site | openmc.org |
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Last verified: 2026-09-02
