Openmc

Use when writing OpenMC Python code for neutron/photon transport simulations, geometry construction (CSG surfaces, cells, universes, lattices), material definition (nuclides, elements, enrichment, density), settings configuration (eigenvalue, fixed-source, batches, particles, source distributions), or when debugging OpenMC Python errors. Covers fission reactor design and research neutronics workflows.

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itaybnv/openmc-claude-skills/tree/main/skills/openmc commit 685d91bd9a

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npx skillmds@latest add itaybnv/openmc