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openmclisted

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.
itaybnv/openmc-claude-skills · ★ 0 · AI & Automation · score 78
Install: claude install-skill itaybnv/openmc-claude-skills
# OpenMC Knowledge Base **Coverage:** Materials · CSG Geometry · Composite Geometry · Lattice Geometry · Settings · Settings Advanced · Tallies · Statepoint · Model API · **Model Mutation** · **Kinetics** · **Sensitivity** · Gotchas · Depletion · MGXS · Nuclear Data · **Data Functions** · openmc.lib · Patterns · Overview · Parallel · Plots · Mesh · Filters · Sources · Volume & Tracks · **CMFD** · **DAGMC** · **Random Ray** **Not yet covered** (Phase 6): README, community docs, public release ## Task Routing ## Entry Points **Broad / orientation / "how does OpenMC work" / "getting started" / unclear scope:** Read `docs/overview.md` **Complete example / end-to-end script / "full simulation" / "copy-paste runnable" / workflow:** Read `docs/patterns.md` --- Load the docs relevant to the task at hand. For narrow questions, load one file (e.g., only tallies.md for a filter question). For cross-cutting tasks, load the files that span the question (e.g., tallies.md + statepoint.md for construction-through-readback). Never load all docs indiscriminately — choose based on what the task actually requires. --- ## Cluster A — Build a Model **Materials / nuclides / enrichment / density / get_activity / get_decay_heat / decay heat / waste classification / waste_disposal_rating / NCrystal / from_ncrystal / ncrystal_cfg / Macroscopic / material Mixture / mix_materials / Nuclide class / Element class / activation material:** Read `docs/materials.md` **CSG geometry — surfaces, cells