mc · GPU neutron transport
mc runs Monte Carlo neutron transport in the browser using WebGPU. Build 3D material regions, configure a neutron source, run fixed-source or eigenvalue calculations, and inspect spatial flux and reaction rates. All calculations run on the local device.
Start with Pin cell for a two-group eigenvalue model, Fixed source for leakage and capture, or Continuous energy for evaluated H-1 transport. Each preset states its assumptions. Run, pause, and reset controls operate on real GPU transport; plots appear only after histories are scored.
This is a research solver. The included multigroup materials are model cross sections. The continuous-energy example uses ENDF/B-VIII.0 H-1 at 0 K and contains no fission. General reactor data, temperature treatment, and safety qualification are outside this release.
- Physics and limitations
- GPU architecture
- Nuclear data and preprocessing
- Validation and references
- Measured performance
- Local agent interface
- Deployment and public checks
- Configuration presets
- Open the application
Configurations save in browser storage or export as JSON. Importing and exporting files requires a user action. No calculation inputs or results are submitted to a backend. Runtime downloads contain only public application assets and bundled nuclear data. Clear browser storage to remove saved configurations.
Keyboard: use Tab to reach controls and geometry regions. On the canvas, arrow keys orbit, +/− zoom, and Home restores the camera. Pointer drag orbits; Shift-drag pans; the wheel zooms. Volume and Z-slice views consume the scored GPU field. Choose a group, reaction channel, scale, and color map in the inspector.
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