fusion-energy / cad_to_dagmc
Convert CAD geometry (STP files) or Cadquery assemblies to DAGMC h5m files
☆30Updated 2 weeks ago
Alternatives and similar repositories for cad_to_dagmc:
Users that are interested in cad_to_dagmc are comparing it to the libraries listed below
- code to target the conversion from a step-file to a h5m-geometry for neutronics☆32Updated 2 weeks ago
- Meshing library for nuclear workflows☆49Updated 3 weeks ago
- Create parametric 3D fusion reactor CAD and neutronics models☆77Updated last week
- A tool for making reproducible materials and standardizing use across several neutronics codes☆24Updated 6 months ago
- Creates a plasma source as an openmc.source object from input parameters that describe the plasma☆33Updated 4 months ago
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆67Updated last week
- Parametric 3-D CAD modeling toolset for stellarator fusion devices☆30Updated last week
- Native plotting GUI for model design and verification☆57Updated 3 months ago
- Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.☆40Updated 10 months ago
- A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB☆16Updated this week
- A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.☆11Updated 11 months ago
- Workflow and Template Toolkit for Simulation (WATTS)☆30Updated last week
- a CAD to MC geometry conversion tool☆48Updated last year
- List of open source projects related to OpenMC☆25Updated 7 months ago
- Direct Accelerated Geometry Monte Carlo Toolkit☆112Updated 2 months ago
- Monte Carlo Particle Lists☆30Updated this week
- ☆21Updated 2 weeks ago
- Convert OpenMC Geometry Components to CAD☆17Updated 5 months ago
- Simulation of transmutation and decay in nuclear systems☆28Updated 3 months ago
- Coupled hydrogen/tritium transport and heat transfer modelling using FEniCS☆101Updated last week
- Tool for Optimization and Group-structure Analysis of nuclear multi-group cross sections☆10Updated 2 years ago
- Sampling nuclear data and uncertainty☆51Updated last month
- Example Jupyter notebooks for OpenMC☆37Updated last month
- Collection of models for reactor physics benchmarks☆52Updated 6 months ago
- open-source Sn software☆29Updated this week
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆41Updated this week
- Format description for ACE nuclear data files☆21Updated 2 years ago
- Tools to translate between different CSG geometry types☆37Updated last year
- OpenMC Monte Carlo Code☆17Updated 2 months ago
- API for parsing and post-processing many MC simulations related files☆15Updated this week