repositony / vscode_mcnpLinks
MCNP6 Syntax highlighting and code snippets for VSCode. Written primarily for MCNP6.x input decks as a placeholder for the full LSP is implemented.
☆13Updated last year
Alternatives and similar repositories for vscode_mcnp
Users that are interested in vscode_mcnp are comparing it to the libraries listed below
Sorting:
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆48Updated this week
- Thermal Scattering Library created with NJOY+NCrystal☆9Updated 3 years ago
- API for parsing and post-processing many MC simulations related files☆17Updated last week
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆72Updated last month
- Automated V&V for fusion relevant CAD-based benchmarks☆23Updated 3 weeks ago
- A Python package for parsing FISPACT-II output. https://deepwiki.com/fispact/pypact☆29Updated last week
- Computational inputs, reference outputs and experimental data for V&V systems☆13Updated 4 months ago
- Python ENDF Parser☆34Updated last year
- List of open source projects related to OpenMC☆26Updated 2 months ago
- Stochastic Calculator Of Neutron transport Equation☆47Updated last week
- Tool for converting MCNP input files to OpenMC classes/XML☆44Updated 2 months ago
- ☆83Updated last year
- A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB☆16Updated 3 weeks ago
- Python package for reading, writing, verifying and translating ENDF-6 formatted files☆19Updated 2 weeks ago
- JADE, a novel nuclear data libraries V&V tool☆31Updated 2 months ago
- Native plotting GUI for model design and verification☆60Updated 6 months ago
- Convert OpenMC Geometry Components to CAD☆18Updated 8 months ago
- Accelerated Discretized Geometry☆17Updated 3 weeks ago
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 3 years ago
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆61Updated 3 months ago
- A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.☆12Updated 3 months ago
- Tools to translate between different CSG geometry types☆39Updated 2 months ago
- NJOY for the 21st Century☆79Updated last year
- ☆31Updated 2 weeks ago
- Tool for modeling distributional particle sources for Monte Carlo simulations, with Kernel Density Estimation.☆22Updated 2 months ago
- a companion for writing MCNP input decks☆12Updated 4 years ago
- OpenMC Monte Carlo Code☆18Updated 6 months ago
- ONIX is an open-source depletion software for nuclear reactor simulations and nuclear archaeology. It is written in Python 3 and offers c…☆41Updated 7 months ago
- A workshop covering a range of fusion relevant analysis and simulations with OpenMC, DAGMC, Paramak and other open source fusion neutroni…☆156Updated this week
- open-source Sn software☆37Updated this week