repositony / vscode_mcnpLinks
MCNP6 Syntax highlighting and code snippets for VSCode. Written primarily for MCNP6.x input decks.
☆12Updated 10 months ago
Alternatives and similar repositories for vscode_mcnp
Users that are interested in vscode_mcnp are comparing it to the libraries listed below
Sorting:
- 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
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆46Updated this week
- Tool for converting MCNP input files to OpenMC classes/XML☆42Updated 2 weeks ago
- Computational inputs, reference outputs and experimental data for V&V systems☆13Updated last month
- ☆79Updated last year
- Python ENDF Parser☆33Updated 10 months ago
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 3 years ago
- Automated V&V for fusion relevant CAD-based benchmarks☆23Updated this week
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆70Updated 3 weeks ago
- A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB☆16Updated 3 weeks ago
- JADE, a novel nuclear data libraries V&V tool☆29Updated this week
- a companion for writing MCNP input decks☆12Updated 4 years ago
- Python package for reading, writing, verifying and translating ENDF-6 formatted files☆16Updated this week
- List of open source projects related to OpenMC☆26Updated last week
- Convert OpenMC Geometry Components to CAD☆18Updated 6 months ago
- Tools to translate between different CSG geometry types☆37Updated last week
- Stochastic Calculator Of Neutron transport Equation☆43Updated 3 weeks ago
- A Python package for parsing FISPACT-II output☆28Updated last week
- Collection of tools for efficiency improvements in developing a CAD model for neutronics analysis☆14Updated 4 years ago
- Accelerated Discretized Geometry☆17Updated this week
- OpenMC Monte Carlo Code☆18Updated 4 months ago
- Native plotting GUI for model design and verification☆58Updated 4 months ago
- Contains slides and example problems for OpenMC workshops and presentations☆23Updated 2 years ago
- A two-cycle full-core PWR depletion benchmark based on a commercial nuclear power plant☆34Updated 2 months ago
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆57Updated last month
- Toolkit for reading and interacting with ENDF-6 formatted files☆41Updated last week
- Catalogs a plethora of reactor inputs for OpenMC☆24Updated 3 months ago
- A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.☆11Updated last month
- Sampling nuclear data and uncertainty☆51Updated last month