repositony / vscode_mcnp
MCNP6 Syntax highlighting and code snippets for VSCode. Written primarily for MCNP6.x input decks.
☆9Updated 6 months ago
Alternatives and similar repositories for vscode_mcnp:
Users that are interested in vscode_mcnp are comparing it to the libraries listed below
- Thermal Scattering Library created with NJOY+NCrystal☆9Updated 2 years ago
- A Python package for parsing FISPACT-II output☆26Updated 10 months ago
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 3 years ago
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆59Updated last month
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆34Updated this week
- Python ENDF Parser☆27Updated 5 months ago
- ☆18Updated 5 months ago
- ☆71Updated 7 months ago
- A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB☆13Updated 7 months ago
- a companion for writing MCNP input decks☆11Updated 3 years ago
- List of open source projects related to OpenMC☆22Updated 4 months ago
- Tools to translate between different CSG geometry types☆36Updated last year
- A tool for making parametric material cards for use in neutronics codes. Original developed for the Paramak☆19Updated 3 years ago
- Tool for converting MCNP input files to OpenMC classes/XML☆37Updated 3 weeks ago
- JADE, a novel nuclear data libraries V&V tool☆25Updated 3 weeks ago
- OpenMC Monte Carlo Code☆17Updated 4 months ago
- An OpenMC-based repo for studying proliferation issues in ARC-class fusion reactors.☆11Updated 4 months ago
- Collection of tools for efficiency improvements in developing a CAD model for neutronics analysis☆13Updated 3 years ago
- Native plotting GUI for model design and verification☆54Updated last month
- Contains slides and example problems for OpenMC workshops and presentations☆23Updated 2 years ago
- Toolkit for reading and interacting with ENDF-6 formatted files☆35Updated 2 months ago
- A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.☆11Updated 8 months ago
- A two-cycle full-core PWR depletion benchmark based on a commercial nuclear power plant☆33Updated 2 weeks ago
- ENDF/B pre-processing codes (PREPRO)☆16Updated last year
- XSPlot - Neutron cross section plotter for isotopes☆14Updated last year
- Sampling nuclear data and uncertainty☆49Updated last week
- ☆29Updated last year
- Catalogs a plethora of reactor inputs for OpenMC☆19Updated 3 years ago
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆53Updated 7 months ago
- Tool for Optimization and Group-structure Analysis of nuclear multi-group cross sections☆10Updated 2 years ago