IAEA-NDS / endf-parserpyLinks
Python package for reading, writing, verifying and translating ENDF-6 formatted files
☆22Updated 4 months ago
Alternatives and similar repositories for endf-parserpy
Users that are interested in endf-parserpy are comparing it to the libraries listed below
Sorting:
- Python ENDF Parser☆39Updated last month
- Tool for converting MCNP input files to OpenMC classes/XML☆52Updated last week
- NJOY for the 21st Century☆84Updated last year
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆61Updated 2 months ago
- Toolkit for reading and interacting with ENDF-6 formatted files☆41Updated last month
- List of open source projects related to OpenMC☆30Updated 6 months ago
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆48Updated this week
- Nuclear data processing with legacy NJOY☆122Updated this week
- A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB☆18Updated last week
- Thermal Scattering Library created with NJOY+NCrystal☆10Updated 3 years ago
- A Python package for parsing FISPACT-II output. https://deepwiki.com/fispact/pypact☆32Updated last month
- Automated V&V for fusion relevant CAD-based benchmarks☆23Updated last week
- JADE, a novel nuclear data libraries V&V tool☆34Updated last week
- A workshop covering a range of fusion relevant analysis and simulations with OpenMC, DAGMC, Paramak and other open source fusion neutroni…☆171Updated last week
- ☆87Updated last year
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆41Updated 4 years ago
- Stochastic Calculator Of Neutron transport Equation☆53Updated this week
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆83Updated 2 weeks ago
- API for parsing and post-processing many MC simulations related files☆17Updated this week
- Accelerated Discretized Geometry☆21Updated last week
- ENDF/B pre-processing codes (PREPRO)☆21Updated 2 years ago
- Computational inputs, reference outputs and experimental data for V&V systems☆15Updated 3 months ago
- Catalogs a plethora of reactor inputs for OpenMC☆35Updated 9 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…☆42Updated last year
- Tools to translate between different CSG geometry types☆40Updated 6 months ago
- Native plotting GUI for model design and verification☆64Updated 3 weeks ago
- Sampling nuclear data and uncertainty☆54Updated 2 months ago
- Convert OpenMC Geometry Components to CAD☆20Updated last year
- For Updating Data and Generating Evaluations (FUDGE): LLNL code for managing nuclear data☆32Updated last month
- MCNP6 Syntax highlighting and code snippets for VSCode. Written primarily for MCNP6.x input decks as a placeholder for the full LSP is im…☆14Updated last year