openmc-dev / openmc_mcnp_adapterLinks
Tool for converting MCNP input files to OpenMC classes/XML
☆52Updated last week
Alternatives and similar repositories for openmc_mcnp_adapter
Users that are interested in openmc_mcnp_adapter are comparing it to the libraries listed below
Sorting:
- List of open source projects related to OpenMC☆30Updated 6 months ago
- Python ENDF Parser☆39Updated last month
- API for parsing and post-processing many MC simulations related files☆17Updated last month
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆41Updated 4 years ago
- Automated V&V for fusion relevant CAD-based benchmarks☆23Updated this week
- Native plotting GUI for model design and verification☆64Updated 2 weeks ago
- Toolkit for reading and interacting with ENDF-6 formatted files☆40Updated 3 weeks ago
- Accelerated Discretized Geometry☆19Updated last week
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆48Updated this week
- ENDF/B pre-processing codes (PREPRO)☆21Updated 2 years ago
- A convenience interface for examining and reassigning metadata in DAGMC models with PyMOAB☆17Updated this week
- Convert OpenMC Geometry Components to CAD☆20Updated last year
- Computational inputs, reference outputs and experimental data for V&V systems☆15Updated 2 months ago
- ☆87Updated last year
- JADE, a novel nuclear data libraries V&V tool☆34Updated this week
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆81Updated last week
- Tools to translate between different CSG geometry types☆40Updated 6 months ago
- A two-cycle full-core PWR depletion benchmark based on a commercial nuclear power plant☆40Updated 2 months ago
- Tool for Optimization and Group-structure Analysis of nuclear multi-group cross sections☆11Updated 3 years ago
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆61Updated 2 months ago
- Sampling nuclear data and uncertainty☆53Updated 2 months ago
- Stochastic Calculator Of Neutron transport Equation☆53Updated last week
- Repository for running SINBAD benchmarks with OpenMC and UQ☆10Updated 3 years ago
- OpenMC Monte Carlo Code☆23Updated 10 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
- NJOY for the 21st Century☆84Updated last year
- A Python package for plotting OpenMC regular mesh tally results with underlying geometry from neutronics simulations.☆14Updated 7 months ago
- For Updating Data and Generating Evaluations (FUDGE): LLNL code for managing nuclear data☆32Updated last month
- ☆33Updated 4 months ago
- Open-source Sn software☆52Updated this week