njoy / NJOY21Links
NJOY for the 21st Century
☆79Updated last year
Alternatives and similar repositories for NJOY21
Users that are interested in NJOY21 are comparing it to the libraries listed below
Sorting:
- Nuclear data processing with legacy NJOY☆111Updated this week
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆59Updated 3 months ago
- Tool for converting MCNP input files to OpenMC classes/XML☆43Updated last month
- Direct Accelerated Geometry Monte Carlo Toolkit☆112Updated 5 months ago
- Python ENDF Parser☆34Updated 11 months ago
- Tools to translate between different CSG geometry types☆39Updated last month
- ☆82Updated last year
- List of open source projects related to OpenMC☆26Updated last month
- Format description for ACE nuclear data files☆21Updated 3 years ago
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆47Updated this week
- A two-cycle full-core PWR depletion benchmark based on a commercial nuclear power plant☆38Updated last week
- Stochastic Calculator Of Neutron transport Equation☆45Updated this week
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆71Updated 3 weeks ago
- Native plotting GUI for model design and verification☆59Updated 5 months ago
- API for parsing and post-processing many MC simulations related files☆17Updated this week
- Toolkit for reading and interacting with ENDF-6 formatted files☆40Updated last week
- Automated V&V for fusion relevant CAD-based benchmarks☆23Updated last week
- Thermal Scattering Library created with NJOY+NCrystal☆9Updated 3 years ago
- Sampling nuclear data and uncertainty☆52Updated 3 weeks ago
- A method of characteristics code for nuclear reactor physics calculations.☆165Updated last year
- ENRICO: Exascale Nuclear Reactor Investigative COde☆65Updated last year
- open-source Sn software☆36Updated this week
- Convert OpenMC Geometry Components to CAD☆18Updated 7 months ago
- JADE, a novel nuclear data libraries V&V tool☆31Updated last month
- 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
- ☆31Updated 3 weeks ago
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 3 years ago
- Workflow and Template Toolkit for Simulation (WATTS)☆32Updated 3 weeks ago
- OpenMC wrapped as a Moose App☆16Updated 2 years ago
- Ducted Assembly Steady State Heat Transfer Software (DASSH)☆18Updated last week