openmc-dev / dataLinks
Collection of scripts for managing data for OpenMC
☆33Updated 6 months ago
Alternatives and similar repositories for data
Users that are interested in data are comparing it to the libraries listed below
Sorting:
- Collection of validation scripts, notebooks, results☆11Updated last year
- Sampling nuclear data and uncertainty☆52Updated last month
- Native plotting GUI for model design and verification☆60Updated 6 months ago
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 3 years ago
- ENRICO: Exascale Nuclear Reactor Investigative COde☆65Updated 2 years ago
- Python ENDF Parser☆34Updated last year
- Direct Accelerated Geometry Monte Carlo Toolkit☆112Updated 5 months ago
- Stochastic Calculator Of Neutron transport Equation☆47Updated last week
- OpenMC wrapped as a Moose App☆16Updated 2 years ago
- List of open source projects related to OpenMC☆26Updated 2 months ago
- Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.☆42Updated last year
- Collection of models for reactor physics benchmarks☆54Updated 10 months ago
- OpenMC Monte Carlo Code☆18Updated 6 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…☆41Updated 7 months ago
- MC/DC: Monte Carlo Dynamic Code☆43Updated 3 months ago
- JADE, a novel nuclear data libraries V&V tool☆31Updated 2 months ago
- Automated V&V for fusion relevant CAD-based benchmarks☆23Updated 3 weeks ago
- Workflow and Template Toolkit for Simulation (WATTS)☆32Updated last month
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆61Updated 3 months ago
- open-source Sn software☆37Updated last week
- Tool for converting MCNP input files to OpenMC classes/XML☆44Updated 2 months ago
- ENDF/B pre-processing codes (PREPRO)☆19Updated last year
- A deterministic finite-element-based solver for the neutron transport equation designed for the evaluation of acceleration methods.☆27Updated 4 years ago
- For Updating Data and Generating Evaluations (FUDGE): LLNL code for managing nuclear data☆30Updated last week
- code to target the conversion from a step-file to a h5m-geometry for neutronics☆33Updated last month
- Tools to translate between different CSG geometry types☆39Updated 2 months ago
- MontePy is the most user friendly Python library (API) to read, edit, and write MCNP input files.☆48Updated this week
- NJOY for the 21st Century☆79Updated last year
- Tool for Optimization and Group-structure Analysis of nuclear multi-group cross sections☆11Updated 2 years ago
- Format description for ACE nuclear data files☆22Updated 3 years ago