mit-crpg / opendepleteLinks
A depletion framework for OpenMC
☆15Updated 7 years ago
Alternatives and similar repositories for opendeplete
Users that are interested in opendeplete are comparing it to the libraries listed below
Sorting:
- Collection of validation scripts, notebooks, results☆11Updated 2 years ago
- OpenMC Monte Carlo Code☆20Updated 9 months ago
- Collection of scripts for managing data for OpenMC☆35Updated 9 months ago
- ERSN-OpenMC is a Graphical User Interface for OpenMC Monte Carlo particle transport simulation code, originally developed by Jaafar EL Ba…☆13Updated 8 years ago
- MCNP(X) project builder using C++☆16Updated this week
- Some Monte Carlo tools☆52Updated 3 weeks ago
- Simulation of transmutation and decay in nuclear systems☆31Updated this week
- Native plotting GUI for model design and verification☆63Updated last week
- Collection of models for reactor physics benchmarks☆56Updated last month
- Plugins and command extensions for Coreform Cubit☆18Updated 4 months ago
- Sampling nuclear data and uncertainty☆52Updated 3 weeks ago
- Tool for Optimization and Group-structure Analysis of nuclear multi-group cross sections☆11Updated 3 years ago
- ENRICO: Exascale Nuclear Reactor Investigative COde☆65Updated 2 years ago
- Direct Accelerated Geometry Monte Carlo Toolkit☆116Updated 8 months ago
- Convert MCNP input files to a general CAD format☆38Updated 4 months ago
- Toolkit for reading and interacting with ENDF-6 formatted files☆40Updated this week
- MC/DC: Monte Carlo Dynamic Code☆45Updated this week
- code to target the conversion from a step-file to a h5m-geometry for neutronics☆34Updated 3 months ago
- Three-Dimensional Discrete Ordinates Neutron, Photon, Electron and Positron Transport Code☆10Updated 3 years ago
- Tool for converting MCNP input files to OpenMC classes/XML☆47Updated last month
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆41Updated 4 years ago
- Materials for the PHYSOR 2016 OpenMC workshop.☆10Updated 9 years ago
- open-source Sn software☆48Updated last week
- continuous energy monte carlo neutron transport in general geometries on GPUs☆45Updated 8 years ago
- Tools to translate between different CSG geometry types☆40Updated 5 months ago
- Python Interface and Utility Functions for MCNP6☆21Updated 8 years ago
- OpenMC wrapped as a Moose App☆16Updated 2 years ago
- NJOY for the 21st Century☆82Updated last year
- A utility code for generating material cards for MCNP☆21Updated 4 years ago
- A set of radiation transport mini-applications used for performance optimization on HPC systems.☆29Updated 7 years ago