tarekbardouni / ERSN-OpenMCLinks
ERSN-OpenMC is a Graphical User Interface for OpenMC Monte Carlo particle transport simulation code, originally developed by Jaafar EL Bakkali & Tarek EL Bardouni, members of Radiation and Nuclear Systems Group ERSN at University Abdelmalek Essaady in Tetouan (Morocco).
☆31Updated last year
Alternatives and similar repositories for ERSN-OpenMC
Users that are interested in ERSN-OpenMC are comparing it to the libraries listed below
Sorting:
- Collection of scripts for managing data for OpenMC☆38Updated last month
- Repository for Moltres, a code for simulating Molten Salt Reactors☆77Updated this week
- Simulation of transmutation and decay in nuclear systems☆35Updated this week
- Materials for the PHYSOR 2016 OpenMC workshop.☆10Updated 9 years 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
- ENRICO: Exascale Nuclear Reactor Investigative COde☆66Updated 2 years ago
- A method of characteristics code for nuclear reactor physics calculations.☆178Updated last year
- Native plotting GUI for model design and verification☆64Updated last week
- High-Fidelity Multiphysics☆123Updated last week
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆61Updated 4 months ago
- Sampling nuclear data and uncertainty☆55Updated 3 months ago
- OpenMC wrapped as a Moose App☆16Updated 2 years ago
- Online reprocessing for molten salt reactors☆22Updated 9 months ago
- Direct Accelerated Geometry Monte Carlo Toolkit☆120Updated 11 months ago
- The Cyclus Nuclear Fuel Cycle Simulator☆86Updated 2 weeks ago
- Tools to translate between different CSG geometry types☆40Updated 8 months ago
- Nuclear Reactor Theory☆20Updated 8 years ago
- continuous energy monte carlo neutron transport in general geometries on GPUs☆46Updated 8 years ago
- OpenMC Monte Carlo Code☆24Updated last year
- THOR is a radiation transport code for unstructured meshes.☆16Updated last year
- 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
- Collection of validation scripts, notebooks, results☆11Updated last month
- Stochastic Calculator Of Neutron transport Equation☆55Updated last week
- Collection of models for reactor physics benchmarks☆58Updated 4 months ago
- Workflow and Template Toolkit for Simulation (WATTS)☆33Updated 7 months ago
- Plugins and command extensions for Coreform Cubit☆18Updated 7 months ago
- Nuclear data processing with legacy NJOY☆123Updated last week
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆41Updated 4 years ago
- A workshop covering a range of fusion relevant analysis and simulations with OpenMC, DAGMC, Paramak and other open source fusion neutroni…☆175Updated this week
- code to target the conversion from a step-file to a h5m-geometry for neutronics☆38Updated 6 months ago