tarekbardouni / ERSN-OpenMC
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).
☆29Updated 7 months ago
Alternatives and similar repositories for ERSN-OpenMC:
Users that are interested in ERSN-OpenMC are comparing it to the libraries listed below
- Collection of scripts for managing data for OpenMC☆29Updated 2 weeks ago
- ERSN-OpenMC is a Graphical User Interface for OpenMC Monte Carlo particle transport simulation code, originally developed by Jaafar EL Ba…☆13Updated 7 years ago
- Repository for Moltres, a code for simulating Molten Salt Reactors☆67Updated this week
- ENRICO: Exascale Nuclear Reactor Investigative COde☆64Updated last year
- Native plotting GUI for model design and verification☆55Updated last week
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆53Updated 7 months ago
- A method of characteristics-based nuclear reactor physics simulator☆14Updated 2 years ago
- Collection of models for reactor physics benchmarks☆50Updated 3 months ago
- Simulation of transmutation and decay in nuclear systems☆28Updated 2 weeks ago
- OpenMC wrapped as a Moose App☆16Updated last year
- Online reprocessing for molten salt reactors☆20Updated last week
- code to target the conversion from a step-file to a h5m-geometry for neutronics☆32Updated last week
- Direct Accelerated Geometry Monte Carlo Toolkit☆106Updated last week
- Nuclear Reactor Theory☆19Updated 7 years ago
- A set of radiation transport mini-applications used for performance optimization on HPC systems.☆28Updated 6 years ago
- A selection of resources for learning openmc with particular focus on simulations relevant for fusion energy☆39Updated 3 years ago
- Sampling nuclear data and uncertainty☆49Updated 3 weeks ago
- Simulate PWR,CANDU and VVER reactor (Assembly Model) using OpenMC https://openmc.readthedocs.io/en/stable/☆11Updated 5 years ago
- Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.☆39Updated 7 months ago
- The Cyclus Nuclear Fuel Cycle Simulator☆81Updated last month
- Stochastic Calculator Of Neutron transport Equation☆39Updated this week
- High-Fidelity Multiphysics☆99Updated this week
- Nuclear data processing with legacy NJOY☆101Updated this week
- A tool to convert CAD to CSG & CSG to CAD for Monte Carlo transport codes☆59Updated last month
- The National Reactor Innovation Center's (NRIC) Virtual Test Bed Repository☆51Updated this week
- Plugins and command extensions for Coreform Cubit☆18Updated 7 months ago
- Creates a plasma source as an openmc.source object from input parameters that describe the plasma☆31Updated last month
- continuous energy monte carlo neutron transport in general geometries on GPUs☆43Updated 7 years ago
- Contains slides and example problems for OpenMC workshops and presentations☆23Updated 2 years ago
- Tools to translate between different CSG geometry types☆36Updated last year