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).
☆30Updated 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:
- Repository for Moltres, a code for simulating Molten Salt Reactors☆70Updated last week
- Collection of scripts for managing data for OpenMC☆33Updated 6 months ago
- Simulation of transmutation and decay in nuclear systems☆29Updated 6 months ago
- ENRICO: Exascale Nuclear Reactor Investigative COde☆65Updated 2 years ago
- Online reprocessing for molten salt reactors☆20Updated 3 months 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
- A set of radiation transport mini-applications used for performance optimization on HPC systems.☆29Updated 6 years ago
- Collection of validation scripts, notebooks, results☆11Updated last year
- Nuclear Reactor Theory☆19Updated 7 years ago
- Native plotting GUI for model design and verification☆59Updated 6 months ago
- Sampling nuclear data and uncertainty☆52Updated last month
- The Cyclus Nuclear Fuel Cycle Simulator☆82Updated this week
- Monte Carlo Particle Lists☆32Updated 2 weeks ago
- continuous energy monte carlo neutron transport in general geometries on GPUs☆43Updated 8 years ago
- An object-oriented component library supporting radiation transport applications.☆56Updated 3 months ago
- Direct Accelerated Geometry Monte Carlo Toolkit☆112Updated 5 months ago
- A method of characteristics code for nuclear reactor physics calculations.☆166Updated last year
- Aurora is A Unified Resource for OpenMC (fusion) Reactor Applications.☆41Updated last year
- OpenMC Monte Carlo Code☆18Updated 5 months ago
- OpenMC wrapped as a Moose App☆16Updated 2 years ago
- A deterministic finite-element-based solver for the neutron transport equation designed for the evaluation of acceleration methods.☆27Updated 4 years ago
- A suite of parsers designed to make interacting with SERPENT output files simple and flawless☆59Updated 3 months ago
- A library for multiphysics solution transfer. ARCHIVED☆50Updated last year
- Stochastic Calculator Of Neutron transport Equation☆47Updated this week
- MC/DC: Monte Carlo Dynamic Code☆43Updated 3 months ago
- Workflow and Template Toolkit for Simulation (WATTS)☆32Updated 3 weeks ago
- Format description for ACE nuclear data files☆21Updated 3 years ago
- High-Fidelity Multiphysics☆114Updated this week
- detailed cad model of the msre☆38Updated 5 months ago
- THOR is a radiation transport code for unstructured meshes.☆14Updated 10 months ago