ERSN-OpenMC, a Java-based GUI for OpenMC Monte Carlo code
Journal of Radiation Research and Applied Sciences, 2016 OpenMC is a new Monte Carlo transport particle simulation code focused on solving two types of neutronic problems mainly the k-eigenvalue criticality fission source problems and external fixed fission source problems.
Jaafar El Bakkali, Seyedmostafa Safavi
exaly +5 more sources
Nuclear data processing capabilities in OpenMC [PDF]
EPJ Web of Conferences, 2017 This work describes newly developed features of the OpenMC code for nuclear data processing. OpenMC, in addition to being a transport code, includes a rich, extensible Python API that enables programmatic pre- and post-processing.
Romano Paul, Harper Sterling
doaj +4 more sources
Inter-Code Comparison of Computational VERA Depletion Benchmark Using OpenMC, OpenMC-ONIX and DRAGON
Atom Indonesia, 2022 This research focuses on the comparative analysis of the PWR fuel assembly based on VERA depletion benchmark problems using community-developed open source Monte Carlo code OpenMC, python based burnup code system ONIX (a coupling interface for Monte ...
A. Islam, T. A. Rahim, A. S. Mollah
doaj +4 more sources
OpenMC(TD): Current status and next development stages [PDF]
EPJ Web of ConferencesThis work presents the current state and the next stages in the development of the modified code OpenMC(TD), a modified version of the Monte Carlo code OpenMC which includes the time dependence related to the emission of β-delayed neutrons, but instead ...
Romero-Barrientos Jaime +3 more
doaj +3 more sources
VERIFICATION OF THE OpenMC HOMOGENIZED MYRRHA-1.6 CORE MODEL [PDF]
EPJ Web of Conferences, 2021 The OpenMC code is being employed both as a multi-group nodal macroscopic cross-section generator and a reference multi-group Monte Carlo (MGMC) solution.
Hernandez-Solis Augusto +4 more
doaj +3 more sources
Neutronics and burnup analysis of VVER-1000 LEU and MOX assembly computational benchmark using OpenMC Code [PDF]
Nuclear Energy and Technology, 2022 A handful of computational benchmarks that incorporate VVER-1000 assemblies having low enriched uranium (LEU) and the mixed oxide (MOX) fuel have been put forward by many experts across the world from the Nuclear Energy Agency.
Md. Imtiaj Hossain +3 more
doaj +4 more sources
Neutronic calculations for the VVER-1000 MOX core computational benchmark using the OpenMC code [PDF]
Nuclear Energy and Technology, 2023 The goal of this study is to perform neutronic calculations of the VVER-1000 MOX core computational benchmarks with an OpenMC code along with ENDF/B-VII.1 nuclear data library.
Md Imtiaj Hossain +3 more
doaj +4 more sources
The Implementation and Application of a Saudi Voxel-Based Anthropomorphic Phantom in OpenMC for Radiological Imaging and Dosimetry [PDF]
DiagnosticsObjectives: This study aimed to implement a high-resolution Saudi voxel-based anthropomorphic phantom in the OpenMC Monte Carlo (MC) simulation framework. The objective was to evaluate its applicability in radiological simulations, including radiographic
Ali A. A. Alghamdi
doaj +2 more sources
Study on application of DAG-OpenMC in fusion neutronics analysis
He jishu, 2022 BackgroundFusion reactor engineering models are extremely complex, which makes the modeling of the neutronics analysis rather tedious and time consuming.
ZHONG Gangqi +4 more
doaj +2 more sources
Simulation of NuScale-Like SMR Benchmark with OpenMC Code
Journal of Nuclear EngineeringCompared to traditional large-scale reactors, the more heterogeneous, boron-free SMR cores create additional challenges for accurate multiphysics simulations. Therefore, advanced modelling and simulation tools should be used to create high-fidelity, high-
Abdo Ez Aldeen +5 more
doaj +2 more sources