Inter-Code Comparison of Computational VERA Depletion Benchmark Using OpenMC, OpenMC-ONIX and DRAGON [PDF]
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
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 +4 more sources
Calculation of kinetic parameters βeff and Λ with modified open source Monte Carlo code OpenMC(TD) [PDF]
Nuclear Engineering and Technology, 2022 This work presents the methodology used to expand the capabilities of the Monte Carlo code OpenMC for the calculation of reactor kinetic parameters: effective delayed neutron fraction βeff and neutron generation time Λ.
J. Romero-Barrientos +7 more
doaj +4 more sources
The OpenMC Monte Carlo particle transport code [PDF]
Annals of Nuclear Energy, 2012 A new Monte Carlo code called OpenMC is currently under development at the Massachusetts Institute of Technology as a tool for simulation on high-performance computing platforms.
Forget, Benoit Robert Yves +1 more
core +5 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
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 +2 more sources
Implementation and benchmarking of the local weight window generation function for OpenMC [PDF]
Nuclear Engineering and Technology, 2022 OpenMC is a community-driven open-source Monte Carlo neutron and photon transport simulation code. The Weight Window Mesh (WWM) function and an automatic Global Variance Reduction (GVR) method was recently developed and implemented in a developmental ...
Yuan Hu, Sha Yan, Yuefeng Qiu
doaj +3 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
Perhitungan Shutdown Margin Teras NuScale Menggunakan OpenMC [PDF]
Jurnal Fisika Unand, 2023 One of the important reactor safety parameters to study is the issue of shutdown margin (SDM). This study aims to obtain an effective safety design of the NuScale reactor in reviewing the SDM value parameter.
Raflis, Helen +2 more
core +3 more sources
Validation of OpenMC Code for Low-cycle and Low-particle Simulations in the Neutronic Calculation [PDF]
JIF (Jurnal Ilmu Fisika)Validation of Low-Cycle and Low-Particle OpenMC Simulation Codes for Neutronics Calculations has been conducted. This study validates OpenMC, an evolving open-source neutron analysis code.
Ahmad Muzaki Mabruri +4 more
doaj +3 more sources