Results 11 to 20 of about 538 (156)

An Efficient Scheme for Coupling OpenMC and FLUENT with Adaptive Load Balancing

Science and Technology of Nuclear Installations, 2021
This paper develops a multi-physics interface code MC-FLUENT to couple the Monte Carlo code OpenMC with the commercial computational fluid dynamics code ANSYS FLUENT.
Qingyang Zhang, Tianji Peng, Guangchun Zhang, Jie Liu, Xiaowei Guo, Chunye Gong, Bo Yang, Xukai Fan   +7 more
doaj   +3 more sources

Implementation and benchmarking of the local weight window generation function for OpenMC

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

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, J.I. Márquez Damián, F. Molina, M. Zambra, P. Aguilera, F. López-Usquiano, B. Parra, A. Ruiz   +7 more
doaj   +2 more sources

OpenMC: Towards Simplifying Programming for TianHe Supercomputers

Journal of Computer Science and Technology, 2014
Modern petascale and future exascale systems are massively heterogeneous architectures. Developing productive intra-node programming models is crucial toward addressing their programming challenge. We introduce a directive-based intra-node programming model, OpenMC, and show that this new model can achieve ease of programming, high performance, and the
Xiangke Liao, Canqun Yang, Tao Tang 0001, Huizhan Yi, Feng Wang 0050, Qiang Wu, Jingling Xue   +6 more
openaire   +3 more sources

The OpenMC Monte Carlo particle transport code [PDF]

Annals of Nuclear Energy, 2013
Abstract 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. Given that many legacy codes do not scale well on existing and future parallel computer architectures, OpenMC has been developed from scratch with a focus on high ...
Romano, Paul Kollath, Forget, Benoit Robert Yves   +1 more
openaire   +5 more sources

Modeling and simulation of VERA core physics benchmark using OpenMC code

Nuclear Engineering and Technology, 2023
Detailed analysis of the neutron pathway through matter inside the nuclear reactor core is exceedingly needed for safety and economic considerations. Due to the constant development of high-performance computing technologies, neutronics analysis using ...
Abdullah O. Albugami, Abdullah S. Alomari, Abdullah I. Almarshad   +2 more
doaj   +2 more sources

Language and design evolution of the OpenMC Monte Carlo particle transport code [PDF]

EPJ Nuclear Sciences & Technologies
The OpenMC Monte Carlo particle transport code has been continuously developed for 13 years by a large community of contributors. In that time span, the codebase has undergone significant changes that have redefined what OpenMC is and made it an enduring
Romano Paul, Tramm John, Shriwise Patrick   +2 more
doaj   +3 more sources

OpenMC: A State-of-the-Art Monte Carlo Code for Research and Development [PDF]

SNA + MC 2013 - Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo, 2014
This paper gives an overview of OpenMC, an open source Monte Carlo particle transport code recently developed at the Massachusetts Institute of Technology. OpenMC uses continuous-energy cross sections and a constructive solid geometry representation, enabling high-fidelity modeling of nuclear reactors and other systems.
Nelson, Adam G., Romano, Paul Kollath, Horelik, Nicholas Edward, Herman, Bryan R, Forget, Benoit Robert Yves, Smith, Kord S.   +5 more
openaire   +6 more sources

Preliminary analysis of TREAT free-field experiments using OpenMC [PDF]

EPJ Web of Conferences
This work analyses activation calculations for dosimetry materials during a steady-state irradiation in the Transient Reactor Test (TREAT) reactor core.
Ferney Paul, Holschuh Tommy, Chipman Andrew, Jensen Colby   +3 more
doaj   +2 more sources

Development of continuous-energy sensitivity analysis capability in OpenMC [PDF]

Annals of Nuclear Energy, 2017
Abstract The iterated fission probability (IFP) method and the Contributon-Linked eigenvalue sensitivity/Uncertainty estimation via Track length importance CHaracterization (CLUTCH) method have been implemented in several Monte Carlo codes to perform sensitivity calculations.
Xingjie Peng, Jingang Liang, Abdulla Alhajri, Benoit Forget, Kord Smith   +4 more
openaire   +3 more sources