Results 141 to 150 of about 898 (171)
Some of the next articles are maybe not open access.
The Bimodal Epithermal Astronautical Reactor in OpenMC
Journal of the British Interplanetary SocietyFrom 2020 through to 2022, the Universities Space Research Association's Center for Space Nuclear Research undertook neutronic simulations of a NERVA PeeWee-derived nuclear thermal rocket reactor in both Serpent and MCNP. The most recent (unpublished) configuration of the reactor was rebuilt in the open source program OpenMC, in order to test the ...
openaire +1 more source
DAG-OpenMC: CAD-Based Geometry in OpenMC
Transactions of the American Nuclear Society - Volume 122, 2020A. Davis, P. Shriwise, X. Zhang
openaire +1 more source
OpenMC Model Validation of the TRIGA Mark II Reactor
2023The development of open-source applications in the nuclear field has recently attracted interest from the scientific community, due to the potential mutual support useful both in the research field and in the analysis of new nuclear reactor concepts.
Lorenzo Loi +2 more
openaire +1 more source
Modeling Decay Heat with a Simplified Depletion Chain in OpenMC
OpenMC can be used to computationally model depletion and produce estimates of decay heat. As an input to depletion simulations, OpenMC requires a depletion chain that details nuclide transmutation pathways. The simplified CASL depletion chain was designed to track relatively few nuclides while still accurately modeling the effective neutron ...Gupta, Tanmay, Forget, Benoit
openaire +1 more source
FEniCSx-OpenMC Coupling for Neutronic Calculation with Temperature Feedback
2023The state of an operating nuclear reactor depends on several interdependent physical phenomena, which can be considered simultaneously by modelling the system using a multi-physics (MP) approach. MP allows a higher level of detail of the system’s properties at the expense of code complexity and computational burden, whereas, in the past, single-physics
Stefano Riva +3 more
openaire +1 more source
Calculation of kinetic parameters β and Λ with modified open source Monte Carlo code OpenMC(TD)
Nuclear Engineering and Technology, 2022J Romero-Barrientos, F Molina
exaly
Verification of depletion capability of OpenMC using VERA depletion benchmark
Annals of Nuclear Energy, 2022Jiankai Yu, Benoit Forget
exaly
Benchmarking and verification of the OpenMC code for accelerator-based neutron source analyses
Fusion Engineering and Design, 2021Yuan Hu, Yuefeng Qiu, Yudong Lu
exaly
Integration of OpenMC methods into MAMMOTH and Serpent
2016Leslie Kerby, Mark DeHart, Aaron Tumulak
openaire +1 more source
Extension and benchmarking of the OpenMC code for accelerator-based neutron source applications
Fusion Engineering and Design, 2020Yuan Hu, Yuefeng Qiu
exaly