Results 61 to 70 of about 898 (171)

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

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   +1 more source

Validation of Multiphysics Simulation Using OpenFOAM–PRAGMA Coupled Code for Heat Pipe Cooled Microreactor Core With the KRUSTY Experiment

International Journal of Energy Research, Volume 2024, Issue 1, 2024.
A heat pipe cooled microreactor (HPMR) offers advantages such as compact design, ease of transportation, and improved system reliability and safety. The core of the HPMR consists of a solid structure called a “monolith,” which contains multiple fuel rods and heat pipes (HPs).
Myung Jin Jeong, Jaeuk Im, San Lee, Hyoung Kyu Cho, Mahmoud Ahmed   +4 more
wiley   +1 more source

Neutronic Analysis of Accident‐Tolerant Cladding Materials in 3D Full Core BEAVRS PWR Benchmark Using OpenMC Code

International Journal of Energy Research, Volume 2024, Issue 1, 2024.
After the Fukushima Daiichi nuclear accident in 2011, the performance of nuclear fuel during accidents became a matter of great concern. To address this, a new type of fuel technology called accident‐tolerant fuel (ATF) has been developed with the goal of enhancing the ability of light water reactors (LWRs) to withstand severe accident conditions. Iron‐
Khalid A. Alamri, Meshari M. Alqahtani, Abdullah S. Alomari, Abdullah I. Almarshad, Andrey Kuzmin   +4 more
wiley   +1 more source

Comparison of Hexagonal and Square Fuel Pin Arrangement with UN-PuN Fuel in PWR

Computational and Experimental Research in Materials and Renewable Energy
Indonesia is experiencing an increasing demand for electrical energy, which can be met through alternative sources such as nuclear energy generated in nuclear reactors at Nuclear Power Plants (NPPs).
Muhammad Syu’bi Alwi, Ratna Dewi Syarifah, M. Khanifuddin Zuhri, Wenny Maulina   +3 more
doaj   +1 more source

Discrepancy across various OpenMC versions due to thermal neutron scattering law [PDF]

EPJ Web of Conferences
The performance of neutron transport calculations is heavily reliant on the fidelity of nuclear data. The Free Gas Model (FGM) is no longer applicable at low neutron energy range due to nucleus binding effects, necessitating the implementation of the ...
Wang Tianxiang, Chen Shengli, Xu Shuqi, Li Zhuo   +3 more
doaj   +1 more source

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   +1 more source

OpenMC In Situ Source Convergence Detection

, 2016
We designed and implemented an in situ version of particle source convergence for the OpenMC particle transport simulator. OpenMC is a Monte Carlo based-particle simulator for neutron criticality calculations. For the transport simulation to be accurate, source particles must converge on a spatial distribution.
Garrett Aldrich, Soumya Dutta, Jonathan Woodring   +2 more
openaire   +2 more sources

Coupled neutronic-thermal-mechanical simulation of the KRUSTY heat pipe microreactor

Frontiers in Nuclear Engineering
Multiphysics analysis has become a common technique for nuclear reactor design validation, with neutronic-thermal analysis being the typical choice for understanding reactor dynamics.
William Reed Kendrick, Benoit Forget
doaj   +1 more source

On-the-fly doppler broadening of unresolved resonance region cross sections via probability band interpolation [PDF]

, 2016
In this work we present a scheme for computing temperature-dependent unresolved resonance region cross sections in Monte Carlo neutron transport simulations.
Brown, Forrest B., Forget, Benoit Robert Yves, Smith, Kord S., Walsh, Jonathan Alan   +3 more
core  

Direct, on-the-fly calculation of unresolved resonance region cross sections in Monte Carlo simulations [PDF]

, 2015
The theory, implementation, and testing of a method for on-the-fly unresolved resonance region cross section calculations in continuous-energy Monte Carlo neutron transport codes are presented.
Brown, Forrest B., Forget, Benoit Robert Yves, Kiedrowski, Brian C., Smith, Kord S., Walsh, Jonathan Alan   +4 more
core