Results 121 to 130 of about 538 (156)

Development and benchmarking of the Weight Window Mesh function for OpenMC

Fusion Engineering and Design, 2021
Abstract OpenMC is a community-driven open-source Monte Carlo neutron and photon transport simulation code. It supports geometry modeling using constructive solid geometry, and is capable of performing fixed source particle transport simulation based on continuous-energy and group-wise nuclear cross-section data.
Yuan Hu, Yuefeng Qiu
exaly   +3 more sources

Development and verification of LOOP: A Linkage of ORIGEN2.2 and OpenMC

Annals of Nuclear Energy, 2017
Abstract This work presents the development of a new burn-up code named LOOP ( L inkage of ORIGEN2.2 and Op enMC) and its verification against various high burn-up benchmarks. The LOOP code links OpenMC (for static analysis of the reactor core), and ORIGEN2.2 (for time dependent fuel inventory). This code is written in Python3. LOOP uses the
Anas Gul, Khurrum Saleem Chaudri
exaly   +2 more sources

Calculation of adjoint-weighted reactor kinetics parameters in OpenMC

Annals of Nuclear Energy, 2019
Abstract Knowledge of adjoint-weighted kinetics parameters plays an importance role in analyzing the reactor safety and in describing the transient behavior of nuclear reactors during normal operation or accidents. In this work, the capability of calculating adjoint-weighted kinetic parameters, including the effective neutron generation time and the ...
Jingang Liang, Benoit Forget, Kord Smith   +2 more
exaly   +2 more sources

OpenMC Model Validation of the TRIGA Mark II Reactor

, 2023
The 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, Antonio Cammi, Carolina Introini   +2 more
openaire   +2 more sources

Investigation of the Fuel Shape Impact on the MTR Reactor Parameters Using the OpenMC Code

Processes, 2023
The goal of this study was to evaluate the impact of simulating different fuel shapes for the material testing reactor (MTR). Two OpenMC codes were built, and the first OpenMC model was simulated using a curved shape fuel element to mimic the real ...
Ahmed Alghamdi
exaly   +2 more sources

Memory efficient indexing algorithm for physical properties in OpenMC

, 2015
OpenMC is an open source Monte Carlo code designed at MIT with a focus on parallel scalability for large nuclear reactor simulations. The target problem for OpenMC is a full core high-fidelity multi-physics coupled simulation. This encompasses not only nuclear physics, but also material science and thermohydraulics.
Lax, Derek Michael
openaire   +2 more sources

The Bimodal Epithermal Astronautical Reactor in OpenMC

Journal of the British Interplanetary Society
From 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

Modeling of neutronic beam for PARR-II using OpenMC

2016 International Conference on Emerging Technologies (ICET), 2016
The objective of this study is to model a neutronic beam at Pakistan Research Reactor-II in order to obtain neutrons of epithermal energy for the therapy of deep seated tumors like glioblastoma multiforme etc. which have no other best alternative cure so far.
Amina Zulfiqar, Muhammad Sohail
openaire   +1 more source

DAG-OpenMC: CAD-Based Geometry in OpenMC

Transactions of the American Nuclear Society - Volume 122, 2020
A. Davis, P. Shriwise, X. Zhang
openaire   +1 more source

Calculation of kinetic parameters β and Λ with modified open source Monte Carlo code OpenMC(TD)

Nuclear Engineering and Technology, 2022
J Romero-Barrientos, F Molina
exaly