Results 131 to 140 of about 898 (171)

OFELIA: Openmc-FEnicsx for muLtiphysics tutorIAl

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3D neutronic analysis on compact fusion reactors: PHITS-OpenMC cross-comparison [PDF]

Ferrero, Gabriele, Hartwig, Zachary, Laviano, Francesco, Ledda, Federico, Meschini, Samuele, Pettinari, Davide, Sparacio, Simone, Testoni, Raffaella, Torsello, Daniele, Trotta, Antonio, Zucchetti, Massimo   +10 more
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OpenMC Interpretation of FNS SINBAD Shielding Benchmark Experiments

Fusion Science and Technology
Bamidele Ebiwonjumi, Stefano Segantin, Ethan Peterson   +2 more
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AtOMC: A Port of OpenMC for Atomic Transport and Plasma Interaction

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Implementation and verification of PyNE R2S with DAG-OpenMC

Fusion Engineering and Design, 2020
Abstract The mesh-based Rigorous-Two-Step (R2S) method has been widely used in the accurate estimation of the shutdown dose rate (SDR) of fusion systems. Several mesh-based R2S code has been implemented based on Monte Carlo particle transport code MCNP5 or DAG-MCNP5 and inventory calculation code such as FISPACT-II, ALARA or ACAB.
Xiaokang Zhang, Patrick C Shriwise, Songlin Liu   +2 more
exaly   +2 more sources

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

ERSN-OpenMC-Py: A python-based open-source software for OpenMC Monte Carlo code

Computer Physics Communications
M Lahdour, O El Hajjaji, H Ziani
exaly   +2 more sources

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
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