Results 1 to 10 of about 647 (119)
A New Coupled CFD/Neutron Kinetics System for High Fidelity Simulations of LWR Core Phenomena: Proof of Concept [PDF]
Science and Technology of Nuclear Installations, 2014 The Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is investigating the application of the meso- and microscale analysis for the prediction of local safety parameters for light water reactors ...
Jorge Pérez Mañes +3 more
doaj +7 more sources
Hybrid microscopic depletion model in nodal code DYN3D [PDF]
Annals of Nuclear Energy, 2016 The paper presents a general hybrid method that combines the micro-depletion technique with correction of micro- and macro- diffusion parameters to account for the spectral history effects.
Yuri Bilodid +2 more
exaly +8 more sources
Explicit decay heat calculation in the nodal diffusion code DYN3D [PDF]
Annals of Nuclear Energy, 2018 The residual radioactive decay heat plays an important role in some accident scenarios and, therefore, needs to be accurately calculated when performing accident analyses.
Yuri Bilodid, Emil Fridman, Dan Kotlyar
exaly +6 more sources
MODELLING ASTRID-LIKE SODIUM-COOLED FAST REACTOR WITH SERPENT-DYN3D CODE SEQUENCE [PDF]
EPJ Web of Conferences, 2021 This study explores the feasibility of applying the Serpent-DYN3D sequence to the analysis of Sodium-cooled Fast Reactors (SFRs) with complex core geometries, such as the ASTRIDlike design.
Rydlewicz Wojciech +2 more
doaj +3 more sources
Science and Technology of Nuclear Installations, 2014 A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses.
Surian Pinem +2 more
doaj +2 more sources
TRANSIENT CALCULATIONS OF SPERT III EXPERIMENTS [PDF]
EPJ Web of Conferences, 2021 Cold-startup and hot-standby reactivity accident tests conducted at the SPERT III E-core research reactor are analysed with the coupled neutron-kinetic/thermal-hydraulic code system DYN3D-ATHLET.
Pautz Andreas, Zwermann Winfried
doaj +1 more source
NEUTRONIC MODELLING OF THE FFTF CONTROL ROD WORTH MEASUREMENTS WITH DIFFUSION CODES [PDF]
EPJ Web of Conferences, 2021 This paper presents an assessment of three deterministic core simulators with the focus on the neutronic performance in steady-state calculations of small Sodium cooled Fast Reactor cores.
Nikitin E. +4 more
doaj +1 more source
NEUTRON NOISE PATTERNS FROM COUPLED FUEL-ASSEMBLY VIBRATIONS [PDF]
EPJ Web of Conferences, 2021 The neutron flux fluctuation magnitude of KWU-built PWRs shows a hitherto unexplained correlation with the types of loaded fuel assemblies. Also, certain measured long-range neutron flux fluctuation patterns in neighboring core quadrants still lack a ...
Viebach M. +4 more
doaj +1 more source
COUPLED TRANSIENT ANALYSIS OF A CORE WITH FUEL ASSEMBLY BOWING WITH A HYBRID CTF/DYN3D MODEL [PDF]
EPJ Web of Conferences, 2021 Fuel assembly bowing is a known phenomenon observed in many PWR reactors all over the world. The phenomenon is relevant to safety as it can lead to increased water gaps between assemblies which results in higher pin peaking factors.
Périn Y, Travleev A, Zilly M
doaj +1 more source
DYN3D and CTF Coupling within a Multiscale and Multiphysics Software Development (Part I)
Energies, 2021 Understanding and optimizing the relation between nuclear reactor components or physical phenomena allows us to improve the economics and safety of nuclear reactors, deliver new nuclear reactor designs, and educate nuclear staff.
Sebastian Davies +7 more
doaj +1 more source