Results 51 to 60 of about 10,412 (178)

Impact of sintering parameters on the microstructure of homogeneous U1‐xCexO2+δ ceramics

open access: yesJournal of the American Ceramic Society, Volume 108, Issue 5, May 2025.
Abstract The effects of atmosphere and cerium content on the densification and the final microstructure of homogeneous U1‐xCexO2+δ solid solutions (x = 0.10; 0.25; 0.50) were investigated. Dilatometric studies first revealed that while the cerium content only slightly modifies sintering under an Ar/H2 atmosphere, a change of the gas to argon ...
Nicolas Clavier   +6 more
wiley   +1 more source

Fundamentals of 3-D Neutron Kinetics and Current Status [PDF]

open access: yes, 2008
This lecture includes the following topics: 1) A summary of the cell and lattice calculations used to generate the neutron reaction data for neutron kinetics, including the spectral and burn up calculations of LWR cells and fuel assembly lattices, and ...
Aragonés Beltrán, José María
core   +1 more source

Exergy analysis of a PWR nuclear steam supply system – Part I, general theoretical model [PDF]

open access: yes, 2018
The paper provides an alternative, novel methodology to perform the exergetic analysis of a Pressurized Nuclear Reactor (PWR) based on the strictest definition of fission temperature to get to a careful evaluation of Exergy Destruction and exergetic ...
Ferroni, Luisa, NATALE, ANTONIO
core   +1 more source

Development and Verification of a Spatial Dynamics Code RESTA‐3D for Light Water Reactors

open access: yesInternational Journal of Energy Research, Volume 2025, Issue 1, 2025.
A neutronics/thermal‐hydraulics (TH) coupled spatial dynamics code, reactor system transient analysis‐three dimensional (RESTA‐3D), has been developed for both static and transient analysis of large advanced light water reactors (LWRs), with a focus on control rod (CR) ejection accidents.
Yong Cui   +4 more
wiley   +1 more source

Dynamic Analysis of Emergency Reset of Topaz Type Space Nuclear Reactor Safety Rod

open access: yesYuanzineng kexue jishu
The safety rod system is the actuator for reactivity control and nuclear safety protection in Topaz type space nuclear reactors. Due to the uneven distribution of reactor core temperature field during the operation of a Topaz type space nuclear reactor ...
LI Jingwei, LIU Shihang, ZHANG Guanhua, YAO Chengzhi, PENG Zhaohui, GUO Zhijia
doaj   +1 more source

Assessment of Different Turbulence Models on Melt Pool Natural Convection Simulations With Internal Heat Source

open access: yesInternational Journal of Energy Research, Volume 2025, Issue 1, 2025.
In the context of severe nuclear accidents, the migration of corium into the reactor pressure vessel (RPV) poses significant hazards, prompting the proposal of the in‐vessel melt retention (IVR) strategy, particularly the external reactor vessel cooling (ERVC) approach.
Pengya Guo   +6 more
wiley   +1 more source

Numerical Simulation of Entrained Bubbles Flow in the Shell‐Tube Heat Exchanger of MSRs Based on Population Balance Model

open access: yesInternational Journal of Energy Research, Volume 2025, Issue 1, 2025.
In molten salt reactors (MSRs), a small amount of inert gas could be entrained from the free liquid surface to the primary loop, which may have obvious impacts on the heat transfer performance of the heat exchanger, the reactivity of the core, and the migration of insoluble fission products.
Ziye Wang   +6 more
wiley   +1 more source

Geometrical optimization of PWR spacer grids using GeN-Foam and Genetic Algorithms

open access: yesBrazilian Journal of Radiation Sciences
This paper presents the results of Computational Fluid Dynamics (CFD) oriented geometrical optimization using the GeN-Foam solver applied to subchannels of the fuel assembly in a PWR-type nuclear reactor.
Carlos Rodrigo Dias   +8 more
doaj   +1 more source

Assessment of the Drift‐Flux Parameter Correlations Implemented in the Nuclear Thermal‐Hydraulic Analysis Code TRACE

open access: yesInternational Journal of Energy Research, Volume 2025, Issue 1, 2025.
The present study assessed the drift‐flux parameter correlations implemented in the TRACE code, the flagship thermal‐hydraulic analysis code developed by the United States Nuclear Regulatory Commission (US NRC). The code is architected on the basis of a two‐fluid model.
Takashi Hibiki   +2 more
wiley   +1 more source

The EPR in crisis [PDF]

open access: yes, 2010
A review of the status of the Areva European Prssurised Water Reactor (EPR)
Thomas, Stephen
core  

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