Results 61 to 70 of about 5,558 (186)
Numerical Study on Laminar-Turbulent Transition Flow in Rectangular Channels of a Nuclear Reactor
Frontiers in Energy Research, 2020 Laminar-turbulent transition flow can be observed in thermal engineering applications, but the flow resistance and heat transfer characteristics are not fully understood.
Zhenying Wang +6 more
doaj +1 more source
GEN-IV LFR development: Status & perspectives [PDF]
, 2018 Since Lead-cooled Fast Reactors (LFR) have been conceptualized in the frame of Generation IV International Forum (GIF), great interest has focused on the development and testing of new technologies related to Heavy Liquid Metal (HLM) nuclear reactors. In
Bassini, Serena +6 more
core +1 more source
Science and Technology of Nuclear Installations, Volume 2026, Issue 1, 2026.Sodium‐cooled fast reactor (SFR) is one of the fourth‐generation nuclear reactor types, which has inherent safety features such as low corrosion and no severe interaction between molten fuel and sodium. However, the severe accident of core melting in SFR needs to be considered.
Yingying Huo +3 more
wiley +1 more source
Thermal-Hydraulics analysis of pressurized water reactor core by using single heated channel model [PDF]
Journal of Applied and Computational Mechanics, 2017 Thermal hydraulics of nuclear reactor as a basis of reactor safety has a very important role in reactor design and control. The thermal-hydraulic analysis provides input data to the reactor-physics analysis, whereas the latter gives information about the
Reza Akbari +2 more
doaj +1 more source
International Journal of Energy Research, Volume 2025, Issue 1, 2025.This study presents a numerical modeling approach for analyzing the electrochemical and thermal–hydraulic characteristics of solid oxide electrolysis (SOE) cells and stacks for hydrogen production. A semilumped parameter‐based simulation code was developed and validated using experimental data and literature references.
Jiwon Ahn +3 more
wiley +1 more source
Assessment of RELAP/SCDAPSIM/MOD3.4 Prediction Capability with Severe Fuel Damage Scoping Test
Science and Technology of Nuclear Installations, 2017 The Power Burst Facility (PBF) was designed to provide experimental data to determine the thresholds for failure during accident conditions. Thus, the PBF benchmark using severe accidental analysis codes is essential to designing reactor for current ...
Noppawan Rattanadecho +4 more
doaj +1 more source
Reactivity feedback effect on loss of flow accident in PWR
Nuclear Engineering and Technology, 2018 In this work, the reactor kinetics capability is used to compute the design safety parameters in a PWR due to complete loss of coolant flow during protected and unprotected accidents.
Basma Foad +2 more
doaj +1 more source
DEVELOPMENT OF THE ALTERNATE PRESSURIZED THERMAL SHOCK RULE (10 CFR 50.61a) IN THE UNITED STATES [PDF]
, 2009 In the early 1980s, attention focused on the possibility that pressurized thermal shock (PTS) events could challenge the integrity of a nuclear reactor pressure vessel (RPV) because operational experience suggested that overcooling events, while not ...
Departament d'Arquitectura i Projectes Urbans +1 more
core +2 more sources
International Journal of Energy Research, Volume 2025, Issue 1, 2025.Increased energy supply reliability, environmental sustainability, more competitive businesses, and improved standard of living are all made possible by more efficient energy conversion processes. In view of this, the present study aims at analysis and optimization of the operational parameters of a dual‐fuel‐fired boiler using Taguchi design method to
Sunday O. Oyedepo +5 more
wiley +1 more source
Analysis on Steady-State Operation and Heat Loss of Chinese Integrated Pressurized Water Reactor
Science and Technology of Nuclear Installations, 2014 Chinese integrated pressurized water reactor (CIPWR) has compact configuration and high inherent safety, which is appropriate for nuclear power plants of small and medium scale.
Zhang Fan +4 more
doaj +1 more source