Lipeng Du , Xiang Chen , Wenchao Zhang , Jianchuang Sun , Weihua Cai
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引用次数: 0
Abstract
The thermal–hydraulic analysis of coolant flow in the reactor core plays a significant role in the optimized design of nuclear fuel assemblies and nuclear safety. This article reviews the experimental and Computational Fluid Dynamics (CFD) methods for the thermal–hydraulic characteristics of subchannels with water as the coolant in different rod bundle geometries and flow conditions over the past few decades. It summarizes the effects of design parameters such as rod spacing and spacer grid type, as well as flow parameters like pressure, mass flow rate, and heat flux on fluid flow and heat transfer. Additionally, various models used for numerical simulations of rod bundle channels are introduced, providing valuable references for research and practice in related fields.
期刊介绍：
Nuclear Engineering and Design covers the wide range of disciplines involved in the engineering, design, safety and construction of nuclear fission reactors. The Editors welcome papers both on applied and innovative aspects and developments in nuclear science and technology.
Fundamentals of Reactor Design include:
• Thermal-Hydraulics and Core Physics
• Safety Analysis, Risk Assessment (PSA)
• Structural and Mechanical Engineering
• Materials Science
• Fuel Behavior and Design
• Structural Plant Design
• Engineering of Reactor Components
• Experiments
Aspects beyond fundamentals of Reactor Design covered:
• Accident Mitigation Measures
• Reactor Control Systems
• Licensing Issues
• Safeguard Engineering
• Economy of Plants
• Reprocessing / Waste Disposal
• Applications of Nuclear Energy
• Maintenance
• Decommissioning
Papers on new reactor ideas and developments (Generation IV reactors) such as inherently safe modular HTRs, High Performance LWRs/HWRs and LMFBs/GFR will be considered; Actinide Burners, Accelerator Driven Systems, Energy Amplifiers and other special designs of power and research reactors and their applications are also encouraged.
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