About this Research Topic
Nuclear fission energy plays an important role in wouldwide energy supplies and the safety is the focus of the nuclear industry development. After the Fukushima disaster, the concept of Accident Tolerant Fuels (ATF) is proposed to improve the inherent safety during accidents. However, the existing UO2/Zircolay-4 fuel-cladding system can hardly meet the requirements of ATF. Therefore, it is urgent to develop fuels and cladding alternative materials with improved performance under operation and accident conditions. Accordingly, many problems such as the design and fabrication of new fuel and cladding materials, the irradiation and corrosion induced performance degradation mechanism of new materials and behaviors of novel and engineering nuclear materials under accident conditions, remain to be studied.
One of the major goals of research in current nuclear materials science is to develop new generation of fuel-cladding system materials with improved performance under operation and accident conditions for fission reactors. The candidates for ATF fuels include but are not limited by Uranium alloys, and the U containing compounds(e.g. uranium nitrides, uranium silicides) or composites with enhanced thermomechanical properties, and the well known examples for cladding materials are FeCrAl alloys, coated Zircalloy, SiC composites and so on. In this Research Topic, we aim to provide new insight into the design and evaluation of promising candidate materials for high performance fuel pellets and thin cladding tubes and thus drive the advances in modern nuclear materials. The articles will cover works by experimental and theoretical investigations at different scales on the performance characterization of nuclear materials under both operation and accident conditions. We believe that the worldwide contributions of this collection can offer diverse perspectives towards solving scientific problems of common interest for the nuclear materials community, such as fission gas release behavior for advanced fuels, irradiation resistance and mechanical property degradation for claddings, mechanisms of irradiation and corrosion induced microstructure evolutions of materials, etc.
Most manuscripts in this Research Topic will be Original Research articles, but we also welcome a few Review articles that provide outlook to guide the nuclear materials community. The scope for the topics is as follows:
• New Materials for Nuclear Application: the studies on potential nuclear materials such as fuels (ceramic, metallic or composite fuels), claddings (both Ferrous and non-ferrous metals, specialty composites) or important components for reactors and their key physical properties and advantages related to performance for ATF application.
• Progress in Nuclear Material Fabrication and Characterization: the new development in fabrication and characterization technologies for novel and engineering nuclear materials.
• New Models for Nuclear Material Service Performance and Behavior Prediction: the new models for evaluation of microstructure and properties evolutions under operation and accident conditions.
• New Understanding of Materials In-Core Behaviors and Damage Mechanisms: The experimental and theorical studies that give new insight into the in-core behaviors and damage mechanisms for advanced nuclear materials (such as fuels, metals, specialty ceramics or composites) under conditions such as irradiation, corrosion, swelling and creep.
One of the major goals of research in current nuclear materials science is to develop new generation of fuel-cladding system materials with improved performance under operation and accident conditions for fission reactors. The candidates for ATF fuels include but are not limited by Uranium alloys, and the U containing compounds(e.g. uranium nitrides, uranium silicides) or composites with enhanced thermomechanical properties, and the well known examples for cladding materials are FeCrAl alloys, coated Zircalloy, SiC composites and so on. In this Research Topic, we aim to provide new insight into the design and evaluation of promising candidate materials for high performance fuel pellets and thin cladding tubes and thus drive the advances in modern nuclear materials. The articles will cover works by experimental and theoretical investigations at different scales on the performance characterization of nuclear materials under both operation and accident conditions. We believe that the worldwide contributions of this collection can offer diverse perspectives towards solving scientific problems of common interest for the nuclear materials community, such as fission gas release behavior for advanced fuels, irradiation resistance and mechanical property degradation for claddings, mechanisms of irradiation and corrosion induced microstructure evolutions of materials, etc.
Most manuscripts in this Research Topic will be Original Research articles, but we also welcome a few Review articles that provide outlook to guide the nuclear materials community. The scope for the topics is as follows:
• New Materials for Nuclear Application: the studies on potential nuclear materials such as fuels (ceramic, metallic or composite fuels), claddings (both Ferrous and non-ferrous metals, specialty composites) or important components for reactors and their key physical properties and advantages related to performance for ATF application.
• Progress in Nuclear Material Fabrication and Characterization: the new development in fabrication and characterization technologies for novel and engineering nuclear materials.
• New Models for Nuclear Material Service Performance and Behavior Prediction: the new models for evaluation of microstructure and properties evolutions under operation and accident conditions.
• New Understanding of Materials In-Core Behaviors and Damage Mechanisms: The experimental and theorical studies that give new insight into the in-core behaviors and damage mechanisms for advanced nuclear materials (such as fuels, metals, specialty ceramics or composites) under conditions such as irradiation, corrosion, swelling and creep.
Keywords: nuclear fuels, cladding, structural materials, nuclear reactor
Important Note: All contributions to this Research Topic must be within the scope of the section and journal to which they are submitted, as defined in their mission statements. Frontiers reserves the right to guide an out-of-scope manuscript to a more suitable section or journal at any stage of peer review.
