摘要
核燃料元件的包壳材料是反应堆安全的重要屏障。随着核动力反应堆向高燃耗、长燃料循环寿命、高安全性趋势的发展,传统Zr合金包壳材料因其铀燃耗极限(62 MW·d/kg)、高温腐蚀、氢脆、蠕变、辐照生长、芯/壳反应等缺陷,已不能满足未来第四代核能系统燃料元件对包壳材料的苛刻要求。Si C因其更小的中子吸收截面、低衰变热、高熔点及优异的辐照尺寸稳定性等优点,以Si C为基体的陶瓷基复合材料成为新一代包壳材料研究的热点。结合Si C的晶体结构、热物理特性,对其在第四代核反应堆包壳材料中的设计思路、中子辐照效应、热一力性能、与UO,的化学反应等进行了概述,对SiC基复合材料在未来核能领域的应用前景进行了展望。
Fuel cladding materials are the essential barrier for the safety of nuclear reactor.With the fuel development tendency of high burn-up,long cycling life and high safety,issues of fuel consumption limit(62 MW · d/kg U),corrosion at high temperature,hydrogen embrittlement,creep deformation,irradiation growth and fuel-cladding reaction of zirconium alloys can not meet special requirements for fuel elements of Generation IV nuclear system calling for new cladding materials.Due to the smaller neutron absorption cross-section,low decay heat,high melting point and irradiation size stability,the nuclear-grade SiC/SiC composites are considered attractive and promising materials for fission system fuel cladding.According to the crystal structure and thermos-physical properties of SiC,the design concept,neutron irradiation effect,thermal-mechanical property and the chemical reaction with fuel UO_2 are summarized,and the future prospects of SiC/SiC composites in nuclear fuel applications are proposed.
引文
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