反应堆用SiC陶瓷基复合包壳材料研究进展

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。

Current Status and Recent Research Achievements in SiC Composites for Fuel Cladding

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 embrit-tlement, 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 UO2 are summarized, and the future prospects of SiC/SiC composites in nuclear fuel applications are proposed.