反应堆用 SiC 陶瓷基复合包壳材料研究进展

摘要

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.%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。SiC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SiC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SiC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-力性能、与UO2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。