反应堆用SiC陶瓷基复合包壳材料研究进展
|
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.
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
[1]H.Feinroth,M.Ales,E.Barringer,G.Kohse,et al.Mechanical Strength of CTP Triplex SiC Fuel Clad Tubes after Irradiation in MIT Research Reactor under PWR Coolant Conditions[J].Ceramic Engineering and Science Processing,2009,30(10):47-55.
[2]Daejong Kim,Hyun-Geun Lee,Ji Yeon Park,et al.Fabrication and Measurement of Hoop Strength of SiC Triplex Tube for Nuclear Fuel Cladding Applications[J].Journal of Nuclear Materials,2015,458:29-36.
[3]L.L.Snead,S.J.Zinkle,D.P.White.Thermal Conductivity Degradation of Ceramic Materials Due to Low Temperature,Low Dose Neutron Irradiation[J].Journal of Nuclear Materials,2005,340:187,202.
[4]Yutai Katoh,Kazumi Ozawa,Chunghao Shin,et al.Continuous SiC ber,CVI SiC Matrix Composites for Nuclear Applications:Properties and Irradiation Effects[J].Journal of Nuclear Materials,2014,448:448-476.
[5]M.Ferraris,M.Salvo,V.Casalegno,et al.Joining of SiC-based Materials for Nuclear Energy Applications[J].Journal of Nuclear Materials,2011,417:379-382.
[6]Mohd Idzat Idris,Saishun Yamazaki,Katsumi Yoshida,et al.Recovery Behavior of High Purity Cubic SiC Polycrystals by Post-irradiation Annealing up to 1673 K after Low Temperature Neutron Irradiation[J].Journal of Nuclear Materials,2015,465:814-819.
[7]Eric Rohmer,Eric Martin,Christophe Lorrette.Mechanical properties of SiC/SiC Braided Tubes for Fuel Cladding[J].Journal of Nuclear Materials,2014,453:16-21.
[8]Chinthaka M.Silva,Yutai Katoh,Stewart L.Voit,et al.Chemical Reactivity of CVC and CVD SiC with U02 at High Temperatures[J].Journal of Nuclear Materials,2015,460:52-59.
[9]Y.Katoh,L.L.Snead,I.Szlufarska,et al.Radiation Effects in SiC for Nuclear Structural Applications[J].Current Opinion Solid State Material Science,2012,16:143-152.
[2]Daejong Kim,Hyun-Geun Lee,Ji Yeon Park,et al.Fabrication and Measurement of Hoop Strength of SiC Triplex Tube for Nuclear Fuel Cladding Applications[J].Journal of Nuclear Materials,2015,458:29-36.
[3]L.L.Snead,S.J.Zinkle,D.P.White.Thermal Conductivity Degradation of Ceramic Materials Due to Low Temperature,Low Dose Neutron Irradiation[J].Journal of Nuclear Materials,2005,340:187,202.
[4]Yutai Katoh,Kazumi Ozawa,Chunghao Shin,et al.Continuous SiC ber,CVI SiC Matrix Composites for Nuclear Applications:Properties and Irradiation Effects[J].Journal of Nuclear Materials,2014,448:448-476.
[5]M.Ferraris,M.Salvo,V.Casalegno,et al.Joining of SiC-based Materials for Nuclear Energy Applications[J].Journal of Nuclear Materials,2011,417:379-382.
[6]Mohd Idzat Idris,Saishun Yamazaki,Katsumi Yoshida,et al.Recovery Behavior of High Purity Cubic SiC Polycrystals by Post-irradiation Annealing up to 1673 K after Low Temperature Neutron Irradiation[J].Journal of Nuclear Materials,2015,465:814-819.
[7]Eric Rohmer,Eric Martin,Christophe Lorrette.Mechanical properties of SiC/SiC Braided Tubes for Fuel Cladding[J].Journal of Nuclear Materials,2014,453:16-21.
[8]Chinthaka M.Silva,Yutai Katoh,Stewart L.Voit,et al.Chemical Reactivity of CVC and CVD SiC with U02 at High Temperatures[J].Journal of Nuclear Materials,2015,460:52-59.
[9]Y.Katoh,L.L.Snead,I.Szlufarska,et al.Radiation Effects in SiC for Nuclear Structural Applications[J].Current Opinion Solid State Material Science,2012,16:143-152.