核燃料ZrC涂层的缺陷、超微结构设计制备及抗辐照耐腐蚀性能研究

Zirconium carbide (ZrC) has good physical and chemical properties such as high melting point, good high temperature mechanical properties, low neutron absorption cross-section as well as good resistance to the corrosion of nuclear fission products. Due to its favorable high temperature characteristics mentioned above, ZrC is being considered as one of the candidate structural materials for the production of the coated fuel particles in the generation-IV nuclear reactor systems. However, the free carbon and carbon defects is very hard to control in the preparation of ZrC coating, and the microstructure becomes more prone to coarsening at high densification temperature, which limited the radiation resistance improvement, the resistance to corrosion of fission products and its application in generation-IV nuclear reactor systems. Based previous studies of our group, the proposal aims to analysis and adjust the carbon defects in the ZrC1-x during the preparation of ZrC coating by the fluidized bed chemical vapor deposition method. The effects of carbon defects on the densification and microstructure evolution are studied. Multi-layer coating processes are used to coating the middle layers of W and Si. The solid solution reactions of W in ZrC lattice and the chemical reactions of Si with ZrC matrix are also studies, which are used to control the carbon defect, free carbon, superfine microstructure and the grain boundary strengthening during the densification of ZrC-based coating. The ion radiation resistance and the resistance to the corrosion of typical nuclear fission products are investigated. The mechanisms for the absorption of the radiation defects and the self-heal of radiation damage in the ZrC-based coating are elucidated. And the effects of defects and superfine microstructures on the corrosion resistance are also revealed. At last, the coordinated design and preparation of ZrC-based coating with good resistance to the radiation and corrosion could be realized. This proposal could supply important knowledge accumulations, data and technique supports about the application of ZrC-based coating in the generation-IV nuclear reactor systems as the coating of fuel particles.

碳化锆具有高熔点、良好的高温机械性能、低中子吸收截面、耐核裂变产物腐蚀等特点，是第四代核能系统燃料颗粒包覆层的重要候选结构材料。但ZrC涂层的自由碳和碳缺陷难精确控制，显微结构易粗化，限制了涂层的抗辐照、耐腐蚀性能及其应用。本项目拟在课题组已有工作基础上，从流化床化学气相沉ZrC基涂层的制备入手，通过对涂层组分及晶格碳缺陷的精确分析和调控来阐明碳缺陷对涂层致密化和超微结构的影响机制，进一步采用多步包覆工艺引入W和Si中间层，利用中间层与ZrC基体间发生固溶和化学反应生成晶格固溶缺陷和SiC增强相来解决涂层致密化中晶格缺陷、自由碳、超微结构及晶界强度控制的问题，系统研究ZrC基涂层抗离子辐照及耐典型裂变产物腐蚀性能，阐明材料缺陷和微结构对辐照缺陷的捕获-复合机制及耐腐蚀作用的原理，实现新型抗辐照、耐高温、耐腐蚀ZrC基涂层的协同设计及制备，为其应用提供重要的科学依据、数据积累与技术支持。

碳化锆具有高熔点、良好高温机械性能、低中子吸收截面、耐核裂变产物腐蚀等特点，是第四代核能系统的重要候选结构材料。本项目以先进核能系统用ZrC涂层为研究对象，采用磁控溅射法研究了溅射工艺参数对ZrC单相陶瓷、ZrC/SiC复合涂层、ZrC-SiC复相涂层的相组成、结晶性和微结构的影响规律，探讨ZrC涂层的碳缺陷、微结构和界面结构对涂层硬度、弹性模量等常温力学性能关系，在此基础上研究了ZrC材料的He2+、Au2+离子辐照性能，明确了材料组分、缺陷和微结构与辐照性能的关系；同时研究了W添加剂在ZrC晶格固溶的过程及其对材料微结构和界面结构影响，揭示材料烧结致密化机理，阐明ZrC在1800℃下高温力学性能的增强机制。以ZrO2微球模拟UO2燃料，制备了ZrC为基体的FCM芯块材料，研究PVB粘结剂和Si添加剂对芯块材料致密化、微结构、力学性能和热学性能的关系，阐明芯块材料微结构和热学性能改善的机理。同时，也探讨了ZrC1-x材料的腐蚀性能。本项目为先进核能系统用耐高温ZrC材料的研制提供重要的借鉴和理论指导作用。