国产新型压水堆燃料包壳材料SiC/SiC的辐照效应及微观机理研究

After the Fukushima nuclear accident in Japan, fiber reinforced SiC/SiC composite has been considered as an indispensable new material to replace Zircaloy nuclear fuel cladding for Pressurized Water Reactor (PWR). Its excellent high temperature strength, good resistance to radiation damage and high temperature chemical inertness can ensure no nuclear leakage occurring during nuclear accidents. However, currently,this kind of materials is seriously lack of the reliable in-pile service data under the operation conditions of Pressurized Water Reactor in the world.Based on undertaking the national related projects for development and fabrication of the SiC/SiC nuclear fuel cladding tubes in our laboratory, and using new SiC/SiC composites developed by ourselves,this project will focus more on the study of the microstructural evolution and its consequent results in mechanical property, thermal conductivity and irradiation swelling/creep evolution rules. On these grounds, to study the failure behaviors, the related micro-mechanisms and the controlling factors of the SiC/SiC composites under PWR　conditions, and to preliminary establish a physical model for assessing the material service life.Furthermore,combined with the micro-structural evolution and macro property changes of the material, to deduce the degradation mechanisms of the material after in-pile PWR irradiation conditions and to initially establish the material performance database under irradation conditions. Then to provide a basis for design optimization of the SiC/SiC composite material for nuclear fuel cladding, and to provide a technical support for cladding fabrication and in-pile assessment.The project conforms to the national strategy demandings. Besides international synchronized,this research has its pioneering and innovative nature. This project is significant important not only for developping a new generation of the domestic anti-nuclear accident fuel cladding tubes for PWR reactors, but also for developping a advanced nuclear fuel cladding material for the advanced fast breeder reactors in the future.

福岛事故后，国际一致认为纤维增韧SiC/SiC复合材料是替代压水堆燃料锆包壳不可或缺的新材料，其优异的高温强度、耐辐照和高温化学惰性可有效避免事故时发生核泄漏。但目前国内外都十分缺乏该材料在压水堆况下可靠的服役数据，迫切需要掌握该材料在相应堆况下的辐照行为。本项目以我室承担压水堆SiC/SiC核燃料包壳管研制的国家任务为基础，采用自行研制的SiC/SiC复合材料，将重点研究压水堆况辐照下该材料的微观结构演化及其力学性能、热导性和辐照肿胀/蠕变演变规律：据此研究SiC/SiC在压水堆况下的失效行为、微观机理及其控制性因素，初步建立评估该材料服役寿命的物理模型，为核包壳管用SiC/SiC复合材料的优化设计提供依据，为包壳管结构设计/制备和下堆考核提供理论支撑。该课题符合国家战略需求，研究与国际同步，具有前沿性和创新性，可为国产新型抗核事故压水堆SiC/SiC复合材料燃料包壳管的发展奠定基础。

福岛核事故后，研制可代替锆包壳材料，正常运行时性能与锆合金相当或更安全、经济；事故时能在足够长时间内保持堆芯完整，确保核燃料、裂变产物和放射性气体不泄露的新一代压水堆耐事故燃料包壳材料成为国际上安全发展核电所迫切需要解决的关键问题。连续碳化硅纤维增强碳化硅复合材料以其较小的中子吸收截面、耐辐照性以及优异的高温稳定性和化学惰性被公认为是新一代最耐事故的燃料包壳材料。本项目以国产压水堆SiC/SiC复合材料包壳管的研制为基础，采用国产三代SiC纤维（Amosic-3）增强的一维和二维SiC/SiC复合材料，开展了质子、离子以及中子等辐照。主要内容包括：研究了一维SiC/SiC复合材料在室温及300℃经2.8 MeV质子辐照至0.002~0.003dpa 的微观结构演化及力学失效行为，结果表明由于辐照剂量较低，复合材料各结构组元中均未有明显非晶化区域形成，这也导致复合材料在辐照前后的拉伸性能未发生显著变化。研究了二维SiC/SiC复合材料在300℃经300 keV硅离子辐照（最大剂量~10dpa）的微观结构演化及力学失效行为，结果表明纤维和基体在辐照剂量小于1.5dpa时产生不同程度的非晶化，这导致二者的纳米硬度和杨氏模量均有一定程度下降。研究了二维SiC/SiC复合材料在300℃经300 keV硅离子辐照（最大剂量~115dpa）的微观结构演化及力学失效行为，结果表明随辐照剂量增加纤维和基体均发生辐照非晶化到辐照再结晶的演变。此外由于辐照导致纤维及基体的键合结构破坏、非晶化形成以及再结晶致使纤维和基体的纳米硬度和杨氏模量呈不同程度下降。研究了二维SiC/SiC复合材料在40℃经中子辐照至0.03和0.2dpa的微观结构演化及力学失效行为，结果表明随辐照剂量增加复合材料的力学性能降低，微观结构分析认为引起复合材料宏观性能衰退主要是由于中子辐照诱导的纤维局部非晶化，界面相部分失效以及基体局部非晶化的综合作用所造成的。本项目采用不同辐照手段所获得的实验数据有力的证明了开展国产耐事故燃料包壳SiC/SiC复合材料研究是可行的。同时，本项目也为国产耐事故燃料SiC/SiC复合材料的优化设计奠定了良好的基础。