新型燃料元件LOCA条件下复杂行为及安全特性研究

Loss of Coolant Accident (LOCA) of nuclear power plant is one of the design basis accidents. The behavior of nuclear fuel pellet and cladding in the event of LOCA is one of the most important research directions in various countries in the world. In order to make up for the lack of research on the key behavioral characteristics of chinese new zirconium alloy N36 in LOCA, this program intends to carry out researches of N36 zirconium alloy basic behaviors, typical experiments and numerical methods. In this program, high temperature steam oxidation, hydrogen absorption and phase transformation behaviors of N36 zirconium alloy will be studied. High temperature steam oxidation and ring compression test of N36 zirconium alloy after quench will be carried out to establish the kinetic model of N36 high temperature oxidation, hydrogen absorption model and phase transformation model. Failure threshold of N36 zirconium alloy will be studied. Experimental study of quenching process and heat transfer mechanism of fuel rod with the N36 zirconium alloy as the cladding material will be studied, and the new minimum film boiling temperature model as well as the advanced quench model will be proposed. Based on the results of the above model studies, a numerical simulation method of N36 zirconium alloy behavior will be studied for the typical failure characteristics under LOCA. In this program, high temperature oxidation model, hydrogen absorption model as well as phase transition model will be built, N36 zirconium alloy cladding failure threshold will be obtained and minimum film boiling model and quench code of N36 zirconium alloy will be established. A numerical simulation mothed of N36 zirconium alloy under LOCA will be built preliminarily.

核电站反应堆冷却剂丧失事故是设计基准事故之一，核燃料元件芯块和包壳在失水事故下的行为是世界各国重点研究的方向之一。为弥补我国自主新型N36锆合金在失水事故下关键特性研究的匮乏，拟进行核电站失水事故下N36锆合金材料基础行为、典型实验和数值分析方法的研究。本项目将进行N36锆合金高温氧化、吸氢及相转变特性研究，结合N36锆合金高温蒸汽氧化实验和淬火后环压实验，建立N36锆合金高温氧化、吸氢和相转变动力学模型；研究N36锆合金包壳失效阈值；基于N36锆合金包壳进行骤冷实验研究、沸腾转变机理研究，建立最小膜态沸腾温度模型、骤冷模型；基于上述模型研究成果，针对失水事故下典型失效特征研究N36锆合金包壳行为数值模拟分析方法。本项目将建立N36锆合金高温氧化、吸氢、相转变模型，得到N36锆合金包壳失效阈值，建立最小膜态沸腾模型和骤冷分析程序，初步建立N36锆合金包壳在失水事故下行为数值模拟分析方法。

压水堆冷却剂丧失事故下核燃料元件芯块和包壳的行为是当前世界各国重点研究的方向之一。对核电站失水事故下我国自主新型N36锆合金材料基础行为、典型实验和数值分析方法开展研究，可弥补N36锆合金在失水事故下关键特性研究的匮乏，对燃料组件国产化自主化具有重要意义。针对N36锆合金包壳燃料元件失水事故下面临的极端工况，结合燃料设计基准事故，系统性地开展了N36锆合金失水事故下氧化、吸氢和相转变、骤冷试验、沸腾转变机理和模型及数值模拟分析研究等方面的研究。首次开展了N36锆合金高温氧化、吸氢及相转变特性研究，攻克了模拟LOCA事故条件下锆合金高温蒸汽氧化实验和淬火后环压实验技术，创新建立了N36锆合金高温氧化、吸氢和相转变动力学模型；率先建立了N36锆合金包壳失效阈值；突破N36锆合金包壳骤冷实验研究技术、探究沸腾转变机理，创新建立最小膜态沸腾温度模型、骤冷分析模块；基于上述模型研究成果，针对失水事故下典型失效特征创新建立了N36锆合金包壳行为数值模拟分析方法。项目共发表重要期刊和会议论文26篇，其中含SCI期刊论文13篇，EI论文8篇；申请专利7项，软件著作权2项；攻克了包括N36锆合金包壳失效阈值、最小膜态沸腾温度模型和数值模拟方法在内的5项关键技术；培养了博士后1名，博士生3名，硕士生3名。依托项目，以组委会成员组织并参加国际学术会议1次，参加国内外学术会议多次，有力地推动了事故下核燃料元件芯块和包壳行为研究进程。