压水堆一回路管道棘轮变形失效机理及设计理论

Primary coolant circuit pipes (PCCPs) are a crucial protective barrier in pressurized water reactor (PWR) nuclear power plants. Thermal aging and fatigue damage result in accelerating ratcheting deformation and failure in PCCPs, which has drawn great attention from domestic and abroad researchers. However, no mature design standard concerning this problem has been accomplished in the world. In this project, as piping ratcheting deformation of primary coolant circuit pipes (PCCPs) in AP1000 PWR, we take the pressurized structure of the straight and elbow pipe as the object of study. Aiming at the following goals: Mechanisms of influence of loading rate, multiaxial loading, loading history, dynamic strain aging, thermal aging and fatigue damage on ratcheting deformation, as well as behaviours of cyclic plasticity of 316 type stainless steel used for PCCPs in AP1000 PWR will be investigated at elevated temperatures; Viscoplastic constitutive model to predict the accumulation of ratcheting deformation will be developed by incorporating multi-mechanism coupling; Tests of ratcheting deformation in the straight and elbow pipes will be conducted under pressure and bend multiaxial cyclic loading conditions; Time-dependent dynamic ratcheting boundaries of piping structure will be determined by analyzing the accumulation of ratcheting strain in typical structures with the newly developed viscoplastic model embedded in the commercial finite element software ANSYS. In one word, this project will establish a design theoretical basis and approach of considering total life cycle interaction of ratcheting deformation and thermal aging and fatigue damage, and finally provide for the building-up of design criteria of preventing ratcheting deformation in PCCPs of PWR. The achievements of this project will be also of significant theoretical meaning and engineering application value to the establishment of standard as well as the design of pressure equipments (such as pressure vessel and piping).

压水堆核电站一回路管道是核电站的重要屏障,而热老化和疲劳损伤导致棘轮应变加速并失效倍受关注,国内外对此并没有成熟的设计规范。针对AP1000压水堆一回路管道棘轮变形设计问题，以典型的承压结构-直管和弯管为研究对象，研究一回路管道用316型不锈钢高温循环塑性性能及棘轮变形规律，澄清加载率、多轴载荷、载荷历史效应、动态应变时效、热老化和疲劳损伤等对棘轮变形影响的机制；建立考虑这些因素影响用于预测棘轮应变积累的多机制耦合的粘塑性循环本构模型；进行直管和弯管部件在各种内压和弯曲组合多轴循环载荷下的棘轮变形测试；通过在ANSYS中嵌入建立的多机制粘塑性循环本构模型，进行典型结构的棘轮应变累积分析，建立与时间相关的管道结构动态棘轮边界，从而建立考虑全寿命周期棘轮变形与热老化和疲劳损伤交互的核压力管道设计理论和方法。研究成果对承压设备（如压力容器和管道）的设计及规范的制定具有重要理论意义和工程应用价值。

压水堆核电站一回路管道是核电站的重要屏障,而热老化和疲劳损伤导致棘轮应变加速并失效倍受关注,国内外对此并没有成熟的设计规范。针对AP1000压水堆一回路管道棘轮变形设计问题，以典型的承压结构-直管和弯管为研究对象，研究一回路管道用316型不锈钢高温循环塑性性能及棘轮变形规律，澄清加载率、多轴载荷、载荷历史效应、动态应变时效、热老化和疲劳损伤等对棘轮变形影响的机制；建立考虑这些因素影响用于预测棘轮应变积累的多机制耦合的粘塑性循环本构模型；进行直管和弯管部件在各种内压和弯曲组合多轴循环载荷下的棘轮变形测试；通过在ANSYS中嵌入建立的多机制粘塑性循环本构模型，进行典型结构的棘轮应变累积分析，建立与时间相关的管道结构动态棘轮边界，从而建立考虑全寿命周期棘轮变形与热老化和疲劳损伤交互的核压力管道设计理论和方法。主要创新点：（1）系统研究了多种加载条件下316LN不锈钢的循环塑性和低周疲劳性能，揭示了材料的棘轮变形和疲劳裂纹萌生机制，发现并解释了316LN不锈钢棘轮-疲劳寿命随平均应力增加不降反增的现象，建立了棘轮-疲劳寿命预测模型；（2）在各向同性强化律中考虑循环软化记忆效应，在随动强化律中引入反映高温下材料动态应变时效作用的强化因子和热老化影响因子，与塑性应变幅、温度、屈服应力与材料晶粒尺寸相关联，建立了描述316LN材料循环行为的统一粘塑性循环本构方程，能很好预测复杂载荷及热老化下材料的棘轮变形；（3）通过实验方法和数值模拟研究了直管和弯管在内压弯矩等复杂载荷条件下的棘轮行为，证实了所提模型对复杂载荷下管道棘轮边界的预测能力，并通过建立屈服极限和多轴参数与老化时间的关系，可以预测热老化后的棘轮边界。研究成果对承压设备（如压力容器和管道）的设计及规范的制定具有重要理论意义和工程应用价值。