加工表面完整性对核电高温水环境下奥氏体不锈钢应力腐蚀行为影响研究

Austenitic stainless steels (SS) are widely used to produce critical structural components in nuclear power plants due to their excellent corrosion resistance and high mechanical properties at elevated temperatures. Nonetheless, stress corrosion cracking (SCC) has occurred in some stainless steel structural components exposed to pressurized water reactor (PWR) primary environment.During manufacture, stainless steel components pass through a number of different processes,which result in heavy residual stresses,retained strain,microstructural modifications in the material and electrochemical changes in the interface.The variations from fabrication process affect the mechanical and corrosion resistance properties.However，the effects of surface cold work on SCC is still unclear. The objectives of this work are the following: Firstly, a detailed analysis of machined surface and the surface layer is performed for SS,the surface integrity is quantified by means of surface roughness parameters, strain-hardening effects and associated residual stresses. Extensive characterization of the deformed microstructure of austenitic SS, employing high resolution electron backscattered diffraction will be conducted. Secondly, Oxide films on cold-worked surface exposed to simulated PWR water will be characterized. The nature of the components presents in the oxide film and the influence of initial surface finish on the microstructures of the oxide film formed on cold worked surface will be studied. Thirdly,the effect of machined surface on stress corrosion cracking (SCC) susceptibility of SS will be investigated in a simulated high temperature water environment. The test material was subjected to different levels of surface integrity. We will identify a number of relationships between the SCC initiation behavior and, for example, strain localization and microstructural parameters produced by machining. Finally, an engineering model will be presented for assessing key influences of surface cold-work in SCC of austenitic SS in PWR environments. the model incorporates a useful measure of surface cold-work relating the surface integrity and strain hardening properties. The research results can give further information to unraveling the SSC initiation mechanisms of the machined surface with relevance to the surface microstructures, improving lifetime management of components in nuclear power station, supporting the optimization of manufacturing processes.

奥氏体不锈钢是核电站设备重要的结构材料，应力腐蚀是其重要失效形式,而材料冷加工变形会加速应力腐蚀裂纹扩展。核电站结构构件在制造与装配过程中进行的表面机械加工、表面处理会影响材料表层的微观结构、金属物理特性和界面反应，从而影响材料在核电站高温水环境中的服役性能，但表面冷加工对材料应力腐蚀行为的影响机制还不清楚。本项目首先定量分析研究机械加工对表面完整性的影响；然后，在核电压水堆高温水条件下，通过应力腐蚀实验和腐蚀表面微观组织分析，研究机械加工表面对氧化膜成膜特性和应力腐蚀性能的影响，揭示材料表面机械处理导致的材料宏/微观尺度性能参数与应力腐蚀开裂动力学之间内在联系的规律性。最后，结合模拟核电站高温水中材料服役性能评价试验，建立机械加工表面对应力腐蚀耐受性评价模型。研究成果可为建立核电站高温水环境下零部件表面加工规范、优化加工工艺参数和评价材料服役可靠性提供基础理论支持。

奥氏体不锈钢是核电设备重要结构材料，应力腐蚀是其重要失效形式。核电站不锈钢构件在制造时进行的切削加工会导致表面完整性的改变，从而影响结构应力腐蚀性能。.课题研究了切削加工对奥氏体不锈钢表面完整性的影响。在理论上，根据切削过程的热-力耦合作用，建立了残余应力计算模型；基于相变动力学和位错密度理论，提出了加工表层马氏体相变和硬度变化预测模型。通过理论模型，实现了从切削力到表面微结构、残余应力的模拟。在实验上，通过EBSD等技术研究了切削导致的变质层微观结构的变化，提出了基于EBSD取向差参数的加工变质层残余变形的评价方法。.课题研究了模拟核电压水堆环境下加工表面氧化膜成膜和退化行为。通过检测不同加工表面氧化膜微观结构和元素分布，明确了加工对氧化膜各化学元素分布和价态影响，揭示了加工表面微结构与表面氧化膜成膜特性的关联关系。分析表明加工影响氧化膜厚度、降低膜厚度均匀性。表面加工变形程度影响界面处Ni、Cr和Fe元素的扩散。研究还发现铣削加工表面存在氧化膜-基体早期剥落。加工导致的微结构变化和氧化导致的基体去合金化造成材料性能退化是导致氧化膜下基体局部剥落的原因。.课题研究了奥氏体不锈钢不同加工表面在沸腾氯化镁溶液和模拟核电高温水环境中应力腐蚀开裂行为，揭示了加工表面残余应力、微结构对应力腐蚀萌生的影响规律。研究表明316和304不锈钢发生应力腐蚀裂纹萌生的残余应力阈值为190-200 MPa。这一阈值为评价加工表面残余应力对应力腐蚀微裂纹的影响提供了重要参数。研究还发现，微裂纹易在加工导致的滑移带密集区萌生，裂纹初期扩展与滑移带分布有较大关联。最后，建立了面向应力腐蚀的加工表面完整性评价模型。引入加工表面名义应力强度因子和等效裂纹深度作为加工表面评价参数，二者与表面应力腐蚀裂纹萌生有良好的关联，可作为表征加工表面抗应力腐蚀的评价指标。.课题的研究为改进加工工艺、降低加工表面应力腐蚀性能提供了科学依据。