【作者】 杨富强；

【导师】 薛河；

【作者基本信息】 西安科技大学 ， 矿山机电工程， 2014， 博士

【摘要】 奥氏体不锈钢和镍基合金在核电站高温水环境中的环境致裂(EAC)行为严重威胁着核电站的安全，准确预测EAC裂纹扩展速率(CGR)已成为核电安全评价的关键问题之一。为了获得EAC裂纹扩展速率，本文采用理论分析和有限元结合的方法，研究了裂尖蠕变力学特征，建立了基于裂尖蠕变与膜破裂的核电结构材料EAC裂纹扩展速率定量预测模型。主要研究内容有：（1）以滑移溶解理论和Ford-Andresen模型为基础，根据裂尖蠕变主导裂尖钝化膜破裂的思想，重新诠释了裂尖钝化膜的破裂机理及EAC裂纹扩展历程，推导建立了以裂尖蠕变率为主要力学参量的EAC裂纹扩展速率定量预测模型。（2）采用理论分析与数值计算结合的方法，得到了裂纹扩展速率与裂尖电化学参量之间的关系，结果表明裂纹扩展速率与钝化膜形成时间和阳极溶解电流密度正相关，与电流衰减指数负相关。通过建立裂尖电化学环境计算模型，研究了裂纹缝隙中物质的传输和缝隙电化学行为，得到了金属阳极电位、溶液pH值及缝隙尺寸对裂尖局部电化学环境的影响规律。（3）以标准紧凑拉伸试样为研究对象，通过建立裂尖蠕变的有限元模型，研究出了蠕变对裂尖力学特征的影响，确定将裂尖正前方位置的蠕变率用于预测模型；分析得到了不同电化学参数下，材料力学性能和裂尖力学状态对蠕变率、Mises应力、等效塑性应变等裂尖蠕变力学特征及裂纹扩展速率预测模型的影响。（4）对新建立的EAC定量预测模型、FRI模型、裂尖应变梯度模型及实验数据进行对比，获得了不同参数下各模型的特点及与实验数据的差别，结果表明采用蠕变致膜破裂的裂纹扩展速率预测模型进行高温水环境中EAC裂纹扩展速率预测是可行的。（5）焊接造成的非匀质在核电关键结构中普遍存在，本文通过建立焊接接头的简化有限元模型，研究了蠕变对材料力学性能不均匀裂尖力学特征的影响，结果表明蠕变将显著改善材料力学性能不均匀引起的裂尖应力分布差异。通过将基于蠕变与膜破裂的CGR定量预测模型应用到焊接接头中，得出了非匀质材料中EAC裂纹扩展速率的规律。本文提出的基于蠕变致裂尖膜破裂机理的EAC定量预测模型，为核电结构材料安全评价和寿命预测提供了新的思路和依据。更多还原

【Abstract】 The safety of nuclear power plants is threatened by environmentally assistedcracking(EAC) of austenitic stainless steels and nickel-based alloys in high temperature waterenvironments. And the accurately prediction of EAC crack growth rate(CGR) has become oneof the key issues in safety assessment of nuclear power plants. By adopting theoreticalanalysis and finite element method, the creep characteristics of crack tip was studied, and thequantitative predictive model of EAC in structure materials of nuclear power plants was alsobuilt based on passivation film rupture induced by the creep at the crack tip. The main worksin this dissertation are as follow:(1) The quantitative predictive model of EAC crack growth rate was built according thethe idea that the rupture of crack tip passivation film in high temperature water environmentsis dominant by the creep of metal based on slip dissolution theory and Ford-Andresen model.The mechanism of the crack tip passivation film ruptures and EAC crack growth process werereinterpreted.(2) The relationship between CGR and electrochemical parameters at the crack tip werestudied by theory and numerical calculation, the results indicates that the CGR is positivelycorrelated with duration of constant current density and oxidization current density, and isnegatively correlated with exponent of current decay curve. The mass transport andelectrochemical behavior in the crack were studied according to the electrochemicalenvironment modeling built, and the effects of anode potential, pH and crack size on localelectrochemical parameters at crack tip were achieved.(3) The effects of the creep on the mechanical characteristics at the crack tip werestudied with a compact tension specimen and finite element method, and the creep strain ratein front of the crack tip was selected to use in the EAC rate prediction model. The effects of material mechanical properties and stress status of crack tip on mechanical characteristics ofcreep crack tip such as creep strain rates, Mises stress and equivalent plastic strain underdifferent electrochemical parameters were obtained.(4) The comparisons under different model parameters among predictive CGR of thenew model proposed in this dissertation, FRI model, crack tip strain gradient model andexperimental data were taken, the results indicate that it is practicable to predicate the EACrate in high temperature water by the model based on crack tip creep induced passivation filmrupture.(5) The inhomogeneous of material caused by weld are common in key structures ofnuclear power plants. The effects of creep on mechanical characteristics of weld joints werestudied by finite element model of simplified weld joints, the results indicate that thedifferences of stress distribution caused by weld inhomogeneous would be reduceddramatically by creep. The EAC rates in the weld joint were also achieved by applying theCGR prediction model to the weld joints.The proposed EAC quantitative predictive model, which is based on the crack tip creepinduced passivation film rupture, provides new method to evaluate the safety and predict thelife of structural materials in nuclear power plants.更多还原