材料期刊网

采用数值模拟方法对核电站316L不锈钢弯头的应力腐蚀裂纹扩展行为进行了研究。首先针对不锈钢厚壁弯头(外径355.6 mm,内径275.6 mm)进行有限元建模,在弯头内壁上创建出与实际裂纹相符的半椭圆状3D缺陷作为裂纹形状,其裂纹张开位移(δ_i)由Dugdale模型计算确定;然后根据有限元计算结果,建立裂纹应力强度因子 (K) 随裂纹深度 (a) 及附加应力 (P) 变化的拟合公式,结合实验数据得到管材在2种冷变形量下的应力腐蚀裂纹扩展速率(da/dt)拟合公式,利用迭代方法计算了裂纹穿透管壁所需的时间,为核电站安全评估提供了有效依据。研究显示,当弯头部位的冷变形量较小(硬度为230~245 HV)且在理想情况下 (无初始附加应力),弯头被应力腐蚀裂纹穿透耗时最长(约57 a);当初始附加应力增加至200 MPa,此失效时间约缩减至前者的1/5 (无应力释放)、2/7 (应力释放一半) 以及3/7 (应力完全释放);保持初始附加应力不变(200 MPa)并提高弯头部位冷加工变形量(由硬度为230~245 HV提高到275~300 HV),弯头的大变形部位被穿透时间约缩短至小变形部位失效时间的2/5 （无应力释放）、3/8 (应力释放一半) 以及1/3 (应力完全释放),由此可见应力释放程度的降低和冷加工变形量的增加均导致了核电站316L不锈钢弯头剩余寿命的缩短。

Stress corrosion cracking (SCC) is one of the main ageing mechanism in light water reactor (LWR). 316L austenitic stainless steel was adopted in nuclear industry for its relatively high corrosion resistance. The SCC of austenitic stainless steel may occur as it is subjected to both the tensile stress and the caustic medium, with regard to maintaining the structural integrity of components in nuclear power plant, an accurate prediction and efficient assessment of the component lifetime is significant and necessary. The stress corrosion crack propagation behavior of the 316L stainless steel elbow of nuclear power plant was investigated through a numerical simulation method. Firstly a finite element (FE) model was created for the stainless steel thick-walled elbow (the outer diameter is 355.6 mm, the inner diameter is 275.6 mm), with a semi-elliptical shaped 3D defect introduced at the internal surface of the elbow as the geometry of the crack, which was consistent with a practical crack, the crack opening displacement (δ_i) was determined by the calculations through the Dugdale model; subsequently, according to the FE calculation results, establish the fitting formula of the stress intensity factor (K) varying with the crack depth (a) and additional stress (P), and the fitting formula of the stress corrosion crack propagation rate (da/dt) for elbows under two types of cold work deformation was deduced through the combination with the experimental data, the crack propagation time was then calculated using a iterative method for cracks which evolved from different initial crack depth values to certain crack depth values. The calculation results provided effective reference criterion for the nuclear power plant safety assesment. This investigation demonstrated that, when the cold deformation extent of the elbow part is relatively small ( with the hardness of 230~245 HV) and it is under the ideal condition (no initial additional stress), it takes around 57 a for the stress corrosion crack to penetrate the elbow, when the initial additional stress was elevated to 200 MPa, the elbow failure time was shrinked to 1/5 (no stress release), 2/7 (half-stress release) and 3/7 (total stress release) of the former; keep the same initial additional stress (200 MPa) and increase the cold work deformation extent (the hardness was increased from 230~245 HV to 275~300 HV), the elbow failure time was shortened to 2/5 (no stress release), 3/8 (half-stress release) and 1/3 (total stress release) for the elbow part with higher cold deformation extent compared to the part with lower cold deformation extent, thus it was observed that both the decrease of the extent of stress relaxation and the increase of the extent of cold work deformation contributed to the reduction of the residual life of the nuclear power plant 316L stainless steel elbow.