核電站316L不銹鋼彎頭應力腐蝕行為的壽命預測核電站316L不銹鋼彎頭應力腐蝕行為的壽命預測Life Prediction for Stress Corrosion Behavior of 316L Stainless Steel Elbow of Nuclear Power Plant 采用數值模擬方法對核電站316L不銹鋼彎頭的應力腐蝕裂紋擴展行為進行了研究。首先針對不銹鋼厚壁彎頭(外徑355.6 mm，內徑275.6mm)進行有限元建模,在彎頭內壁上創建出與實際裂紋相符的半橢圓狀3D缺陷作為裂紋形狀,其裂紋張開位移 (δi) 由Dugdale模型計算確定;然后根據有限元計算結果,建立裂紋應力強度因子 (K) 隨裂紋深度 (d) 及附加應力 (P) 變化的擬合公式,結合實驗數據得到管材在兩種冷變形量下的應力腐蝕裂紋擴展速率 (da/dt) 擬合公式,利用迭代方法計算了裂紋穿透管壁所需的時間，為核電站安全評估提供了有效依據.研究顯示，當彎頭部位的冷變形量較小（硬度為230~245 HV）且在理想情況下 (無初始附加應力) 彎頭被應力腐蝕裂紋穿透耗時最長 (約57 a)，當初始附加應力增加至200 MPa此失效時間約縮減至前者的1/5 (無應力釋放)、2/7 (應力釋放一半) 以及3/7 (應力完全釋放)；保持初始附加應力不變(200 MPa)并提高彎頭部位冷加工變形量(由硬度為230~245 HV提高到275~300 HV)，彎頭的大變形部位被穿透時間約縮短至小變形部位失效時間的2/5（無應力釋放）、3/8 (應力釋放一半)以及1/3 (應力完全釋放)，由此可見應力釋放程度的降低和冷加工變形量的增加均導致了核電站316L不銹鋼彎頭剩余壽命的縮短。 Stress corrosion cracking (SCC) is one of the main aging mechanism in LWR (light water reactor). 316LN 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(δi) 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 (d) 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 evolve 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.
|
