Stress Corrosion Cracking behavior of Austenitic Stainless Steels (SS) in a chloride environment has been investigated by several methods. In this investigation failure of emerging structural material for nuclear reactor (316LN grade austenitic stainless steel) and its Friction Stir Welded (FSW) joint due to stress corrosion cracking were analyzed in a chloride environment.FSW joint was fabricated between 316LN plates with optimized process parameters. Initially, hardness and tensile properties were analyzed across the weld zone to evaluate the mechanical properties of the welded joint. Microstructural characterization of base metal (BM) and weld zone were done by Optical Microscope (OM), and Scanning Electron Microscope (SEM). Grain boundary analysis was done by Transmission Electron Microscope (TEM). Microstructural evolution revealed that the FSW process is capable to refine and reduce grain size nearby 12 times that of the base metal by Continuous Dynamic Recrystallization mechanism (CDRX). Improved strength and hardness are also recorded at the weld zone compared to the base metal. Boiling 45 weight % MgCl₂ solution at constant load condition as per ASTM standard G36 was used to assess SCC failure on base metal and FSW joint. Failure due to Chloride Stress Corrosion Cracking (CSCC) on weld samples were investigated and compared with its base metal samples at different tensile stress conditions (20 %, 40%, 60%, 80%, 100% of base metal yield stress values). From the experimental result steady-state elongation rate ($I_{\mathit{SS}}$), Transition time ($t_{\mathit{ss}}$) and time to failure ($t_f$) were evaluated and generalized equations to predict time to failure of base metal and FSW weldment was generated. The impact of microstructural morphology on SCC failure of the FSW joint was investigated. The fracture location of the weld joint, nature of crack advancement of both base metal and FSW joint was analyzed by the SEM study. A study of fracture surface displayed brittle nature transgranular failure for base metal as well as FSW samples.

通过多种方法研究了奥氏体不锈钢（SS）在氯化物环境中的应力腐蚀开裂行为。在本研究中，分析了在氯化物环境下由于应力腐蚀裂纹导致的核反应堆新兴结构材料（316LN级奥氏体不锈钢）及其摩擦搅拌焊接（FSW）接头的失效。在316LN板之间制造了FSW接头，并优化了工艺参数。最初，在整个焊接区域对硬度和拉伸性能进行了分析，以评估焊接接头的机械性能。通过光学显微镜（OM）和扫描电子显微镜（SEM）对贱金属（BM）和焊接区进行了显微组织表征。晶界分析通过透射电子显微镜（TEM）进行。显微组织演变表明，FSW工艺能够通过连续动态重结晶机制（CDRX）细化和减小基体金属晶粒尺寸的12倍左右。与母材相比，焊接区域的强度和硬度也有所提高。沸腾45重量％的MgCl根据ASTM标准G36在恒定负载条件下的_2种溶液用于评估母材和FSW接头上的SCC失效。研究了在不同拉伸应力条件下（母材屈服应力值的20％，40％，60％，80％，100％）由于焊缝样品上的氯化物应力腐蚀开裂（CSCC）导致的失效，并将其与母材样品进行了比较。根据实验结果，稳态伸长率（${\mathrm{一世}}_{\mathit{SS}}$），过渡时间（${Ť}_{\mathit{ss}}$）和失效时间（${Ť}_F$）进行了评估，并建立了广义方程来预测母材和FSW焊件的失效时间。研究了微结构形态对FSW关节SCC破坏的影响。通过SEM研究分析了焊接接头的断裂位置，母材和FSW接头的裂纹扩展性质。断裂表面的研究显示出贱金属和FSW样品的脆性，穿晶破坏。