核用304不銹鋼輻照促進應力腐蝕開裂研究

核用304不銹鋼輻照促進應力腐蝕開裂研究Study on Irradiation Assisted Stress Corrosion Cracking of Nuclear Grade 304 Stainless Steel

采用2 MeV質子束在360 °C對國產核用304不銹鋼試樣進行了輻照實驗，利用高溫高壓水環(huán)境的慢應變速率拉伸實驗（SSRT）和掃描電子顯微鏡、電子背向散射衍射、透射電子顯微鏡等研究了核用304不銹鋼輻照促進應力腐蝕開裂（IASCC）機理。結果表明，慢應變速率拉伸過程中輻照促進材料晶界和表面滑移臺階處形成應變集中，且其程度隨輻照劑量增加而增加。滑移臺階穿過或終止于晶界，終止于晶界的臺階造成晶界處產生不連續(xù)滑移，易將位錯傳輸?shù)骄Ы纾诰Ы鐓^(qū)域形成位錯塞積和殘余應變集中。而臺階不連續(xù)滑移的形成則受毗鄰晶粒的斯密特因子對的類型影響。另一方面，輻照促進晶界發(fā)生貧Cr富Ni元素偏析，其偏析程度隨輻照劑量增加而增加。SSRT實驗后輻照試樣表面發(fā)生明顯沿晶應力腐蝕開裂，且裂紋數(shù)量隨輻照劑量和外加應變增加而增加。同時，裂紋尖端區(qū)域發(fā)生明顯晶界腐蝕，且氧化物寬度和長度隨輻照劑量增加而增加。分析認為，輻照致晶界應變集中和元素偏析的協(xié)同作用造成材料變形行為和晶界腐蝕行為變化是IASCC發(fā)生的關鍵因素。

Irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steel core components is one major concern for maintenance of nuclear power plants. While previous studies reported that changes in deformation modes and interaction of dislocation channels with grain boundary are key factors promoting the IASCC, the mechanism of IASCC of austenitic SS has still remained incompletely understood, such as the effect of material factors (crystal orientation, Schmidt factor, et al.) on grain boundary deformation and cracking. In addition, in the process of stress corrosion cracking (SCC), corrosion occurs simultaneously with localized deformation in the vicinity of the crack tip. This indicates that corrosion is one of the potential contributors to IASCC, while the role of corrosion in IASCC has not received much attentions. In this work, IASCC of proton-irradiated nuclear grade 304 stainless steel (304 SS) were investigated. The IASCC tests were conducted by interrupted slow strain rate tensile (SSRT) tests at 320 °C in simulated primary water of pressurized water reactor containing 1200 mg/L B as H3BO3 and 2.3 mg/L Li as LiOH·H2O, with a dissolved hydrogen concentration of 2.6 mg/L. Following the SSRT tests, the localized deformation, corrosion and IASCC of the specimens were characterized. The results revealed that increasing the irradiation dose promoted residual strain accumulation at slip steps and grain boundaries of nuclear grade 304 SS. Since the slip step usually transmitted or terminated at the grain boundary, it eventually impacted on the grain boundary. Specially, the slip step transmitted at grain boundary led to slip continuity at the grain boundary. In contrast, a slip discontinuity was observed at the grain boundary where the slip step terminated, which caused a much higher strain accumulation by feeding dislocations to the grain boundary region. Further, formation of the slip discontinuity was related to the Schmidt factor pair type of the adjacent grains. The irradiation resulted in a depletion of Cr and an enrichment of Ni at grain boundary, while the magnitude of Cr depletion and Ni enrichment increased with increasing the irradiation dose. Following the SSRT tests, intergranular cracking was observed on surfaces of the irradiated specimens, while the number the cracks were increased by a higher irradiation dose and applied strain, suggesting a higher IASCC susceptibility of nuclear grade 304 SS in the primary water. Meanwhile, significant intergranular oxidation ahead of the crack tip was observed, while both the width and length of the oxide were larger at a higher irradiation dose. The synergic effect of irradiation-promoted deformation and intergranular corrosion was the primary cause for the IASCC of the irradiated steel.

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