综述了近年来在核电高温高压水中不锈钢和镍基合金的腐蚀电化学行为规律, 以及材料表面膜的成分、组织结构和电子特征. 通过系统分析水介质化学和材料微观结构对表面膜的影响, 试图把膜的特性与腐蚀电化学行为联系起来. 讨论了核电高温高压水中材料的腐蚀机制, 提出核电高温高压水中不锈钢和镍基合金的腐蚀机制与常温时不同, 高温下是由电化学与氧化联合控制, 而不是传统意义上常温下的纯电化学腐蚀. 表面膜为双层结构, 外层疏松, 而内层是由纳米晶构成的具有半导体性质的膜层, 内层是控制腐蚀的关键. 材料的微观成分与组织结构、表面加工状态、水化学参数是影响核电用材料高温高压水中腐蚀的重要因素. 腐蚀产生的氧化物的楔形力导致缺陷前端产生局部拉应力, 即使在宏观压应力区, 缺陷前端的局部拉应力仍可导致应力腐蚀开裂的发生与扩展. 氧化物楔形力的作用是促进压应力下产生应力腐蚀开裂的重要原因.

采用扫描电镜、原子力显微镜和表面粗糙度测量仪对具有不同表面状态的690TT合金表面形貌进行了表征与比较。采用零电荷电位测量、动电位扫描和电化学快慢扫描等方法对不同的690TT合金的腐蚀行为进行了比较。结果表明，与机械抛光样品相比较，打磨样品表面起伏较严重，拥有更大的表面粗糙度值；在相同的腐蚀环境中，打磨样品比机械抛光样品表现出更大的腐蚀速度和更高的应力腐蚀开裂敏感性。分析认为，单纯的表面较大粗糙度和残余应变均能够促进690TT合金的腐蚀。实验中打磨690TT样品表现出的较高腐蚀速度和应力腐蚀开裂敏感性是由其较大的表面粗糙度和表面残余应变 综合影响结果。

<p>The effects of surface state and applied stress on the stress corrosion cracking (SCC) behaviors of thermally treated (TT) Alloy 690 in 10 wt% NaOH solution with 100 mg/L litharge at 330&nbsp;&deg;C were investigated using C-ring samples with four kinds of surface states and two different stress levels. Sample outer surfaces of the first three kinds were ground to 400 grit (ground), shot-peened (SP) and electro-polished (EP) and the last one was used as the as-received state. Two samples of every kind were stressed to 100% and 200% yield stress of Alloy 690TT, respectively. The results showed that the oxide film consisted of three layers whereas continuous layer rich in Cr was not found. The poor adhesive ability indicated that the oxide film could not protect the matrix from further corrosion. Lead was found in the oxide film and the oxides at the crack paths and accelerated the dissolution of thermodynamically unstable Cr in these locations and also in the matrix. The crack initiation and propagation on Alloy 690TT were effectively retarded by SP and EP treatments but were enhanced by grinding treatment, compared with the cracks on the as-received surface. The cracking severity was also enhanced by increasing the externally applied stress. The accelerated dissolution of Cr and the local tensile stress concentration in the near-surface layer caused by cold-working and higher applied stress reduced the SCC-resistance of Alloy 690TT in the studied solution.</p>

Oxidation of Alloy 690TT samples either manually ground to 400 and 1500 grit, mechanically polished, or electropolished was performed in a solution of 1500 &#x000D7; 10^-6 B and 2.3 &#x000D7; 10^-6 Li with 2.5 &#x000D7; 10^-6 dissolved H₂, at 325 &#x000b0;C and 15.6 MPa for 60 days. The oxide films grown on samples with different surface states were analyzed using various techniques. Results show that a triple-layered structure was formed after immersion: an outermost layer with large scattered oxide particles rich in Fe and Ni, an intermediate layer with small compact oxide particles rich in Cr and Fe for the ground surfaces and loose needle-like oxides rich in Ni for the polished surfaces, and an inner layer with continuous Cr-rich oxides. The surface state was found to affect not only the surface morphology, but also the corrosion rate. Grinding accelerated the growth of protective oxide films such that the ground samples showed a lower oxidation rate than the polished ones. Samples of ground Alloy 690TT showed superior resistance to intergranular attack (IGA).

The electropolished (EP) alloy 690TT samples were first oxidized in the simulated B and Li containing primary water with 2.5 mg/L H2 at 325 ℃ and 15.6 MPa for 720 h, and then half of the samples were continuously immersed in this solution with 2.0 mg/L O2 for another 720 h. The microstructures and chemical composition of the oxide films formed under the above two conditions were analyzed. The results show that the dual layered oxide film formed under the single hydrogen water chemistry is mainly composed of spinel oxides. The outer layer is composed of big oxide particles rich in Ni and Fe and the underlying loose needle-like oxides rich in Ni. The inner layer is continuous Cr-rich oxides. The oxide film formed on EP alloy 690TT under the hydrogen/oxygen water chemistry also shows a dual layered structure. The surface morphology and chemical composition of the outer layer are similar to the oxide film formed under the hydrogen water chemistry. However, the inner layer is changed to the nano-sized NiO. The stable phase region in the potential-pH diagram for the Ni oxides is enlarged by the later dissolved oxygen. As a result, the oxygen promotes the fast growth of the outer needle-like oxides rich in Ni. Further, the oxygen promotes the dissolution of the inner Cr-rich oxides formed under the hydrogen water chemistry and increases the corrosion rate of the EP alloy 690TT. Electropolishing treatment can not reduce the corrosion rate of alloy 690TT in the simulated primary water with sequentially dissolved hydrogen and oxygen.

将电解抛光态690TT合金样品在325 ℃, 15.6 MPa,含1500 mg/L B, 2.3 mg/L Li, 2.5 mg/L H₂的高温高压水中连续浸泡720 h后, 取出一半样品用于腐蚀产物的分析, 其余样品继续在含2.0 mg/L O₂的该高温高压水中连续浸泡720 h. 采用SEM, GIXRD和TEM分析了在上述2种条件下样品表面生长的氧化膜的微观结构. 结果表明, 电解抛光态690TT合金在单一溶氢的高温高压水中表面生长的氧化膜具有双层结构: 外层是分散的富含Ni和Fe大颗粒氧化物和富含Ni的疏松的针状氧化物; 内层是近连续的富含Cr的氧化物; 内外层氧化物均具有尖晶石结构. 在溶氢/溶氧的溶液中连续浸泡后, 样品表面生长的氧化膜也具有双层结构: 外层的形貌、化学组成和物相结构与在单一溶氢条件下生长的氧化膜相似, 仅针状氧化物的长度明显增加; 而氧化膜内层变成了纳米尺寸的NiO. 后期溶解氧扩大了电位-pH图中含Ni的氧化物稳定存在的相区, 促进了外层富含Ni的针状氧化物的快速生长; 更加重要的是, 溶解氧提高了含Fe和Cr氧化物的腐蚀电位, 促进了在溶氢条件下生长的内层富Cr氧化物的溶解, 破坏了氧化膜的保护性结构, 提高了电解抛光态690TT合金的腐蚀速度. 在一回路溶氢/溶氧连续浸泡过程中, 电解抛光处理并不能降低690TT合金的腐蚀速度.

The microscopic residual stresses beneath scratches on thermally treated (TT) UNS N06690 have been measured using synchrotron radiation x-ray diffraction; the effect of these residual stresses on initiating stress corrosion cracking (SCC) has been studied in 10% sodium hydroxide (NaOH) + 10 g/L lead oxide (PbO) solution at 330°C. Results show that scratching causes local residual tensile stresses parallel to the scratching direction at the scratched surface and local residual compressive stress normal to the scratching direction beneath the scratch. The residual tensile stresses at the scratched surface will act as a precursor of SCC. The residual compressive stresses beneath the scratch cannot stop crack growth of N06690 TT in 10% NaOH + 10 g/L PbO solution at 330°C.

The influence of surface conditions on the corrosion behavior of engineering structures has been paid more attention. However, there is still a lack of systematic research on the effect of cutting parameters on material's microstructure and performance in service. In this paper, the effect of cutting parameters on microstructure and corrosion behaviors of 304 stainless steel in simulated primary water is well investigated. The results show that different cutting parameters can cause the superficial layer a gradient microstructure with nanocrystalized layer on top and deformation band structures underneath. With the similar surface roughness, the deformation microstructure can be very different due to the different cutting parameters. The effect degree on the depth of deformation zone is feed rate > cutting depth > cutting speed. The larger feed rate, lower cutting depth, lower cutting rate may induce a deeper deformation zone. With the increasing depth away from the machined surface, the localized corrosion rate is decreased, and at the same depth the localized corrosion rate along the deformation bands is higher than that along the grain boundaries (GBs). The nanocrystalized surface has a smallest general corrosion rate due to the quick formation of Cr rich oxide film. However, once the corrosion penetrates through this nanocrystalized layer, subsequent preferential corrosion at deformation bands and GBs will dominate and may lead to the significant increase of corrosion rate of the component in high temperature pressurized water.

利用划伤技术研究了690TT合金在325 ℃高温含氧硼锂水中的裂纹萌生和生长情况。试样表面和截面显微分析的结果表明，划伤沟槽底部局部萌生了典型的沿晶应力腐蚀裂纹。由于应力集中，在慢速率拉伸阶段划伤沟槽底部产生了机械裂纹，而机械裂纹成为恒载过程中690TT合金沿晶应力腐蚀裂纹萌生和生长的先导。尖端非常接近晶界或者沿着晶界的机械裂纹可继续形成沿晶应力腐蚀裂纹。690TT合金在恒载荷条件下对应力腐蚀开裂仍有一定的敏感性。

Three kinds of samples with different scratch depths on the surface of alloy 690TT for heat exchanger tubes are prepared. The microstructure changes caused by scratch and the effect of scratches on the corrosion and SCC behavior were systematically investigated. The results show that severe plastic deformation occurred near the scratch bank and the scratched groove. A large number of sliding steps and tear deformation generated near the scratched bank. The scratch depth had no obvious influence on the morphology, composition and distribution of corrosion products in different areas of scratches. As the scratch depth increased both the number and the length of SCC cracks increased, indicating that the SCC sensitivity of the material increased. Defects such as slip steps and micro cracks produced during the scratching process were likely to be the preferential positions of SCC initiation.

制备三种表面有不同深度划伤的690TT合金传热管试样并观察划伤导致的微观组织变化，研究了表面划伤对腐蚀、应力腐蚀行为的影响。结果表明：在传热管表面划伤的划伤堤和划伤谷附近出现了严重的塑性变形，并在划伤堤附近产生了大量的滑移台阶和撕裂变形征；划伤深度对不同划伤区域腐蚀产物的形貌、成分和分布等没有明显的影响；随着划伤深度的增大SCC裂纹的数量和长度都呈增加的趋势，说明材料的SCC敏感性增强；划伤产生的滑移台阶、微裂纹等缺陷，易成为SCC裂纹优先萌生位置。

Alloy 690TT tube samples with different scratch depths were repaired by grinding treatments using abrasive papers of two different particle sizes. The microstructure and stress corrosion cracking (SCC) behavior were studied in detail. During grinding, the plastic accumulation zone vulnerable to SCC was removed. Meanwhile, some residual slip steps remained in the scratched area. Corrosion tests lasting 1000, 2000, 3000, and 4000 h show that the sensitivity and risk of SCC in the scratched area are decreased by grinding. Treatment using abrasive particles of a smaller size is more effective. Nevertheless, deep scratches remained hazardous even after the grinding.

<p>The radiation-induced segregation (RIS) and microstructure evolution such as dislocation loops and cavities are major microstructural causes for the irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steel (SS) core components. While a couple of studies have been reported on the irradiation induced damage in nuclear grade (NG) austenitic SS, the evolution of dislocation loop density and size and its correlation with the mechanical properties have still remained incompletely understood. In addition, the correlation between the segregation at the grain boundary and that at the dislocation loop has received limited attentions. In particular, there is still a lack of a systematic study of the irradiation damage in domestically fabricated NG austenitic SS. In this work, the proton-irradiation induced microstructural damage in domestically fabricated 304NG SS was characterized, in an effort to correlate the RIS and the dislocation loop density and size with the irradiation dose, as well as the dislocation loop density and size with the radiation-induced hardening. The results revealed that the radiation-induced microstructure damage was mainly dislocation loops with a few micro-voids. The loop density was in the order of 10²² m^-3 with an average size of <10 nm. The square root of the product of loop density and size (<i>Nd</i>)^0.5, scaled linearly with the square root of irradiation dose with a factor of 6.8×10³ dpa^-0.5mm^-1. The loops were believed to be mainly responsible for the hardening in 304NG SS, which also scaled linearly with (<i>Nd</i>)^0.5 with a factor of 1.16×10^-2 HV_0.025mm. A comparative analysis about the segregation at the grain boundary and at the dislocation loop was conducted. While the depletion of Cr and enrichment of Ni at the dislocation loop and grain boundary showed no difference, the enrichment of Si at the dislocation loop could be of about 6 times of that at the grain boundary. In addition, the loop density and loop size, as well as RIS and radiation-induced hardening were all increased by a higher dose and tended to saturate by a dose of 3.0~5.0 dpa.</p>

Irradiation assisted stress corrosion cracking (IASCC) of austenitic stainless steel core components is one major concern for maintenance of nuclear power plants. Previous studies on the IASCC had mainly focused on the effect of irradiation on changes in deformation modes and interaction of dislocation channels with grain boundary. The role of corrosion in IASCC, however, has not received sufficient attentions. 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. In this work, IASCC of proton-irradiated nuclear grade 304 stainless steel (304SS) was investigated. The IASCC tests were conducted by interrupted slow strain rate tensile (SSRT) tests at 320 ℃ 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 304SS. Since the slip step usually transmitted or terminated at the grain boundary, it eventually promoted localized deformation at 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 of the cracks was increased by a higher irradiation dose and applied strain. This suggested a higher IASCC susceptibility of nuclear grade 304SS 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.

Localized deformation and corrosion in irradiated 304 nuclear grade stainless steel in simulated primary water were investigated. The investigation was conducted by comparing the deformation structure, the oxide scale formed at the deformation structure, and their correlation with cracking. The results revealed that increasing the irradiation dose promoted localized corrosion at the slip step and grain boundary, which was primarily attributed to the strain concentration induced by enhanced localized deformation and depletion of Cr at grain boundary. Further, a synergic effect of the enhanced localized deformation and localized corrosion at the slip step and grain boundary caused a higher cracking susceptibility of the irradiated steel.