T91和316L钢在氧控铅铋中600小时后腐蚀产物分析

本文开展了铅铋反应堆典型候选结构材料T91和316L钢在温度480℃、氧质量分数10~(-6)%、流速0.3 m/s的铅铋合金环境中600 h的初步腐蚀实验研究。利用显微镜(SEM)、能谱仪(EDS)和X射线衍射仪(XRD)等微观测试手段,对材料的腐蚀界面形貌及产物进行了分析,初步结果显示:在实验条件下T91和316L钢基体均未发生铅铋渗透及溶解腐蚀现象,腐蚀类型均为氧化腐蚀。T91钢样品表面氧化层较厚(约11μm),且分为内外两层:外层结构疏松,主要成分为Fe_3O_4。内层结构致密,主要成分为(Fe,Cr)_3O_4;而316L钢表面单层氧化层较薄(约1μm),主要成分为(Fe,Cr)_3O_4。     

包壳材料316Ti在液态铅铋中的腐蚀氧化层分析

本文利用自主研制的液态铅铋腐蚀试验装置,开展了中国铅基研究实验堆候选包壳材料316Ti在500℃氧饱和液态铅铋中3 000h的腐蚀试验。利用SEM及能谱仪对腐蚀后的样品进行了分析,结果显示316Ti样品腐蚀后形成了双层氧化膜,外层为疏松的Fe3O4氧化物,内层为致密的Fe-Cr尖晶石。随腐蚀时间的增加,外层氧化膜厚度变化不明显,但内层氧化膜逐渐增加,同时内层氧化膜沿晶界向基体生长,呈现出晶间腐蚀的特征。     

燃料元件包壳材料CN-1515不锈钢在可控氧铅铋环境下的腐蚀行为

国产CN-1515不锈钢因其良好的抗辐照肿胀能力和高温力学性能成为铅铋快堆燃料包壳的主要候选材料。在铅铋冷快堆中,由于液态铅铋合金对金属材料具有强烈的腐蚀性,会影响到反应堆的安全稳定运行,因此,铅铋冷快堆中结构材料应用还需充分考虑耐液态铅铋腐蚀性能。本文以国产CN-1515奥氏体不锈钢为研究对象,在自行研发的控氧静态铅铋腐蚀实验装置上,开展了高温铅铋腐蚀实验。实验温度分别为450、500、550、600 ℃,实验时间分别为1 000、3 000、6 000 h,液态铅铋合金中氧含量控制在10^-6%～10^-7%之间。实验结果表明,低温（T≤450 ℃）下,CN-1515不锈钢表面会生成一层保护性氧化膜,但随着腐蚀时间的增加,氧化膜会逐渐疏松而失去其保护作用;然而温度大于500 ℃时,不锈钢发生严重的Ni元素溶解腐蚀,腐蚀深度随温度的升高和时间的延长而增加。     

奥氏体321不锈钢在550 ℃静态铅铋共晶合金中的腐蚀行为

奥氏体321不锈钢常用作核反应堆冷却剂主管道结构材料,铅铋共晶合金是第四代核能系统（Gen Ⅳ）铅冷快堆冷却剂的主要候选材料。为研究321不锈钢与高温液态铅铋共晶合金的相容性,对321不锈钢在550 ℃液态铅铋共晶合金中的200、400、600 h腐蚀现象进行了研究。对不同腐蚀时间后腐蚀试样的表面和截面分别进行了XRD和SEM、EDS检测。结果发现：在321不锈钢试样表面产生了一种随腐蚀时间增加先生长后脱落的含O、Ti、Pb元素的化合物（Ti₂O和Pb₂O₃）;在321不锈钢基体与铅铋共晶合金交界处会产生一层随腐蚀时间增加不断增厚的扩散层;321不锈钢在铅铋共晶合金中发生溶解腐蚀,在Fe、Cr元素不断向铅铋共晶合金中溶解时,伴随着Pb、Bi元素向基体中的渗透。     

奥氏体ODS钢在超临界水中的腐蚀行为

研究了奥氏体ODS钢（316-ODS）在600 ℃/25 MPa超临界水（SCW）中的腐蚀特性。采用腐蚀增重法、SEM、EDS和XRD分析了材料的氧化动力学、氧化膜的形貌、合金元素分布和组织结构。研究结果表明,316-ODS钢在SCW中出现了疖状腐蚀,同时还出现了敏化,其腐蚀增重服从幂指数生长规律。316-ODS钢表面氧化膜为双层结构,内层氧化膜富Cr贫Fe,其主要成分为FeCr₂O₄,而外层氧化膜富Fe贫Cr,其主要成分为Fe₃O₄。     

卧式铅铋堆芯氧化腐蚀特性研究

为开展卧式铅铋堆芯氧化腐蚀特性研究,本研究建立液态铅铋氧化腐蚀模型,并基于计算流体动力学方法,运用输运方程源项自定义方法实现耦合计算。研究表明,基准工况下堆芯燃料棒表面氧化层最厚位于出口位置处,中心位置燃料棒表面氧化层厚度显著高于靠近燃料组件盒燃料棒表面氧化层。10000 h后中心位置燃料棒表面仍保持双层氧化层结构,双层氧化层平均总厚度为1.32μm。本研究提出了铅铋堆芯氧化腐蚀特性数值模拟研究方法,能够用于铅铋堆芯氧化腐蚀的预测。     

304NG在超临界水中腐蚀后氧化膜分析

研究了奥氏体不锈钢304NG(以下简称304NG)在压力为25 MPa,温度分别为500、550、600、650℃超临界水中的腐蚀行为,通过扫描电镜-电子能谱(SEM-EDX)、X射线衍射(XRD)对304NG试样氧化膜微观组织的研究表明:304NG在超临界水中腐蚀后,表面氧化膜由岛状和非岛状2种不同形貌的腐蚀相组成.其中,含岛状腐蚀相的氧化膜具有双层结构,外层为Fe3O4相,内层为Fe3O4和FeCr2O4相;不含岛状腐蚀相的氧化膜为单层结构,氧化膜中含有Fe3O4和FeCr2O4相.同时,304NG在超临界水中氧化膜存在脱落现象,氧化膜脱落程度随温度升高而加剧.     

铁素体/马氏体钢和奥氏体不锈钢的液态铅铋腐蚀行为与机理

液态铅铋共晶合金[liquid lead-bismuth eutectic,LBE,Pb44.5Bi55.5,%(质量分数)]具有优异的热工水力和中子学性能,是第四代液态金属冷却快堆最重要的冷却工质之一。但是,液态铅铋冷却快堆的主要候选材料包括铁素体/马氏体钢(如T91)和奥氏体不锈钢(如316L和15-15Ti)存在液态金属腐蚀问题,一定程度上阻碍了液态铅铋快堆工程化应用进度。本文综述了液态铅铋腐蚀的基本机制以及铁素体/马氏体钢和奥氏体不锈钢的液态铅铋腐蚀行为,总结了抑制液态铅铋腐蚀的主要方法,并展望了未来研究方向。     

Zr(Fe,Cr)_2金属间化合物在500℃过热蒸汽中的腐蚀(英文)

用非自耗电弧炉熔炼了Fe/Cr比值为1.75和4.50的Zr(Fe,Cr)_2金属间化合物,它们的粉末经500℃、10.3MPa过热蒸汽腐蚀不同时间后,用X射线衍射、电子探针和透射电子显微镜分析了腐蚀后生成物的结构及其形貌,以及成分的重新分布。Fe/Cr比值不同的Zr(Fe,Cr)_2腐蚀后的生成物都相同,但是含Cr高的更不易被腐蚀。腐蚀初期的生成物是立方ZrO_2,并析出α-Fe(Cr),在继续腐蚀时,立方ZrO_2逐渐转变为单斜ZrO_2,α-Fe(Cr)也逐渐被氧化成(Fe,Cr)_3O_4。Fe和Cr在偏聚时,Fe原子的扩散速率比Cr原子快。根据实验结果,讨论了第二相影响Zr-4合金腐蚀性能的原因。     

铅铋合金热膨胀性能及室温时效研究

铅铋共晶合金(LBE)热膨胀性能影响堆容器的安全性,LBE中杂质的存在会显著影响其热物性,需要制备高纯铅铋合金并对其热膨胀性能进行研究。本文制备出单个杂质元素含量低于10ppm且杂质元素总含量低于73ppm的高纯铅铋共晶合金。热膨胀性能研究结果显示LBE在液相-固相转变时体积增加了约0.80%,室温条件下固态LBE中β相向富铋γ相的转变是其宏观体积变化的最主要原因。     

Corrosion behavior of heat-treated Fe-Al coated steel in lead-bismuth eutectic under loading

A protective surface alloy coating on steel surfaces can prevent steels from the heavy LBE corrosion in the LBE cooled reactor. Especially, Fe-Al alloy coating on a steel surface is effective for corrosion resistance in LBE due to the self-healing of a thin and stable Al oxide layer on the surface of the coating layer. In order to investigate the utility of the coating layer under the stress due to the hydrodynamic and thermal stress induced in the practical system. The Fe-Al coated 316SS, which was heat-treated after the coating process, was immersed in the stagnant LBE at 650 °C for 250 h with loading to investigate the corrosion behavior of the specimen with the bending stress. The Fe-Al coating layer was not corroded because of the protection by Al oxide scale which was formed on the surface of the coating layer and the interface between the coating and the matrix during the heat treatment process. The coating layer cracked elastically. The LBE penetrated into the cracks and corroded the 316SS matrix and the pre-coating layer. The matrix exhibited the dissolution corrosion caused by the preferential dissolution of Ni and the oxidation forming the Fe oxide and Cr oxide. The coating layer is effective to reduce the surface of the matrix to be corroded by LBE, and can moderate the corrosion of the depth direction.     

Corrosion Resistance of Fe-Al-Alloy-Coated Ferritic/Martensitic Steel under Bending Stress in High-Temperature Lead-Bismuth Eutectic

The corrosion test was performed for ferritic/martensitic steel HCM12A with and without Fe-Al alloy coating in LBE at temperatures of 550 and 650°C under loading for an immersion time up to 500 h. After the corrosion test at 550°C for 500 h, both of the uncoated and Fe-Al-coated HCM12A showed a good corrosion resistance without the influence of the tensile stress on the LBE corrosion. On the other hand, after the corrosion test at 650°C, the Fe-Al coating layer on the specimen surface exhibited no LBE dissolution corrosion because of the formation of a stable oxide protection film on the coating layer surface, although the coating layer cracked. The LBE penetrated into the cracks and corroded the base metal and the precoating layer. The uncoated HCM12A exhibited the double oxide layers of FeO and Fe- Cr spinel. The FeO was damaged in the bent zone by the stress, and the Fe-Cr spinel layer was not destroyed by the influence of cracking and the tensile stress. The cracking and stress did not have a large influence on the overall oxidation corrosion rate.     

Corrosion experiments and materials developed for the Japanese HLM systems

The static corrosion tests in lead-bismuth eutectic (LBE) were conducted from 450 °C to 600 °C to understand corrosion behavior and develop corrosion resistant materials for heavy liquid metal systems. While increase of Cr content in steels enhances corrosion resistance in LBE, the effect approaches a constant value above 12 wt% of Cr. Corrosion depth in LBE increases with increasing temperature and corrosion attack becomes severe above 550 °C even under the condition of high oxygen concentration. Nickel dissolution and Pb-Bi penetration occur in 316SS and JPCA above 550 °C under the condition of high oxygen concentration. When oxygen concentration decreases below the level of Fe oxide formation, corrosion attack on these steels also becomes violent due to dissolution of various elements and grain boundary corrosion. Whereas additions of 1.5 wt% Si to T91 and 2.5 wt% Si to 316SS improve corrosion resistance, the effect is insufficient taking fluctuation of oxygen concentration in LBE into consideration. Furthermore, addition of 1.5 wt% Si to T91 causes rise in DBTT. A new coating method using Al, Ti and Fe powders produces corrosion resistant coating layers on 316SS. The coating layers containing 6-8 wt% Al exhibit good corrosion resistance at 550 °C for 3000 h in LBE containing 10⁻⁶-10⁻⁴ wt% of oxygen.     

Corrosion studies in support of lead-bismuth cooled FBRs

The performance of structural materials in lead or lead-bismuth eutectic (LBE) systems is evaluated. The materials evaluated included several US steels (austenitic steel [316L], carbon steels [F-22, Fe-Si], ferritic/martensitic steels [HT-9 and 410]), and several experimental Fe-Si-Cr alloys that were expected to demonstrate corrosion resistance. The materials were exposed in either a dynamic corrosion cell for periods from 100 to 1,000 h at temperatures of 400, 500, 600 and 700°C, depending on material and exposure location. Weight change and optical SEM or X-ray analysis of the specimen were used to characterize oxide film thickness, corrosion depth, microstructure, and composition changes. The tests conducted with stainless steels (410, 316L and HT-9) produced mass transfer of elements (e.g., Ni and Cr) into the LBE, resulting in degradation of the material. With Fe-Si alloys a Si rich layer (as SiO₂) is formed on the surface during exposure to LBE from the selective dissolution of Fe.     

High temperature oxidation of Fe-9Cr-1Mo steel in stagnant liquid lead-bismuth at several temperatures and for different lead contents in the liquid alloy

This research project deals with the feasibility studies concerning the construction of an hybrid reactor for the transmutation of long-lived radioactive wastes. In this context, the liquid lead-bismuth eutectic (LBE) is considered to be a good candidate for the spallation target material needed for the neutrons production necessary to the transmutation. In this hybrid reactor, the LBE, which is enclosed in a T91 (Fe-9%Cr) steel container, can induce corrosion concerns. If the oxygen content dissolved in Pb-Bi is higher than the needed content for magnetite formation, corrosion proceeds by oxidation of the steel. Previously, specific results were reported, obtained in stagnant liquid LBE at 470 °C. An analytical model taking into account the oxide layer structure has been carried out. It involves iron, oxygen and chromium bulk diffusion and diffusion via preferential paths such as liquid lead-bismuth nano-channels incorporated in the oxide layer structure and grain boundaries. In this paper, experimental results on corrosion kinetics, obtained at different temperatures with different percentages of lead in the lead-bismuth alloy, are presented. The model, adapted to the different experimental conditions, is compared to these kinetics and to experimental points coming from the literature at different temperatures in LBE, in pure lead and in pure bismuth.     

奥氏体304NG不锈钢在550℃/25MPa超临界水中的腐蚀行为

研究了304NG不锈钢在550℃/25MPa超临界水中的腐蚀特性。采用扫描电镜、X射线能谱仪和X射线衍射分析了氧化膜的腐蚀形貌、组织结构和元素成分分布。实验结果表明,在550℃/25MPa的超临界水中腐蚀1000h后,304NG不锈钢显示出优越的耐腐蚀性能,其均匀腐蚀增重速率仅为0.01299mg•dm^-2•h^-1。304NG不锈钢在超临界水中形成均匀致密、但带有疖状腐蚀的双层氧化膜,厚度约为2.0μm,内层氧化膜致密而富Cr和Ni,外层氧化膜疏松而富Fe。     

Experience with irradiation of LiSoR test section no. 3

LiSoR (liquid metal–solid metal reaction) loop is a unique facility which was designed to investigate simultaneously the influence of flowing lead bismuth eutectic (LBE), static stress and irradiation by protons onto steel that might be used as structural material in a future ADS (accelerator driven system) reactor. LiSoR is worldwide the first LBE loop that is under operation while it is irradiated at the same time.Up to now five LiSoR test sections have been irradiated. In this paper the experience on LiSoR no. 3 is presented including irradiation experiment and dissembling of the test section and EDM wire cutting of the samples. Additionally the results obtained by SEM and EDX analyses on LiSoR specimen and on beam window are shown whereby the most interesting outcome is the finding of an oxide layer, which was clearly identified by X-ray mapping. This layer is located directly in the irradiated area having a total thickness less than 1 μm. Additionally solidified LBE was detected on some areas of the samples but no wetting of the steel or penetration of LBE into the steel matrix was observed. No enrichment or dissolution of steel elements such as Fe and Cr could be detected on the surface near area. The steel surface is still smooth without any indications of corrosion attack.