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。     

热处理对锆-4合金中第二相结构和成分的影响(英文)

用电化学方法分离出锆-4合金中第二相,研究了不同热处理制度对第二相结构和成分的影响。锆-4合金经1050℃β相加热空冷后,析出的第二相为立方结构的Zr(Fe,Cr)_2,Fe/Cr比值在2.1～2.5之间。试样重新在600～800℃下加热3h,晶体结构不发生改变,只是Fe/Cr比值逐渐降至1.9;但在700～800℃下加热后,有少量的六方结构Zr(Fe,Cr)_2第二相析出。生产厂提供的锆-4板中第二相是六方结构的Zr(Fe,Cr)_2,重新在700～800℃加热3h,晶体结构不发生变化,Fe/Cr比值由1.9降至1.5左右。这说明在重新加热时,第二相中的Fe和Cr与周围基体中的Fe和Cr会相互扩散置换。试样从β相冷却析出第二相时,Fe原子的扩散比Cr原子快;Cr原子在六方晶格Zr(Fe,Cr)_2中的固溶度比在立方晶格Zr(Fe,Cr)_2中的大。由于这些原因造成了第二相成分随热处理制度不同而变化的现象。     

热处理对锆-4合金中第二相结构和成分的影响

用电化学方法分离出锆-4合金中第二相,研究了不同热处理制度对第二相结构和成分的影响。锆-4合金经1050℃β相加热空冷后,析出的第二相为立方结构的Zr(Fe,Cr)_2,Fe/Cr比值在2.1～2.5之间。试样重新在600～800℃下加热3h,晶体结构不发生改变,只是Fe/Cr比值逐渐降至1.9;但在700～800℃下加热后,有少量的六方结构Zr(Fe,Cr)_2第     

Zr-Fe-Nb合金中Zr(Nb,Fe)2相在400 ℃/10.3 MPa过热蒸汽中的腐蚀行为

Zr-1.0Fe-1.0Nb合金经β相油淬、冷轧变形及580 ℃/5 h退火处理，在静态高压釜中进行400 ℃/10.3 MPa过热蒸汽腐蚀试验，利用带EDS的SEM和HRTEM对合金基体以及腐蚀生成的氧化膜显微组织进行分析。结果表明：合金中主要存在正交的Zr 3Fe和密排六方的Zr(Nb，Fe)₂第二相。Zr(Nb，Fe)₂相在氧化过程中先转变成非晶组织，非晶进一步氧化转化为m-Nb₂O₅和m-Fe₂O₃相纳米晶态氧化物，最后扩散流失到腐蚀介质中；Zr(Nb，Fe)₂相氧化后的Fe、Nb元素发生扩散流失，且Nb的流失速度大于Fe，合金元素的扩散流失在氧化膜中留下大量缺陷，促进氧化膜由柱状晶向等轴晶形态演化而不利于合金的耐腐蚀。     

Zr-4合金α-Zr固溶体中的Fe、Cr含量分析

将电化学分离Zr-4合金中Zr(Fe,Cr)2第二相粒子的技术和原子吸收光谱分析技术相结合，建立了一种分析Zr-4合金α-Zr固溶体中Fe,Cr含量的新方法，并用这种新方法分析了不同热处理状态下Zr-4合金α-Zr固溶体中的Fe,Cr含量。分析结果表明，随着淬火温度的增加。α-Zr固溶体中的Fe,Cr含量和Fe/Cr比值均增加，而Zr(Fe,Cr)2第二相粒子中的Fe/Cr比值则相应降低。结合以前的工作，可得结论：过饱和固溶在α-Zr固溶体中的Fe,Cr含量对Zr-4合金的耐水侧腐蚀性能有重要影响。     

Zr-4合金a-Zr固溶体中的Fe、Cr含量分析

将电化学分离Zr-4合金中Zr(Fe,Cr)2第二相粒子的技术和原子吸收光谱分析技术相结合,建立了一种分析Zr-4合金a-Zr固溶体中Fe、Cr含量的新方法,并用这种新方法分析了不同热处理状态下Zr-4合金a-Zr固溶体中的Fe、Cr含量。分析结果表明,随着淬火温度的增加,a-Zr固溶体中的Fe、Cr含量和Fe/Cr比值均增加,而Zr(Fe,Cr)2第二相粒子中的Fe/Cr比值则相应降低。结合以前的工作,可得结论:过饱和固溶在a-Zr固溶体中的Fe、Cr含量对Zr-4合金的耐水侧腐蚀性能有重要影响。     

Zr-2/Cr扩散反应层物相分析

研究Zr-2/Cr扩散反应层物相，可为判断Zr-2和Cr是否相容提供依据。用热压法（50MPa)获得在1073K时Zr-2/Cr扩散反应层。分别用透射电镜（TEM）和场发射扫描电镜配备的薄窗能谱仪（EDS）对反应层进行结构的成分分析。结果表明，Zr-2/Cr扩散反应生成六方结构（C14型的）Zr(Fe,Cr)2Laves相。     

Zr(Fe,Cr)_2金属间化合物在500℃过热蒸汽中的腐蚀

用非白耗电弧炉熔炼了Fe/Cr比值为1.75和4.50的Zr(Fe,Cr):金属间化合物,它们的粉末经500℃、10.3 MPa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     

Zr-2/Cr反应扩散研究

研究了Zr-2/Cr扩散偶在1023－1123K范围内的反应扩散，用扫描电镜观察反应宽度，用能谱仪（EDS）测定了反应层Zr,Fe和Cr沿扩散方向的浓度分布，研究了反应层的生长动力学，结果表明，反应层生长基本符合抛物经规律，生长过程受扩散控制，用Boltzman-Matano-Heumann模型计算了Cr在反应层Zr（Fe,Cr）2中的互扩散系数D，得到了互扩散系数D与温度的Arrhenius方程，由计算的扩散数据与Cr在Zr中形成稀固溶体时Cr的扩散数据比较可知，Cr在金属间化合物Zr（Fe,Cr）2中的扩散比在稀因溶体中的扩散快5个数量级。     

Zr-4合金α-Zr固溶体中Fe,Cr含量的研究

将电化学分离Zr-4合金中Zr(Fe,Cr)2第二相粒子的技术和原子吸收光谱分析技术相结合,建立了一种分析Zr-4合金α-zr固溶体中Fe和Cr含量的新方法,并用这种新方法分析了不同热处理状态下Zr-4合金α-zr固溶体中的Fe和Cr含量。分析结果表明,随着淬火温度的增加, α-zr固溶体中的Fe,Cr含量和Fe/Cr比值均增加,而Zr(Fe,Cr,)2第二相粒子中的Fe/Cr比值则相应降低,并且在淬火状态,α-zr固溶体中Fe,Cr的含量均大于其最大饱和固溶度。结合以前在高温水、蒸汽中的腐蚀试验结果发现,α-zr固溶体中的Fe,Cr含量对Zr-4合金的耐水侧腐蚀性能起着重要作用,并且,过饱和固溶在α-zr中的Fe,Cr含量存在一临界值,当Fe和Cr含量大于临界值时,Zr-4合金就具备抗疖状腐蚀的能力。     

Characterization of solidified melt among materials of UO2 fuel and B4C control blade

To predict the fundamental phase relationships in the solidified core melt of the Fukushima Daiichi Nuclear Power Plant, solidified melt samples of the various core materials [B₄C, stainless steel, Zr, ZrO₂, (U,Zr)O₂] were prepared by arc melting. Phases and compositions in the samples were determined by means of X-ray diffraction, microscopy, and elemental analysis. With various compositions, the only oxide phase formed was (U,Zr)O₂. After annealing, the stable metallic phases were an Fe-Cr-Ni alloy and an Fe₂Zr-type (Fe,Cr,Ni)₂(Zr,U) intermetallic compound. The borides, ZrB₂ and Fe₂B-type (Fe,Cr,Ni)₂B, were solidified in the metallic part. Annealing at 1773 K under an oxidizing atmosphere (Ar-0.1%O₂) resulted in the oxidation of U and Zr in the alloy and in ZrB₂, and consequently the (Fe,Cr,Ni)₂B and Fe-Cr-Ni alloy became dominant in the metallic part. The experimental phase relationships in the metallic part agreed reasonably with the thermodynamic evaluation of equilibrium phases in a simplified B₄C–Fe–Zr system. The metallic Zr content in the melt was found to be a key factor in determining the phase relationships. As a basic mechanical property, the microhardness of each phase was measured. The borides, especially ZrB₂, showed notably higher hardness than any other oxide or metallic phases.     

Metallurgical characterizations of Fe–Cr–Ni–Zr base alloys developed for geological disposal of radioactive hulls

Alloy melting route is currently being considered for radioactive hulls immobilization. Towards this, wide range of alloys, belonging to Zirconium–Iron binary and Zirconium–Stainless steel pseudo-binary systems have been prepared through vacuum arc melting route. Detail microstructural characterization and quantitative phase analyses of these alloys along with interaction study between Zirconium and Stainless steel coupons at elevated temperatures identify Zr(Fe,Cr)₂, Zr(Fe,Cr), Zr₂(Fe,Cr), Zr₃(Fe,Ni), Zr₃(Fe,Cr), Zr₃(Fe,Cr,Ni), β-Zr and α-Zr as the most commonly occurring phases within the system for Zirconium rich bulk compositions. Nano-indentation studies found Zr(Fe,Cr)₂ and Zr(Fe,Cr) as extremely hard, Zr₃(Fe,Ni) as moderately ductile and β-Zr, Zr₂(Fe,Cr) as most ductile ones among the phases present. Steam oxidation studies of the alloys, based on weight gain/loss procedure and microstructural characterization of the mixed oxide layers, suggest that each of the alloys responded to the corrosive environment differently. Fe₂O₃, NiFe₂O₄, NiO, monoclinic ZrO₂ and tetragonal ZrO₂ are found to be most common constituents of the oxide layers developed on the alloys. Integrating the microstructural, mechanical and corrosion properties, ZrFeCrNi3 (Zr: 84.00, Fe: 11.20, Cr: 3.20, Ni: 1.60, in wt.%) is identified as the acceptable base alloy for disposal of radioactive hulls.     

Reactions between U–Zr alloys and Fe at 923 K

Interdiffusion experiments were carried out at 923 K with the diffusion couples consisting of U–23 at.% Zr/Fe and U–23 at.% Zr–1 at.% Ce/Fe. The reaction layer adjacent to the Fe was a single Zr-depleted UFe₂ phase. The phases in the reaction layers were estimated consistently with the calculated U–Zr–Fe ternary isotherm. The diffusion path obtained in this study was similar to that reported for the U–Pu–Zr/HT9-steel couple at 923 K, when those paths were expressed on the (U+Pu)–Zr–(Fe+Cr) composition triangle. The reaction layers grew in proportion to the square root of the annealing time. The addition of approximately 1 at.% of Ce to the U–23 at.% Zr alloy has little effect on the reaction between U–23 at.% Zr and Fe.     

The effect of alloying modifications on hydrogen uptake of zirconium-alloy welding specimens during corrosion tests

The hydrogen uptake behavior during corrosion tests for electron beam welding specimens made out of Zircaloy-4 and zirconium alloys with different compositions was investigated. Results showed that the hydrogen uptake in the specimens after corrosion tests increased with increasing Cr content in the molten zone. This indicated that Cr element significantly affected the hydrogen uptake behavior. Fe and Cr have a low solubility in α-Zr and exist mainly in the form of Zr(Fe,Cr)₂ precipitates, which is extremely reactive with hydrogen in its metallic state. It is concluded that the presence of Zr(Fe,Cr)₂ second phase particles (SPPs) is responsible for the increase in the amount of hydrogen uptake in the molten zone of the welding samples after corrosion, as Zr(Fe,Cr)₂ SPPs embedded in α-Zr matrix and exposed at the metal/oxide interface could act as a preferred path for hydrogen uptake.     

Irradiated Behavior at High Burnup for HiFi Alloy

Irradiation tests of a BWR advanced Zr alloy (HiFi alloy) and Zircaloy-2 (Zry-2) were carried out in a Japanese commercial reactor and the irradiation performances of the materials were investigated. HiFi alloy and Zry-2 showed excellent resistance to corrosion up to 70 GWd/t, and furthermore, HiFi kept lower hydrogen pickup compared with Zry-2. TEM observation showed that the Fe/(Fe+Cr) ratio of Zr(Fe,Cr)2 type second phase particles (SPPs) for HiFi alloy and Zry-2 tended to decrease as fast neutron fluence increased and to saturate at high fluence. Zr-Fe-Cr SPPs did not completely disappear even for 6 cycles for the irradiated HiFi alloy and Zry-2. In order to clarify the mechanism of hydrogen absorption, an electrochemical technique was used for the oxide film of both materials as part of the out-of-pile test. The relation between the oxide surface potential and the hydrogen pickup fraction was estimated suggesting that the potential difference over the oxide film suppressed hydrogen (proton) diffusion in the oxide film.     

Atomistic modeling of the interaction of cladding elements (Fe, Ni, Cr) with U-Zr fuel

Atomistic simulations of U-Zr fuel and its interaction with Fe, Ni, and Cr using the BFS method for alloys are presented. Results for the γU-βZr solid solution are discussed, including the behavior of the lattice parameter and coefficient of thermal expansion as a function of concentration and temperature. Output from these calculations is used to study the surface structure of γU-βZr for different crystallographic orientations, determining the concentration profiles, surface energy, and segregation behavior. The analysis is completed with simulations of the deposition of Fe, Ni and Cr on U-Zr substrates with varying Zr concentration. All results are discussed and interpreted by means of the concepts of strain and chemical energy underlying the BFS method, thus obtaining a simple explanation for the observed Zr segregation and its influence in allowing for cladding elements diffusion into the U-Zr fuel.     

Effects of Fe and Cr on corrosion behavior of ZrFeCr alloys in 500 °C steam

A study of the corrosion behaviors of ZrFeCr alloy and the influence of microstructure on corrosion resistance are described by X-ray diffraction and scanning electron microscope in this paper. The results show that several ZrFeCr alloys exhibit protective behavior throughout the test and oxide growth is stable and protective. The best alloy has the composition Zr1.0Fe0.6Cr. Fitting of the weight gain curves for the protective oxide alloys in the region of protective behavior, it showed nearly cubic behavior for the most protective alloys. The Zr1.0Fe0.6Cr has the more laves Zr(Fe,Cr)₂ precipitate in matrix and it has the better corrosion resistance. The Zr0.2Fe0.1Cr has little precipitate, the biggest hydrogen absorption and the worst corrosion resistance. The number of precipitates and the amount of hydrogen absorption in Zr alloy plays an important role on corrosion resistance behaviors in 500 °C/10.3 MPa steam.