纳米及超细晶结构锆金属腐蚀速率计算模型

本文在Wagner理论和固体电子理论的基础上,将电导率与电子的平均自由程及晶粒尺寸的关系联系起来,建立了纳米及超细晶结构金属腐蚀速率一晶粒尺寸关系模型,并结合锆金属的性质,模拟计算了不同温度下晶粒尺寸对锆合金腐蚀速率常数的影响,结果表明,与普通品粒相比,纳米晶粒尺寸下锆Zr的腐蚀速率常数和腐蚀增重,远低于普通晶粒纯Zr的腐蚀速率常数和腐蚀增重,同时,随着纳米晶粒尺寸的减小Zr的腐蚀速率常数也降低,显示晶粒纳米化处理可以改善锆金属的腐蚀性能.     

纳米微晶钛膜结构及其热稳定性

介绍了用离子束辅助沉积法制备纳米微晶钛的活性特点。纳米钛膜在较宽范围温度处理后，用X射线衍射线宽法（谢氏公式）测定出Ti及TiHx晶粒尺寸长大有限。β－Ti纳米微晶氢化后转变成TiHx的现象由X射线衍射（XRD）谱图证实。由氢饱和工艺提供的氢钛原子比印证了TiHxXRD谱线代表的原子比。     

锆—2合金在500—800℃空气中氧化过程的研究

研究了Zr-2合金在500—800℃空气中氧化的规律,利用光学显微镜、透射电镜和扫描电镜对氧化膜形貌、组织和晶体结构进行观察,并对氧化膜内表面应力和应力梯度进行了估算。随着氧化膜增厚,氧化膜内表面应力最初增加而后下降,出现极大值;极大应力值和与其相对应的氧化膜厚度都随氧化温度升高而增加,应力梯度随氧化膜增厚呈单调下降。黑色氧化膜内存在三种晶体结构;单斜、四方、立方,最初晶粒细小,模糊不清,随着氧化时间增长,发生回复再结晶后,晶粒衬度才逐渐变得清晰。白色氧化膜为单一的单斜结构。     

纳米微晶钛膜结构及其热稳定性

介绍了用离子束辅助沉积法制备纳米微晶钛的活性特点.纳米钛膜在较宽范围温度处理后,用X射线衍射线宽法(谢氏公式)测定出Ti及TiHx晶粒尺寸长大有限.β-Ti纳米微晶氢化后转变成TiHx的现象由X射线衍射(XRD)谱图证实.由氢饱和工艺提供的氢钛原子比印证了TiHx XRD谱线代表的原子比.     

铀铌合金表面热氧化膜结构研究

采用X射线光电谱分析技术、结合Ar离子枪溅射,研究了大气环境、不同温度(室温、100、200和300℃)条件下铀铌合金样品表面氧化膜结构的变化情况。实验结果表明:随温度升高,氧化膜厚度增大,氧化膜结构发生明显变化。不同温度热氧化处理后,铀铌合金初始表面Nb主要以Nb2O5形式存在,在Nb2O5与金属Nb之间,总存在一定厚度的NbO及少量其他价态氧化物NbOx(0x1,1x2)的混合层。室温～200℃热氧化合金样品表面铀均以含间隙氧的UO2+x(P型)存在,其U4f7/2结合能较UO2低约0.7eV。室温条件下,氧化膜成分主要为UO2;100、200℃热氧化后,氧化膜中除UO2外,还含有少量P型UO2+x,其U4f5/2卫星峰的结合能为396.6eV。300℃热氧化后的合金样品表面为铀的高价氧化物(U3O8或UOx,2x3),U4f特征峰的结合能分别为381.8和392.2eV;氧化层为UO2和金属铀的混合物。热氧化过程中,温度对铀氧化的影响较对Nb的明显得多。     

锆合金氧化膜弹性模量的测定

氧化膜的内应力对锆合金的腐蚀行为有重要影响,弹性模量是计算分析氧化膜内应力的关键物性参数。由于测试困难,锆合金氧化膜的弹性模量通常根据块体材料的数据进行估计。本文利用纳米压痕法测试分析了多种条件下锆合金氧化膜的弹性模量和硬度,研究发现其表面和截面的弹性模量与硬度数值存在差异。与纯水相比,在含锂水介质中腐蚀,锆合金氧化膜的弹性模量及硬度偏小。     

U-2Nb合金热处理后的相结构和成分分析

观察了三种热处理工艺下U 2Nb合金的组织 ,分析了其微观结构和相成分。精确测定了U 2Nb合金中α相和γ相的点阵常数计算出了两相中的Nb含量。在测量中发现 ,机加工后经电解抛光改变了X射线衍射图 ,分析了该异常变化的原因 ,其结果为鉴定U 2Nb合金的双相结构找到了一种有效方法     

XAFS方法研究纳米非晶态NiB和NiP合金的结构

不同退火温度处理后的纳米非晶态NiB和NiP合金催化剂XAFS和XRD的结果表明,在300℃温度退火后,纳米非晶态NiB合金晶化生成纳米晶Ni和晶态Ni3B中间态;纳米非晶态NiP合金直接晶化生成稳定的晶态Ni和Ni3P。在500℃温度退火后,NiB和NiP样品都晶化为金属Ni,但NiB样品中的Ni原子周围的局域结构与金属Ni箔的几乎相同,而NiP样品由于Ni原于受到元素P的影响,生成的晶态Ni的结构有较大的畸变,结构与金属Ni相差很大。     

锆-4合金在高压釜中腐蚀时氧化膜显微组织的演化

用透射电镜．扫描电镜和扫描探针显微镜研究了锆-4样品在高压釜中（360℃／18．6MPa去离子水）腐蚀后氧化膜的断口形貌、氧化膜不同深度处的显微组织和晶体结构。结果表明：氧化膜生成时形成的压应力。使晶体中产生了许多缺陷,稳定了一些亚稳相．在氧化膜底层中有单斜,四方、立方晶体结构甚至非晶相存在;在氧化膜的中间层及表面层中。空位和间隙原子等缺陷发生扩散、湮没和凝聚,内应力发生弛豫,亚稳相转变成稳定的单斜结构;空位被晶界吸收形成了纳米尺寸的孔洞簇,弱化了晶粒间的结合力,在表面张力的作用下,晶粒逐渐成为球形;孔洞簇的形成并发展成裂纹,使氧化膜失去了原有良好的保护性。导致腐蚀转折,这是氧化膜的显微组织在腐蚀过程中发生演化后的必然结果。     

熔化情况下Zr_(-4)合金包壳管氧化膜厚度变化的计算

已有的研究表明,温度超过元件包壳熔点时,Zr-4合金与蒸汽在包壳表层反应所生成的ZrO_2氧化膜将对Zr-4合金熔融物本身起到一定的保护作用。因此这层氧化膜的厚度及完整性对于由堆芯元件过热及由此引起的Zr-4合金熔融物流动所导致的反应堆事故具有重要的影响。本文假定氧的扩散是ZrO_2氧化膜厚度变化的控制因素,通过解扩散方程,导出了对在这种情况下ZrO_2膜厚度变化的计算公式。     

Embrittlement of Zircaloy-4 due to Oxidation in Environment of Stagnant Steam

With a view to examining the embrittlement behavior of Zircaloy due to inner surface oxidation occurring in an LWR loss-of-coolant accident, ring-like Zircaloy-4 cladding specimens were heated at the isothermal oxidation temperature ranging 890~1,194°C in an environment of stagnant steam, which simulated the atmospheric condition inside the ruptured cladding.

The embrittlement of the specimen due to oxidation in an environment of stagnant steam is influenced primarily by the amount of hydrogen absorbed by the Zircaloy-4. Ring compression tests conducted at 100°C on oxidized ring-like cladding specimen showed that Zircaloy containing more than about 500wt.ppm of hydrogen had become brittle.

The results of the present experiment support the idea supposed in the previous paper that the hydrogen absorption on the inner surface of the ruptured cladding, causing severe embrittlement, is strongly related with the atmospheric condition inside the ruptured cladding.     

水化学对Zr-4合金氧化膜形貌的影响

为了研究LiOH水溶液影响Zr-4合金耐腐蚀性能的机理,用扫描电子显微镜和扫描探针显微镜观察Zr-4合金样品在不同介质中腐蚀后氧化膜内外表面的形貌。实验结果表明：由于氧化过程体积膨胀,氧化膜中存在平行于界面的压应力。在压应力的作用下,氧化膜向外鼓起,并在垂直于平面的方向上产生张应力,导致氧化膜破裂,从而使腐蚀加速。LiOH水溶液浓度越高,这个过程越快。     

氧化-氢化引起的锆合金焊接件开裂问题

3块Zr-4板经过800℃真空热轧扩散焊接的样品,在400℃过热蒸汽中腐蚀150天后会发生重新开裂,用光学显微镜观察样品截面上显微组织的变化;用微区成分分析方法研究了焊接面处与基体成分之间的差别.结果表明:焊接面上C元素的污染引起焊接面耐腐蚀性能的变坏,以及氧化-氢化联合作用导致了焊接面的开裂.     

The Effect of Oxide Microstructure on Kinetic Transition in Out-of-pile Steam Corrosion Test for Zircaloy-2 and Nb-added Zircaloy-2

In order to study the mechanism of kinetic transition of corrosion rate for zirconium alloys, oxide films formed on Zircaloy-2 (Zry-2) and Nb-added Zircaloy-2 (0.5Nb/Zry-2) in steam at 673 K and 10.3 MPa were examined with TEM and SIMS.

Kinetic transition occurred at almost the same oxide thicknesses for both Zry-2 and 0.5Nb/Zry-2, but the corrosion rate after the transitions were quite different for the two alloys. Zircaloy-2 showed cyclical oxidation, while the weight gain of 0.5Nb/Zry-2 increased linearly.

The morphology and crystal structure were similar for the oxides of the two alloys and both the oxide films still mainly consisted of columnar grains even after the transition. Interface layers which mainly consisted of a-Zr crystallites were observed for both alloys and the oxygen content in the interface layers increased after the transition.

The solute concentrations of Fe, Cr and Ni became higher, accompanying the increase of oxygen concentrations at columnar grain boundaries in the oxide films after the transition for 0.5Nb/Zry-2. It was thought that the properties of grain boundaries of the 0.5Nb/Zry-2 oxide films changed after the transition, and the increase in oxygen diffusivity at grain boundaries caused the linear increase in weight gain.     

Irradiation Effects on Corrosion Resistance and Microstructure of Zircaloy-4

In order to investigate irradiation effects on nodular corrosion resistance of Zircaloy-4, an out-of-pile corrosion test was conducted using Zircaloy-4 specimens cut from the channel box of a fuel assembly irradiated in the BWR (Monticello reactor) up to the neutron fluence of 1.53×10²⁶ n/m² (E>1MeV). The corrosion test was carried out in high pressure steam of 10.3 MPa at 783 K for 24 h. No nodules appeared on the specimens cut from welded areas of the channel box and nodular corrosion resistance tended to be better with increasing neutron fluence. Microstructural evolution in the form of irradiation-induced release of Fe atoms from Zr(Fe, Cr)2 type precipitates was detected by an analytical electron microscope. It was found that the higher the concentration of dissolved Fe and Cr in the grains of Zircaloy-4, the better the nodular corrosion resistance.     

Ar+离子辐照对锆4合金耐腐蚀性能的影响

为了研究辐照损伤对锆4合金电化学耐腐蚀性能的影响,使用直线加速器产生单电荷载能Ar+,在液氮温度下辐照样品表面,产生缺陷.然后,测量辐照后锆4合金的电化学极化曲线,使用钝化电流密度作为评定腐蚀性能的指标,分析不同注量Ar+离子辐照对锆4合金钝化电流密度的影响.同时使用透射电子显微镜分析不同注量Ar+辐照下锆4合金损伤层的微观结构.实验结果表明在低离子注量范围内(＜3×1014/cm2),随着辐照量的增加钝化电流密度升高,耐腐蚀性能降低;在中等离子注量范围内(3×1014-1×1016/cm2),随着辐照量的增加钝化电流密度急速下降,耐腐蚀性能显著提高;在高离子注量范围内(1×1016-1×1017/cm2),随着辐照量的增加钝化电流密度又开始增加,耐腐蚀性能再次降低.最后,根据原子碰撞理论对实验结果进行了理论分析.     

Zircaloy-4 Cladding Embrittlement due to Inner Surface Oxidation under Simulated Loss-of-Coolant Condition

Embrittlement of Zircaloy-4 cladding by oxidation of the inner surface occurring in an LWR loss-of-coolant accident was studied using simulated fuel containing of A1₂O₃ pellets sheathed in Zircaloy-4 specimen cladding, filled with Ar gas, and sealed. This simulated fuel rod was heated from outside until the isothermal oxidation temperature between 880 and 1,167°C was obtained after the cladding burst. This exposed the inner surface of the cladding to the environmental atmosphere, provided by steam flowing at a constant rate in the range of 0.13–1.6 g/cm²-min.

The embrittlement of the specimen due to inner surface oxidation is influenced primarily by the amount of hydrogen absorbed by the Zircaloy-4. Ring compression tests conducted at 100°C on test pieces constituted of sliced sections of oxidized specimen showed that Zircaloy containing more than 200–300 wt.ppm of absorbed hydrogen became brittle when oxidized at temperatures above 1,000°C. In the range of oxidation temperature 932 to 972°C, brittleness did not appear below 500–750 wt.ppm absorbed hydrogen.

Hydrogen absorbed by the Zircaloy precipitated in the form of fine hydride crystals formed along previous β-phase grain boundaries. Peaks were found in the distribution of hydrogen absorbed on the inner surface, at a distance of 15–45 mm upward and downward of the rupture opening. Within this range, the distance was influenced by the oxidation temperature and steam flow rate.     

Evolution of texture during creep of Zircaloy-4 at 673 K

Measurements of the evolution of texture during creep of Zircaloy-4, at 673 K, are reported. The changes in the (0002) and {101̄0} pole figures with creep strain are interpreted in terms of a contributionof $\text{〈c + a〉}$-slip on {101̄0}, {101̄1}, {112̄1} and {112̄2} planes, in addition to the usual $\text{〈a〉}$-slip on prismatic planes.