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。     

Heat Transfer Characteristics of Zircaloy-2 Oxide Film

A neural network model is under development to predict the local power distribution in a BWR fuel bundle as a high speed simulator of precise nuclear physical analysis model. The relation between ²³⁵U enrichment of fuel rods and local peaking factor (LPF) has been learned using a two-layered neural network model ENET. The training signals used were 33 patterns having considered a line symmetry of a 8x8 assembly lattice including 4 water rods.

The ENET model is used in the first stage and a new model GNET which learns the change of LPFs caused by burnable neutron absorber Gadolinia, is added to the ENET in the second stage. Using this two-staged model EGNET, total number of training signals can be decreased to 99. These training signals are for zero-burnup cases. The effect of Gadolinia on LPF has a large nonliniality and the GNET should have three layers. This combined model of EGNET can predict the training signals within 0.02 of LPF error, and the LPF of a high power rod is predictable within 0.03 error for Gadolinia rod distributions different from the training signals when the number of Gadolinia rods is less than 10. The computing speed of EGNET is more than 100 times faster than that of a precise nuclear analysis model, and EGNET is suitable for scoping survey analysis.     

Hydrogen generation by water radiolysis with immersion of oxidation products of Zircaloy-4

In order to predict the hydrogen gas generation from seawater or water in which debris would be included by the severe accident of nuclear power plant, we investigated the effect of ZrO₂ and the oxidation products of Zircaloy-4 on hydrogen gas generation by radiolysis of water since the radiolytic generation could be affected by materials immersed in water. Powders of well-characterized oxides and oxidation products were immersed in either seawater or distilled water, and irradiated by gamma ray from a Co-60 source. The observed hydrogen yield, G(H₂), was measured as a function of the weight fraction of oxide in water up to 50 wt%. The enhancement of the hydrogen generation by radiolysis of water with the commercial oxides and the oxidation products of Zircaloy-4 was quite small or absent in seawater. But the enhancement was observed in the presence of the oxides or the oxidation products at low weight fraction in distilled water. This enhancement in distilled water seemed to be dependent on specific surface area or particle size, but its dependence on the crystal structure was not apparent in the experimental results. The enhancement was saturated at higher ZrO₂ weight fractions and it was not apparent in the seawater.     

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.     

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.     

Fracture behavior of recrystallized and stress-relieved Zircaloy-4 cladding under biaxial stress conditions

Pellet-cladding mechanical interaction (PCMI) under reactivity-initiated accident conditions may lead to failure of high-burnup fuel rods. The biaxial stress states of Zircaloy cladding tubes induced by PCMI loading presumably makes the tubes more susceptible to failure. To clarify the influence of the anisotropic mechanical properties of Zircaloy cladding tubes on their fracture behavior under biaxial stress conditions, biaxial tensile tests were performed, and the measured stresses and strains were converted to reduced parameters such as equivalent strain, equivalent stress, and stress triaxiality by using the anisotropic constants of the Hill yield function derived in our previous study. The minimum fracture strains of recrystallized (RX) and stress-relieved (SR) specimens were located where the stress ratio of axial to circumferential is 0.75. The equivalent plastic fracture strains tended to decrease monotonously with increasing stress triaxiality, which is a typical trend observed in ductile fracture, in the range of 0.65–0.78 for both specimens. In the case of SR specimens, however, the analysis with stress triaxiality did not reduce the fracture strains well to a single trend curve, showing that another influential factor is required such as an anisotropic fracture behavior or a plastic behavior deviated from that predicted by the yield function.     

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。     

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.     

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.