Zr-Nb合金在LiOH水溶液中耐腐蚀性的研究

比较了Zr-Nb合金与Zr-Sn-Nb-Fe及Zr-4合金的耐腐蚀性，讨论了Nb对Zr-Nb合金在不同介质中耐腐蚀性的影响，实验证明Nb的腐蚀产物可以部分溶解于LiOH水溶液中，因而Zr-Nb合金在LiOH水溶液中腐蚀时，合金中β－Nb第二相腐蚀后生成的腐蚀产物会部分溶解于LiOH水溶液，在氧化膜中形成孔洞，从而导致Zr-Nb合金在LiOH水溶液中的耐腐蚀性变差。     

两种锆合金显微组织与腐蚀机理研究

根据国产C锆合金与低锡Zr-4合金在纯水以及LiOH水溶液中的高压釜腐蚀试验的结果,采用透射电镜（TEM）观察基体和氧化膜显微组织,通过分析氧化增重数据,对C锆合金的腐蚀机理进行了研究。提出了3种腐蚀机理:即Nb元素有效抑制阴离子空位浓度提高,可减少氧元素扩散速率;缺陷阱的数量影响氧扩散带来的腐蚀,且空位阱数量与第二相颗粒总表面积成正比;第二相粒子氧化膨胀造成氧化膜压应力松弛,降低其稳定性并失去保护能力。      

Zr-4合金在 LiOH水溶液中腐蚀机理的概述

综述了近十几年国内外对Zr-4合金在LiOH水溶液中腐蚀的研究,着重讨论了在LiOH水溶液中加速腐蚀的规律和机理,并分析了目前解释这种腐蚀机理的各种假说中存在的问题。     

ZIRLO合金和Zr-4合金在LiOH水溶液中耐腐蚀性能的研究

比较了ZIRLO合金和Zr-4合金两种样品在350℃、16．8MPa、0．04MLiOH水溶液中的耐腐蚀性能，发现Zr4合金样品在腐蚀转折之前的腐蚀增重比ZIRLO合金稍低，这时两种样品的氧化膜相对完整而致密。用二次离子质谱仪(SIMS)测量Li^ 在两种合金样品氧化膜剖面中的分布，发现Li^ 进入Zr-4合金氧化膜的深度比ZIRLO合金浅，但浓度比较高。而腐蚀至68天在Zr—4合金样品腐蚀发生转折后，其腐蚀增重远高于ZIRLO合金，这是因为此时Zr-4合金样品氧化膜因疏松而失去保护作用，而ZIRLO合金样品腐蚀至82天氧化膜仍致密而完整。ZIRLO合金中细小的βNb和Zr—Fe—Nb第二相粒子可能对保持氧化膜的完整性有重要作用。     

扫描探针显微镜在锆合金氧化膜显微组织研究中的应用

Zr-4合金在高压釜中经360 ℃高温水腐蚀后,用扫描探针显微镜(SPM)研究了氧化膜中的显微组织和晶粒形貌.由于SPM具有很高的水平和垂直分辨率,适合于观察表面只有微小起伏的显微组织,所以能够清晰地观察到氧化膜中的裂纹、空洞和晶粒等显微组织.测试样品的制备方法简便.     

Zr-0.2Cu-xNb合金的显微组织及腐蚀行为

对Zr-0.2Cu-xNb（质量分数x=0.2,0.5,1.0,2.5）合金进行真空β相油淬、冷轧及退火处理,并在静态高压釜中进行过热蒸汽腐蚀试验,最后采用扫描电镜和透射电镜研究了合金及其腐蚀生成的氧化膜的显微组织。结果表明,随着Nb含量的增加,Zr-0.2Cu-xNb合金中Zr2Cu第二相的数量逐渐减少,而β-Zr第二相数量逐渐增加;合金中尺寸较小的Zr2Cu第二相对耐腐蚀性能有利;β-Zr第二相在氧化过程中会促进氧化膜微裂纹的产生,降低合金的耐腐蚀性能。Zr-0.2Cu-xNb合金中Nb含量接近其在α-Zr中最大固溶度时,合金具有最优的耐腐蚀性能。      

国产新型锆铌合金水侧腐蚀行为研究

对国产新型锆铌合金进行了元件表面带有热负荷情况下的堆外动水腐蚀实验,同时进行500℃蒸汽腐蚀实验及在氢氧化锂和硼酸水中的静水腐蚀实验,获得了不同腐蚀实验条件下样品的增重或氧化膜厚度,并与改进Zr-4的数据进行了比较.利用光学显微镜(OM)对腐蚀形成的氧化膜进行了分析,采用惰气脉冲红外法测量了样品的氢含量,并用OM观察了基体中氢化物的形貌和分布.实验结果表明,国产新型锆铌合金的抗腐蚀性能优于改进Zr-4,而新型锆铌合金中细小均匀分布的第二相粒子是其具有优异抗腐蚀性能的原因.     

秦山一期核电厂乏燃料棒包壳外表面氧化膜内应力研究

锆合金在压水堆中常用作包壳材料,而包壳的腐蚀限制了燃料的堆内使用寿命。为了增加核燃料的燃耗,有必要研究包壳的腐蚀过程。以秦山一期核电厂乏燃料棒为研究对象,对包壳外表面氧化膜的内应力和物相组成进行分析,对径向氧化膜显微形貌进行观察。结果表明,氧化膜中存在压应力,从燃料棒底端到顶端,应力逐渐减小,当降低到最低值时,应力逐渐稳定下来,最后在气腔处又突然增加;压应力对稳定四方相有着非常重要的作用,随着氧化层中裂纹与孔洞的发展,应力得到释放,氧化膜的物相逐渐转变为单斜相。     

HRTEM原位加热研究Zr-4合金氧化过程中的晶格演化

采用高分辨透射电镜(HRTEM)原位氧化的方法,研究了Zr-4合金腐蚀初期的行为,为认识Zr-4合金在氧化膜与金属基体界面的氧化行为提供参考。原位氧化实验结果显示ZrO_2的形核过程伴随着连续的晶格演化:ZrO_2形成前,氧原子不断固溶到α-Zr的晶格间隙中,形成若干不连续的富氧区,随着氧含量的增加,富氧区晶格之间产生较大应变促使晶格出现调制以及形成面心立方结构ZrO亚氧化物,亚氧化物由取向接近的畴结构组成,随着反应的继续晶格连续演变直到形成m-ZrO_2。     

锆合金的碘致应力腐蚀开裂研究(英文)

研究了不同热处理状态的Zr-2和Zr-4合金在不同浓度的碘介质及实验温度下的应力腐蚀开裂(SCC)行为。并对不同织构取向的试样在350℃下进行了蠕变实验,蠕变实验的载荷值选择与SCC实验相对应的一系列典型载荷。用扫描电子显微镜观察了断口特征,用透射电子显微镜和光学显微镜检查了材料的显微组织,用X-光衍射仪测定了锆合金的织构,分析讨论了材料状态、实验温度、碘浓度以及蠕变对锆合金碘致应力腐蚀行为的影响。     

Corrosion characteristics and oxide microstructures of Zircaloy-4 in aqueous alkali hydroxide solutions

The corrosion characteristics of Zircaloy-4 have been investigated in various aqueous solutions of LiOH, NaOH, KOH, RbOH and CsOH with equimolar M⁺ and OH⁻ at 350°C. The characterization of the oxides was performed using transmission electron microscope (TEM) and scanning electron microscope (SEM) on the samples which were prepared to have an equal oxide thickness in pre-transition and post-transition regimes. At a low concentration (4.3 mmol) of aqueous alkali hydroxide solutions, the corrosion rates decrease gradually as the ionic radius of cation increases. At a high concentration (32.5 mmol), the corrosion rate increases significantly in LiOH solution and slightly in NaOH solution, but in the other hydroxide solutions such as KOH, RbOH and CsOH, the corrosion rate is not accelerated. Even if the specimens have an equal oxide thickness in LiOH, NaOH and KOH solutions, the oxide microstructure formed in the LiOH solution is quite different from those formed in the NaOH or the KOH solutions. In the LiOH solution, the oxides grown in the pre-transition regime as well as in the post-transition regime have an equiaxed structure including many pores and open grain boundaries. The oxides grown in the NaOH solution have a protective columnar structure in the pre-transition regime but an equiaxed structure in the post-transition regime. On the other hand, in the KOH solution, the columnar structure is maintained from its pre-transition regime to the post-transition regime. On the basis of the above results, it can be suggested that the cation incorporation into zirconium oxide would control the oxide microstructure, the oxide growth mechanism at the metal–oxide interface and the corrosion rate in alkali hydroxide solutions.     

Oxidation of Zircaloy-2 studied using a wedge-shaped specimen

Experiments have been carried out to vary the stress in the oxide layer during oxidation of Zircaloy-2. This was achieved by varying the thickness of the metal substrate, using a specimen with a tapered wedge-shaped cross-section. X-ray diffraction measurements confirmed that the compressive stress level in the oxide was reduced when the metal substrate was thinner. The rate of oxidation was also slower for conditions where the stress was reduced. The results can be interpreted such that transitions in the growth result from sequential cracking of the oxide when sufficient elastic strain energy accumulates and that the cracks then enhance access of oxygen to the metal interface.     

Crystal structure and grain size of Zr oxide characterized by synchrotron radiation microdiffraction

Zirconium oxides formed on Zirclaoy-4 and Zr-1.5Nb (in wt%) were characterized by the microbeam X-ray diffraction using a synchrotron radiation. The phase fraction and the grain size were determined as a function of the position in the oxide. It was found that Zr-1.5Nb showed the better corrosion resistance than Zircaloy-4 in 360 °C pure water although the tetragonal phase was more stabilized to a further distance from the metal/oxide interface in the oxide of Zircaloy-4. The calculation of the grain size revealed that the oxide of the Zr-1.5Nb had larger grains than that of Zircaloy-4 with the tetragonal phase being smaller than the monoclinic one. It seems reasonable to suppose that the superior corrosion resistance of Zr-1.5Nb was attributed to the lager grain size of the oxide in which the oxygen diffusion is expected to be lowered when compared to the smaller grain size of the oxide on Zircaloy-4.     

Effect of additives on the susceptibility of zirconium alloys to nodular corrosion

A new approach based on a stability analysis of a uniformly growing oxide film was applied to estimate the effect of alloying additives on the susceptibility of zirconium alloys to nodular corrosion. The analytical results agree with available experimental data on effect of Fe and Ni on resistance of Zircaloy-2 and Zircaloy-4 to the growth of nodular oxide.     

Residual stresses and tetragonal phase fraction characterisation of corrosion tested Zircaloy-4 using energy dispersive synchrotron X-ray diffraction

To improve the understanding of the oxidation mechanism in zirconium alloys for fuel clad applications, detailed residual stress and phase fraction analysis was carried out for the oxides formed on Zircaloy-4 after autoclave exposure at 360 °C for various times by means of synchrotron X-ray diffraction. In a post-transition sample (220 days), significant stress variation through the oxide thickness was found for the monoclinic phase in individual oxide layers, with maximum in-plane compressive stresses located towards the metal–oxide interface and a discontinuity in the residual stress profile. The depth of this discontinuity matched well with the depth at which electron microscopy analysis showed an interface between two distinct oxide layers. Analysis of the tetragonal phase with exposure time demonstrated changes of the total volume of tetragonal phase before and during transition. These observations are put into the context of residual stress evolution presented previously, to provide further insight into the importance of phase transformations and residual stresses in determining the corrosion kinetics of Zr alloys.     

Influence of second phase particles on corrosion resistance of N36 alloy in superheated steam

The nodular corrosion resistance of N36 (Zr–0.8Sn–1Nb–0.3Fe–0.12O, wt%) alloy with different temperatures of interstage annealing and optimized Zircaloy-4 (lower content of tin) cladding tubes were tested in 500 °C and 10.3 MPa superheated steam. The result showed that no nodular spots appeared on surface of N36 alloy cladding tubes after 500 h corrosion, while nodular spots on optimized Zircaloy-4 tubes after 8 h. The differences of weight gain between two kinds of N36 tubes suggested that the percentage (volume fraction) of tetragonal Zr dioxide in oxide film can be effectively influenced by characteristics, such as size, of the second phase particles on the oxide film-matrix boundary. The analysis by SEM and LRS clearly indicated that smaller and fine distributed second phase particles were beneficial to stabilize the oxide via endure the uniform stress so it could slow down the altering process from tetragonal Zr dioxide to monoclinic Zr dioxide.     

Developments in PWR cladding corrosion: materials and models

In pressurized water reactors Zircaloy-4 is a standard fuel cladding material. The aim of this paper is to present and evaluate corrosion data generated both in-reactor, and out-of-reactor on PWR claddings made of both Zircaloy-2 and Zircaloy-4 materials. The oxide thickness measurements of cladding tubes irradiated in the Ringhals 3 reactor, and oxide weight gain measurements carried out in Sandvik autoclaves at 400°C, 10.3 MPa clearly show that the stress relief annealed Zircaloy-2 is more corrosion resistant than Zircaloy-4 produced with an identical fabrication route. Furthermore, autoclave tests indicate that the hydrogen pickup fraction of the two alloys is very similar. The obtained data have been evaluated in regard to chemical composition and heat treatment. In addition, computer models, which simulate thermal and hydraulic reactor conditions and corrosion kinetic processes simultaneously, have been used to predict the in-reactor corrosion behaviour of the claddings.     

Structure of zirconium alloy oxides formed in pure water studied with synchrotron radiation and optical microscopy: relation to corrosion rate

A detailed study was undertaken of oxides formed in 360 °C water on four Zr-based alloys (Zircaloy-4, ZIRLO™,¹ Zr-2.5%Nb and Zr-2.5%Nb-0.5%Cu) in an effort to relate oxide structure to corrosion performance. Micro-beam X-ray diffraction was used along with transmitted light optical microscopy to obtain information about the structure of these oxides as a function of distance from the oxide-metal interface. Optical microscopy revealed a layered oxide structure in which the average layer thickness was inversely proportional to the post-transition corrosion rate. The detailed diffraction studies showed an oxide that contained both tetragonal and monoclinic ZrO₂, with a higher fraction of tetragonal oxide near the oxide-metal interface, in a region roughly corresponding to one oxide layer. Evidence was seen also of a cyclic variation of the tetragonal and monoclinic oxide across the oxide thickness with a period of the layer thickness. The results also indicate that the final grain size of the tetragonal phase is smaller than that of the monoclinic phase and the monoclinic grain size is smaller in Zircaloy-4 and ZIRLO than in the other two alloys. These results are discussed in terms of a model of oxide growth based on the periodic breakdown and reconstitution of a protective layer.     

Deuterium diffusion in LiOH–water-corroded oxide layer of zirconium alloys

In order to clarify the hydrogen diffusion mechanism in the oxide layer of zirconium alloys, in situ hydrogen isotope diffusion in the oxide layer has been examined. The zirconium alloys used were Zircaloy-2, GNF-Ziron (Zircaloy-2 type alloy with high iron content) and VB (zirconium-based alloy with high iron and chromium contents). They were corroded in 1 or 0.1 M LiOH-containing water at 563 K, producing oxide layers of 1.1–2.1 μm in thickness. The diffusion experiments were carried out in the temperature range from 488 to 633 K by using a combined technique of deuterium plasma exposure and nuclear reaction analysis for D (³He,p)⁴He reaction. From the transient deuterium profiles in the oxide layers, it was concluded the LiOH–water-corroded oxides had a single-layer structure, which was in contrast to the double-layer structure previously observed in steam-corroded oxide layers. The diffusion coefficients in the 1 M LiOH–water-corroded oxides evaluated from the deuterium profiles were smaller in the order of Zircaloy-2 > GNF-Ziron > VB at 573 K. For the 0.1 M LiOH–water-corroded oxide of GNF-Ziron, the diffusivity was lower than that of the 1 M LiOH–water-corroded oxide by a factor of 1/4. The present diffusion coefficients of the 1 M LiOH–water-corroded oxides of GNF-Ziron and VB were approximately 7 times larger than the previous data of the corresponding steam-corroded oxides. The deuterium diffusion properties in the oxides of the three alloys obtained in the in situ experiment were roughly consistent with their hydrogen absorption performances in the LiOH–water-corrosion tests, as well as in the previous steam corrosion tests.