Relations of Microstructural Attributes and Strength-Ductility of Zirconium Alloys with Hydrides

As the first safety barrier of nuclear reactors,zirconium alloy cladding tubes have attracted extensive attention because of its good mechanical properties.The strength and ductility of zirconium alloy are of great significance to the service process of cladding tubes,while brittle hydrides precipitate and thus deteriorate the overall performance.Based on the cohesive finite element method,the effects of cohesive strength,interfacial characteristics,and hydrides geometric characteristics on the strength and ductility of two-phase material(zirconium alloy with hydrides)are numerically simulated.The results show that the fracture behavior is significantly affected by the cohesive strength and that the overall strength and ductility are sensitive to the cohesive strength of the zirconium alloy.Furthermore,the interface is revealed to have prominent effects on the overall fracture behavior.When the cohesive strength and fracture energy of the interface are higher than those of the hydride phase,fracture initiates in the hydrides,which is consistent with the experimental phenomena.In addition,it is found that the number density and arrangement of hydrides play important roles in the overall strength and ductility.Our simulation provides theoretical support for the performance analysis of hydrogenated zirconium alloys during nuclear reactor operation.

Microstructural evolution of zirconium alloy under dynamic compression at strain rate of 1000s^(-1)

Microstructural evolution of the zirconium alloy deformed at a strain rate of about 1000 s-1 was investigated. Four different strain levels of the zirconium alloy subjected to dynamic compression were designed by several-times impacting at almost the same strain rate. The results show that abundant low angle boundaries at different strain levels were observed in the deformed microstructures, and the quantity and density of low angle boundary increase dramatically with the strain increasing. Besides low angle boundaries and high angle boundaries observed in grain boundary maps, the twin boundaries including the tensile twins {10 2}, {11 1} and compressive twins {11 2} were distinguished at different strain levels, and most twin boundaries were indexed as {10 2} twins. With the stain increasing, the twin boundary density in the deformed microstructures increases indistinctively. Based on the characterization of the deformed microstructures at the different strain levels, the deformation and evolution processes of the zirconium alloy subjected to dynamic loading were proposed. Microhardness measurements show that the microhardness in the impacted specimens increases gradually with the strain increasing, which should be associated with the strain hardening caused by the tangled dislocation.

EFFECT OF MELT OVER-HEATING AND ZIRCONIUM ALLOYING ON THE MORPHOLOGY OF Al_9FeNi PHASE AND MECHANICAL PROPERTIES OF 2618 ALLOY

The effect of melt over-heating on the morphology of Al_9FeNi phase in 2618aluminum alloy with high contents of Fe and Ni and 0.22 wt. percent zirconium has been investigatedby optical microscopy. SEM and TEM. The mechanical properties of 2618 aluminum alloy after hotextrusion and quenching/aging have been tested. The results show: melt over-heating treatment of2618 alloy with high contents of Fe and Ni at 960 deg C led to finer and better-distributedneedle-like Al_9FeNi phase in cast microstructure and fine Al_9FeNi particles after hot extrusion;the grain size of the alloy after hot extrusion could also be refined evidently by alloying ofzirconium; the ambient and high temperature tensile strength and elongation of 2618 alloy have beenapparently enhanced due to fine Al_9FeNi particles and dispersed Al_3Zr as well as fine grain size.

Existing form and effect of zirconium in pure Mg,Mg-Yb,and Mg-Zn-Yb alloys

The existing form and grain refining effects of small zirconium addition in pure Mg, Mg-Yb and Mg-Zn binary alloys, and Mg-Zn-Yb ternary alloy （ZK60-Yb） were investigated. The results show that Zr element exists mainly in single and cluster particles of pure α-Zr or Zn-Zr compounds inside grains and at grain boundaries. Only the particles located in the interior of grains can act as the nucleus for α-Mg growth and effectively promote the formation of fine equiaxed grains. The broken and dispersed Zr-rich particles produced during the hot extrusion process can form nebulous banded structure in which these fine particles may act as obstacles to dislocation motion in wrought magnesium alloys.

Effects of annealing treatments on forming performance of zirconium alloys

The effects of annealing treatments(ATs)on the microstructure of Zr-Sn-Nb alloy strips were studied.Based on the characteristics of strips for nuclear fuel assemblies,punching experiments were carried out and the formability of zirconium alloy strips was quantitatively evaluated.The results indicate that the proportions of small-angle grain boundaries of the zirconium alloy under conditions of annealing treatment at 580°C(ATⅠ)and annealing treatment at 620°C(ATⅡ)are 14.3%and 23.2%,respectively,while that of the as-received material is 12.4%.And the forming limit margin fields of the zirconium alloy under ATⅠcan reach 0.43%,while the values of the as-received material and the ATⅡare-0.35%and-2.8%,respectively.The annealing process affects the evolution process of the strip recrystallization texture and the grain size.Moreover,the total texture and pole density are closely related to the degree of anisotropy of the strip.Besides,the small-angle grain boundary affects the strain path and crack expansion of the necking unit during the strip punching process,while the grain size affects the hardening exponent of the material.

Interfacial reaction between zirconium alloy and graphite mold/yttrium oxide ceramic mold

Zirconium alloys are active in the molten state and tend to react with the mold during casting. The casting technology of zirconium is not yet well established; especially in selecting the mold materials, which are difficult to determine. In the present work, the interfacial reactions between zirconium casting and casting mold were studied. The zirconium alloy was melted in a vacuum arc skull furnace and then cast into the graphite mold and ceramic mold, respectively. The zirconium casting samples were characterized using SEM, EDS and XRD with an emphasis on the chemical diffusion of elements. A reaction layer was observed at the casting surface. Chemical analysis shows that chemical elements C, O and Y from the mold are diffused into the molten zirconium, and new phases, such as ZrC, Zr30, YO1.335 and Y6ZrO11, are formed at the surface. In addition, an end product of zirconium valve cast in a yttria mold has a compact structure and good surface quality.

The Influence of Different Contents of Bi Addition on the Corrosion Behavior of Various Zirconium-Based Alloys

Zr-4(Zr-1.5Sn-0.2Fe-0.1Cr,wt%), S5(Zr-0.8Sn-0.34Nb-0.39Fe-0.1Cr), T5(Zr-0.7Sn-1.07Nb-0.32Fe-0.08Cr) and Zr-1Nb were adopted to prepare Bi-containing zirconium alloys for systematically investigating the effect of Bi addition on the corrosion resistance of zirconium alloys. The specimens were corroded in superheated steam at 400℃/10.3 MPa, and in lithiated water with 0.01 M LiOH or in deionized water at 360℃/18.6 MPa by autoclave testing. Results show that the corrosion resistance increases with the increasing of Bi content dissolved in α-Zr. But the presence of Bi-con- taining second phase particles (SPPs) is unfavorable for the enhancement of corrosion resistance. This indicates that the Bi dissolved in α-Zr matrix plays an important role in improving the corrosion resistance, while the precipitation of the Bi-containing SPPs does harm to the corrosion resistance.

Different grain refinement mechanisms of minor zirconium and cerium in magnesium alloys

Zirconium and rare earth element cerium were added in magnesium and magnesium alloys to study their different grain refinement mechanisms. The results show that zirconium has an obvious refinement effect on the cast grain of magnesium and its alloys without the alloy element Al because the crystal structure of zirconium is the same as magnesium matrix,and the lattice parameters are close to magnesium. Zirconium can decrease the grain size of magnesium from 150 to 20 μm. The rare earth cerium also has a grain refinement effect on Mg and Mg-Al alloy. The cerium atoms tend to remain in the liquid rather than solidify with the solvent atoms magnesium at the solid-liquid interface. The liquid constitutional undercooling can provide a heterogeneous crystal nucleation. The grain is refined from 200 μm to 40-80 μm. These two elements have different grain refinement mechanism on Mg alloy. The mechanism of zirconium is that it acts as the nuclei of α-Mg. But the mechanism of cerium is that it increases the liquid constitutional undercooling that can provide a heterogeneous crystal nucleation for the alloy.

OUT-OF-PILE CORROSION RESISTANCE OF NEW ZIRCONIUM ALLOYS IN 500 ℃/10 3MPa STEAM

he out of pile corrosion resistance of new Zirconium alloys in 500℃/10 3MPa steam has been investigated and the effect of alloying elements, Sn,Nb,Fe,Cr on the property has been analyzed. The results show that the new alloys have better corrosion resistance than Zircaloy 4. That no nodular corrosion is found during test cycles shows the nodular corrosion resistance can be dramatically improved by addition of alloying elements Nb. The Fe/Cr ratio should be properly controlled if there is Cr addition in the alloys when developing new alloys.

Effects of scandium and zirconium combination alloying on as-cast microstructure and mechanical properties of Al-4Cu-1.5Mg alloy

The influences of minor scandium and zirconium combination alloying on the as-cast microstructure and mechanical properties of Al-4Cu-1.5Mg alloy have been experimentally investigated.The experimental results show that when the minor elements of scandium and zirconium are simultaneously added into the Al-4Cu-1.5Mg alloy,the as-cast microstructure of the alloy is effectively modified and the grains of the alloy are greatly refined.The coarse dendrites in the microstructure of the alloy without Sc and Zr additions are refined to the uniform and fine equiaxed grains.As the additions of Sc and Zr are 0.4% and 0.2%,respectively,the tensile strength,yield strength and elongation of the alloy are relatively better,which are 275.0 MPa,176.0 MPa and 8.0% respectively.The tensile strength is increased by 55.3%,and the elongation is nearly raised three times,compared with those of the alloy without Sc and Zr additions.

Study on Corrosion Resistance of N36 Zirconium Alloy in LiOH Aqueous Solution

Zr-Sn-Nb-Fe alloys are one of the important directions for continuous improvement of zirconium alloys for high burn-up fuel assemblies. The corrosion resistance of Zr-Sn-Nb-Fe alloys is closely related to the alloying element and water chemical condition. To better understand the effect of Sn on corrosion resistance of Zr-Sn-Nb-Fe alloy, the normal N36 (Zr-1Sn-1Nb-0.3Fe) and low-tin N36 (Zr-0.8Sn-1Nb-0.3Fe) alloy sheets were prepared and tested in static autoclave in both of 0.01 mol/L LiOH and 0.03 mol/L LiOH aqueous solution at 360°C and 18.6 MPa. The characteristics of the microstructure and oxide film of alloys were analyzed by TEM and SEM respectively. It was shown that that the corrosion transition of the normal N36 appears earlier and the weight gain is higher than the low-tin N36 in two corrosive mediums. The cracks paralleling to the interface of oxide/metal are formed in the fracture surface of the oxide film and the micrographs at the oxide film/substrate interface appear uneven morphology. With the increasing of corrosion gain, there are more parallel cracks in oxide film and the uneven morphology at the oxide film/substrate interface is more obvious.

氢含量对锆合金蠕变-疲劳行为的影响及机理研究

针对反应堆用锆合金渗氢后易发生蠕变-疲劳失效问题,采用频率修正应变寿命法和频率修正滞回能法研究了再结晶状态的锆合金在320oC、不同渗氢含量下的蠕变-疲劳行为。结果表明:随保持时间增加,无渗氢试样的抗蠕变-疲劳性能降低,但保持时间30 s以上时无明显影响;保持时间对渗氢试样的抗蠕变-疲劳性能无影响。不同保持时间下,未渗氢试样的抗蠕变-疲劳性能最好,0.04%氢含量试样的性能最差,0.005%氢含量在高应变水平(高滞回能)下的抗蠕变-疲劳性能弱于0.02%氢含量试样,而在低应变水平区域优于0.02%氢含量试样。氢含量的影响机理为:固溶氢可提高蠕变-疲劳寿命,氢化物可降低蠕变-疲劳寿命;0.04%氢含量试样中氢化物起主导作用,导致其蠕变-疲劳性能最差。

核用锆合金管氧化过程中光谱发射率特性研究

发射率在锆合金管非接触测温的准确性和失水事故情况下的传热分析上具有重要意义。本文采集了SZA4、SZA6和ZIRLO 3种锆合金在400,500,600℃下氧化100min过程中4~12μm内的光谱发射率数据,分析了温度、氧化膜厚度和合金组分对材料光谱发射率特性的影响规律。结果表明,3种锆合金在氧化过程中发射率变化趋势基本相同,但组分差异导致在发射率的数值和振荡特性上略有差别。通过建立薄膜干涉模型,解释了光谱发射率在波长方向上的振荡现象,并结合振荡极值与干涉效应原理计算了合金表面氧化膜厚度,在0.6~8μm范围内测量精度达到±0.3μm。

乏燃料棒M5锆合金包壳的透射电镜分析

压水堆燃料元件的锆合金包壳,在服役期间会经受高中子注量辐照,其微观组织将发生很大变化,从而影响其宏观性能,因此锆合金包壳的中子辐照行为研究一直是核领域的研究重点。但由于材料经中子辐照后具有较强的放射性,相关的实验必须在热室内进行,因此针对辐照后燃料包壳微观组织的研究也一直是工作的难点。本文在中国原子能科学研究院热室设施上,通过透射电镜分析手段,研究了M5锆合金包壳材料中子辐照后的微观组织。样品来源于国内商业压水堆AFA3G型乏燃料棒,其燃耗分别为14 GW·d/tU和41 GW·d/tU。从燃料棒上截取长度约10 mm的包壳样品,在热室内完成去芯块与化学清洗,获得空包壳样品,然后通过机械制样方法,制备出?3 mm薄片状包壳基体样品,最后采用电解双喷减薄方法,制备出包壳透射电镜观察分析样品。另外,为对比锆包壳辐照后的组织变化,采用同样方法制备了相同材料的冷态观察分析样品。冷态样品与辐照样品的观察分析结果表明:冷态Zr合金包壳基体组织内部存在原生的第二相粒子,基体内部整体较为干净,纳米析出相稀少,未观察到明显的位错结构;辐照后,基体内原生的第二相粒子尺寸和分布与冷态样品差异不明显,但出现了明显的纳米析出相和高密度位错组织;随着燃耗的增加,纳米析出相尺寸有增加的现象;低燃耗与高燃耗样品位错组织具有相似性,表明在14 GW·d/tU燃耗下,锆合金包壳内由辐照产生的位错组织已基本趋于饱和状态;电子选取衍射结果表明,辐照后,基体内原生的第二相粒子虽存在一些非晶组织,但仍以bcc晶体结构为主,表明在41 GW·d/tU燃耗下,第二相粒子保持了一定的辐照稳定性;另外,第二相的EDS结果表明,随着燃耗的增加,Nb元素的含量有贫化趋势;分析认为,Zr合金经中子辐照,第二相粒子中的Nb原子扩展至Zr基体内,将促进Nb元素以纳米富Nb相形式在Zr基体中析出。

基于内压法的锆合金薄壁管泊松比试验研究

锆合金包壳管在核电应用中存在变形及破裂等问题,严重影响核电运行安全。为了更好地评价锆合金包壳管等薄壁管材受内压时的变形性能,给燃料组件的设计、计算、验证提供参考,设计基于内压法的泊松比测试方法。根据薄壁圆筒受均匀内压时的应力解,推导出基于内压加载的薄壁管泊松比理论公式。按照该理论公式设计合理的内压试验方法,采用该方法对N36锆合金薄壁管开展20~400℃下的泊松比试验,同时通过拉伸法进行相应温度下的对比试验。结果表明,内压试验方法可以有效获得不同温度下锆合金薄壁管的泊松比,且内压试验方法得到的锆合金薄壁管泊松比与拉伸法结果较为一致,相对误差均在2%以内。试验结果同时表明:N36锆合金薄壁管的泊松比随试验温度的增加呈增大趋势。

聚能装药水下破坏双层靶板数值模拟研究

针对水下爆破舰船以及构筑物装药的高效要求,采用锆基非晶合金作为聚能装药药型罩材料,并采用有限元数值计算方法,对比分析了锆基非晶合金和传统紫铜材料作为药型罩对水下双层靶板毁伤性能分析。结果表明,锆基非晶合金材料药型罩聚能装药较之紫铜材料药型罩在金属射流速度上性能提升14.9%,破板后效(穿靶后射流剩余速度)提升37.3%,可为爆破钢结构构筑物提供参考。

锆合金氧化膜及基体中氧的扩散

锆合金氧化膜及基体中氧的扩散系数是锆合金腐蚀动力学中的重要参数,目前文献报道的氧在锆及氧化膜中的扩散系数数值差异较大。本文通过真空退火试验,得到不同温度下氧化膜中氧浓度分布,计算了氧在锆合金基体中的扩散系数;通过氧化膜的等效扩散模型,由腐蚀转折前的腐蚀增重曲线,估算锆合金氧化膜中氧的扩散系数,得到Zr-Sn-Nb合金基体中氧的扩散系数随温度的变化规律为D_(Zr)(cm^(2)/s)=0.18exp(-180 000/RT);通过转折前的腐蚀增重曲线,估算得到氧化膜中氧的扩散系数随温度的变化规律为D_(ox)(cm^(2)/s)=3×10^(-7)exp(-101 550/RT)。

电感耦合等离子体原子发射光谱法测定锆合金中5种元素的含量

提出了电感耦合等离子体原子发射光谱法测定锆合金中锡、铌、铁、铬、镍含量的方法。取样品0.1 g,加入5 mL水、5 mL硝酸和0.5 mL氢氟酸,于100℃加热5 min,用水定容至25 mL。设置仪器射频功率为1300 W,雾化气流量为0.80 L·min^(-1),进样量为2.0 mL·min^(-1),测定5种元素在Sn 242.949 nm、Nb 309.418 nm、Fe 259.941 nm、Ni 221.648 nm、Cr 267.716 nm等分析谱线处的响应强度,基质匹配法定量。结果表明:锡、铌的质量浓度在150μg·mL^(-1)以内,铁、铬、镍的质量浓度在15μg·mL^(-1)以内分别与对应的响应强度呈线性关系;锡检出限(3s)为16μg·g^(-1),其他元素检出限(3s)均低于3.0μg·g^(-1);各元素方法重复性的相对标准偏差(n=10)均小于2.0%,加标回收率为99.0%~109%;采用锆合金标准物质SRM360b进行验证,测定值均在认定值的不确定度范围内。

涂层锆合金包壳管切向微动磨损数值预测模型研究

为建立涂层锆合金包壳微动磨损的有限元计算分析模型,预测切向工况下的最大磨损深度,基于包壳-格架的实际几何特征建立了涂层锆合金包壳-刚凸接触系统的有限元模型,采用Archard磨损模型计算磨损量,结合ABAQUS的ALE功能和UMESHMOTION子程序模拟磨损过程。根据微动磨损试验结果分析磨损系数随磨损周次的变化情况,并拟合其函数关系。有限元计算的最大磨损深度和最大深度的位置与试验结果相符,表明所建立的模型可用于涂层锆合金包壳切向微动磨损最大深度的预测。

中子毒物硼和钆在乏燃料后处理中的应用及合金相图研究

随着我国核电站在建及运行的反应堆日渐增多,乏燃料的产生、运输和储存成为核工业面临的严峻难题。对卸载出的巨量核废乏燃料,采用闭环处理循环利用,比传统深埋更加环保和安全。采用添加中子毒物的核结构材料制造相关设备和仪器,可以保障乏燃料后处理过程的运行安全。基于核工业应用要求,针对乏燃料后处理常用的316L不锈钢、Ti35、Zr等材料添加B或者Gd的相图展开调研,结果显示B和Gd可明显提高结构材料的中子吸收效果,但B和Gd的化学性质与过渡族金属Fe,Ti,Zr等相差甚远,除α-Zr最多固溶2.8%Gd(原子数分数)以外,B和Gd均不能大量固溶于常规金属合金体系中,强制添加势必对合金力学性能或耐蚀性能带来巨大危害。可采用团簇模型的合金设计理念,通过含B和Gd化合物相均匀析出或者复合的制造工艺以及Zr-Gd、Ti-Gd体系的多元合金化,引入互溶的中间相,增加难以固溶的元素的固溶度,以此解决中子毒物B和Gd在乏燃料后处理中的应用瓶颈。