锆合金带电离子辐照效应及氦泡演化行为研究

发布时间：2018-06-07 14:55

本文选题：锆合金 + 辐照效应　；参考：《电子科技大学》2015年博士论文

【摘要】：核电技术发展至今已有60多年历史,国际原子能机构(IAEA)2014年年度报告显示,目前全球核电运行机组共有434台,其发电总量为371.7GW,约占世界发电总量的16%。随着化石能源的逐渐消耗,全球能源变得十分稀缺。核能作为一种清洁能源,可以在满足增加的电能需求的同时减少碳排放量,实现国家能源安全和可持续发展的目标,因而核电技术的发展将会得到更高的重视。核能工业的发展促进了对适应于反应堆环境材料的研究与开发,评估材料在核反应堆中的适用性能的主要标准为材料的机械强度和高温水热环境中的耐腐蚀性能。锆及其合金具有较好的机械性能、较强的耐腐蚀性能、低的热中子吸收截面等特性,因而被作为核能工业中的战略材料,在核反应堆中主要被用作结构材料和燃料棒包壳等。反应堆运行时,锆合金将受到中子束流及多种离子束、!射线的辐照。辐照会导致锆合金微观结构变化,进而影响合金的机械性能和耐腐蚀性能,最终导致材料失效。因此,研究锆合金的耐辐照性能具有重要意义。自1984年中国第一座核反应堆通过技术鉴定以来,经过30年在核电领域的快速发展,中国现已成为核电领域的先进国家,由此增加了国家对优质锆合金的需求和开发具有自主知识产权的高性能锆合金的紧迫性。N18锆合金是一种中国自主研发的锆合金,堆外性能测试结果证实N18具有更优异的耐腐蚀性能、吸氢性能和力学性能。但是,对该合金的耐离子束辐照性能、氦泡的演化行为及其氢化行为的研究还很少。本文将从以上三个方面内容展开详细研究,评估N18锆合金的抗辐照性能、氦泡演化行为及氢化行为,主要研究内容分为四个部分:(1)研究了辐照温度为360 oC时2 MeV的质子束辐照N18锆合金的辐照效应,结果表明:锆合金中的第二相粒子主要为密排六方结构的Zr(Fe,Cr,Nb)2,辐照剂量达到3.9 dpa时,第二相粒子开始发生部分非晶化转变,颗粒呈现出核壳结构,即第二相粒子内部为晶体,边缘为非晶态;当辐照剂量为8.2 dpa时,第二相粒子发生完全非晶化转变。对非晶化转变前后的第二相粒子采用能谱仪进行线扫描研究,结果显示第二相粒子发生非晶化转变的机制与Fe元素由第二相颗粒扩散进入基体紧密相关。(2)在透射电子显微镜中原位研究了辐照温度为310 oC时500 keV的Ne离子束辐照N18锆合金的辐照效应,结果表明:N18锆合金中第二相粒子Zr(Fe,Cr,Nb)2经Ne离子辐照至0.5 dpa时开始发生非晶化转变;当辐照剂量增加至1.0 dpa时,第二相粒子完全非晶化。此外,对完全非晶化后的第二相粒子进行原位退火研究,结果显示退火温度为450 oC时第二相粒子开始发生再结晶,并在温度为600 oC时完全结晶。完全结晶后形成的纳米颗粒具有与辐照前Zr(Fe,Cr,Nb)2相同的密排六方结构。此外,400 keV的重离子Fe辐照N18锆合金后会导致其显微硬度增大,这是由于辐照在合金中产生大量的位错所致。(3)采用离子注入方法在N18锆合金中引入He离子,注入能量为400keV,研究了合金中形成的氦泡与注入剂量和辐照温度的关系,以及加热过程中原位研究氦泡的演化行为,结果表明:合金中氦泡的尺寸随辐照剂量的增加而增大,辐照温度越高越有利于氦泡的长大;氦泡的原位加热研究结果显示,温度低于650 oC时,氦泡的成长速度缓慢,温度高于650 oC时,氦泡迅速长大,氦泡的长大遵循气泡的迁移合并机制。此外,采用纳米压痕仪研究了锆合金的显微硬度与注氦量的关系,表明合金表层硬度值随注氦量的增加而增大。(4)在扫描电镜中原位研究了N18锆合金的氢化反应,系统中的氢来源于Ga离子束激发有机金属气体MeCpPtIVMe3分解。形成的锆氢化物具有针状形貌,面心立方结构的!-ZrH1.5-1.66。针状的锆氢化物锆排列有序,并与锆合金基体"-Zr具有[010]"//[1-10]!,(001)"//(111)!的外延取向关系。氢化物的形成被认为与锆合金内应力释放导致的应变驱动力相关。在透射电子显微镜中采用选区电子衍射和电子能量损失谱原位研究了锆氢化物的热稳定性,结果表明:加热前锆氢化物为!-ZrH1.5-1.66相,温度升高至450 oC时,!-ZrH1.5-1.66相发生分解,分解产物为#-ZrH0.25-0.5相,升温至700 oC后,#-ZrH0.25-0.5相仍然稳定存在。
[Abstract]:Nuclear power technology has been developed for more than 60 years. The annual report of the International Atomic Energy Agency (IAEA) in 2014 shows that there are 434 units in the global nuclear power unit at present. The total amount of power generation is 371.7GW, and the 16%. of the world's total power generation is deplete with the gradual depletion of fossil energy, and nuclear energy is a clean energy. The development of nuclear power technology will be paid more attention to reducing carbon emissions and achieving national energy security and sustainable development at the same time to meet the increased demand for electrical energy. The development of nuclear energy industry promotes the research and development of the materials adapted to the reactor environment and the evaluation of the application performance of the materials in the nuclear reactor. The main standard is the mechanical strength of materials and the corrosion resistance in high temperature hydrothermal environment. Zirconium and its alloys have good mechanical properties, strong corrosion resistance, low thermal neutron absorption cross section and so on. Therefore, they are used as strategic materials in nuclear energy industry, and are mainly used as structural materials and fuel rod cladding in nuclear reactors. During the operation of the reactor, the zirconium alloy will be subjected to a neutron beam and a variety of ion beams. Radiation irradiation. Radiation will lead to the change of the microstructure of the zirconium alloy, which will affect the mechanical and corrosion resistance of the alloy, and eventually lead to the failure of the material. Therefore, it is of great significance to study the radiation resistance of the zirconium alloy. The first nuclear reaction in China in 1984. After 30 years of rapid development in the field of nuclear power, China has now become an advanced country in the field of nuclear power, which has increased the national demand for high quality zirconium alloys and the urgency of developing high performance zirconium alloys with independent intellectual property rights..N18 zirconium alloy is a kind of self developed zirconium alloy developed by China. The test results show that N18 has better corrosion resistance, hydrogen absorption and mechanical properties. However, there are few studies on the ion beam radiation resistance of the alloy, the evolution behavior of helium bubbles and their hydrogenated behavior. This paper will discuss the radiation resistance of N18 zirconium alloys, the behavior of helium bubble evolution and the behavior of the helium bubble from the above three aspects. The main research content of hydrogenation is divided into four parts: (1) the irradiation effect of N18 zirconium alloy irradiated by proton beam of 2 MeV at radiation temperature of 360 oC is studied. The results show that the second phase particles in the zirconium alloy are mainly the Zr (Fe, Cr, Nb) 2 of the dense six square structure, and the second phase particles begin to be partially amorphous when the amount of irradiation agent reaches 3.9 DPA. The particles present a nuclear shell structure, that is, the second phase particles are crystalline and the edges are amorphous. When the irradiation dose is 8.2 DPA, the second phase particles are completely amorphous. The second phase particles before and after the amorphous transition are scanned by the energy spectrometer. The result shows the mechanism of the non crystalline transition of the second phase particles and the Fe The element is closely related to the diffusion of the second phase particles into the matrix. (2) in the transmission electron microscope, the irradiation effect of N18 zirconium alloy irradiated by Ne ion beam of 500 keV when radiation temperature is 310 oC is investigated. The results show that the second phase particle Zr (Fe, Cr, Nb) 2 of the N18 zirconium alloy begins to undergo amorphous transition when the Ne ions are irradiated to the 0.5 DPA; When the dose increased to 1 DPA, the second phase particles were completely amorphous. In addition, the second phase particles were in situ annealed. The results showed that the second phase particles began to recrystallization and crystallized at the temperature of 600 oC when the annealing temperature was 450 oC. The nanocrystalline particles formed after complete crystallization were with Zr (F) before irradiation. E, Cr, Nb) 2 the same six square structure. In addition, 400 keV heavy ion Fe irradiation of N18 zirconium alloy will lead to the increase of the microhardness, which is caused by a large number of dislocations in the alloy. (3) the ion implantation method is used to introduce He ions in the N18 zirconium alloy and the injection energy is 400keV. The helium bubble formed in the alloy and the injection are studied. The relationship between the dose and irradiation temperature and the study of the evolution of helium bubbles in situ during heating show that the size of helium bubbles in the alloy increases with the increase of irradiation dose, and the higher the irradiation temperature is, the higher the helium bubble growth. The results of the helium bubble in situ show that the growth rate of helium bubble is slow when the temperature is below 650 oC. When higher than 650 oC, the helium bubble grows rapidly and the growth of the helium bubble follows the migration and consolidation mechanism of the bubble. In addition, the relationship between the microhardness of the zirconium alloy and the amount of helium is studied by the nano indentation instrument. It shows that the hardness value of the alloy surface increases with the increase of the amount of helium. (4) the hydrogenation reaction of the zirconium alloy in the central plain of the scanning electron microscope is studied in the system. The hydrogen is derived from the decomposition of the organic metal gas excited by the Ga ion beam. The zirconium hydride formed with a needle like morphology and a face centered cubic structure! -ZrH1.5-1.66. needle like zirconium hydride zirconium is arranged in order, and the zirconium alloy matrix "-Zr has [010]" //[1-10]!, (001) "/ / (111)!" epitaxy orientation relationship. The formation of hydride is considered to be with zirconium. The thermal stability of zirconium hydride is in situ studied by electoral electron diffraction and electron energy loss spectrum in a transmission electron microscope. The results show that the zirconium hydride is the -ZrH1.5-1.66 phase before heating, when the temperature rises to 450 oC, the -ZrH1.5-1.66 phase breaks down and the decomposition product is #-ZrH0.25 The #-ZrH0.25-0.5 phase remained stable after -0.5 phase heating up to 700 oC.
【学位授予单位】：电子科技大学
【学位级别】：博士
【学位授予年份】：2015
【分类号】：TL341

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