锂陶瓷微球释氘行为及其与辐照缺陷的相关性研究

【作者】 肖成建；

【导师】 汪小琳；

【作者基本信息】 中国工程物理研究院 ， 核燃料循环与核材料， 2013， 博士

【摘要】 氚增殖剂是聚变堆产氚包层的核心功能材料,其在中子场下的氚释放行为和辐照性能是氘氚燃料循环工艺、包层工程设计所关心的重要内容。中国已开展了20余年的固态氚增殖剂制备工作,如LiAlO2、Li2ZrO3.Li2TiO3和Li4SiO4等,但对这些材料性能,尤其是辐照产氚性能研究得较少,缺乏对材料辐照效应和氚相关基础问题的认识,无法满足聚变工程设计的要求。针对目前的研究需求,在缺少聚变中子源的实验条件下,本文采用裂变反应堆辐照锂陶瓷氚增殖剂,在堆外热解吸(TDS)实验平台上系统研究材料的氚释放行为及其影响因素,掌握氚释放过程的速控步骤；运用伽玛射线、电子束和中子辐照产生增殖剂材料的体相缺陷,比较研究辐照前后的微观组织变化,采用电子自旋共振实验技术研究辐照缺陷的顺磁特征；利用加热退火的方法加速体相缺陷的恢复速度,研究辐照缺陷湮灭的动力学行为；分析氚释放行为和体相辐照缺陷演变的相关性,建立氚与缺陷的相互作用模型；通过离子注入的方式将氘引入增殖剂材料,研究表面辐照缺陷与氢同位素热解吸行为的关系。以不同升温速率进行堆外热解吸实验,研究结果表明,以LiOH为原料、冷冻成型工艺制备出的Li4SiO4陶瓷微球(～80%T.D)具有较低的氚解吸活化能(40.0±4.2kJ/mol),表现出适宜的低温放氚窗口(500～800K).采用球形扩散模型计算得到增殖氚在Li4SiO4微球中的扩散动力学参数,结合表观热解吸活化能,确定以纯氦为载气的条件下,氚在晶粒内的扩散是氚释放过程的速控步。陶瓷微球晶粒的扩散动力学数据与粉末晶体相近,说明氚在晶界内和孔道内的扩散对氚释放的影响较小。氚扩散系数随中子注量的增大而减小,说明辐照缺陷的增多会阻滞氚在晶粒中的扩散。论文同样得到LiAlO2微球的氚解吸活化能为128.7士28.6kJ/mol,氚释放主要分布在750～1000K,说明LiAlO2需要相对较高的温度窗口提氚。采用He,He+0.1%H2或He+0.1%H2O为载气,分别研究载气组成对锂陶瓷释氚形态的影响。实验结果表明,氚释放形态由晶粒内氚扩散、表面热解吸和氢同位素交换反应竞争控制。Li4SiO4中氚的扩散速度快,表面热解吸活化能小,表面-OT与载气中H2的同位素交换反应贡献小,因此氚释放形态受载气组成的影响小；LiAlO2中氚的扩散速度慢,热解吸活化能与表面氢同位素交换反应(-OT与H2)的活化能在相近的能阈范围,因此释氚形态受载气组成的影响大,即氦气加H2会增加HT的释放比例。此外,锂陶瓷微球表面负载少量的催化活性元素(Pt, Pd, Ru或Ir),可以显著增强表面氢同位素交换反应,加快表面滞留氚的释放,促进分子氚(HT)在低温段的回收。在辐照效应研究方面,采用低剂量的γ射线、电子束和热中子辐照锂陶瓷微球,不会对表面或体相造成严重的辐照损伤,但会诱生缺陷色心,改变材料颜色。采用ESR实验技术检测顺磁缺陷信号,分析辐照缺陷特征,发现Li20主要产生F+中心,而三元锂氧化物辐照后主要产生E’中心和氧缺陷中心(O--center, O2-center)。辐照剂量大于500kGy后,Li4SiO4的非中子辐照缺陷与中子辐照缺陷表现出相似的ESR谱形特征,为模拟研究中子辐照效应提供了参考。采用退火的方法使与缺陷相关的扩散、捕获或脱陷等反应过程加速,研究中子辐照缺陷演变的动力学行为。在等速升温(5K/min)过程中,各种离位原子和电子扩散复位,Li4SiO4的中子辐照缺陷逐渐消失。当温度超过798K,表面氚化物分解并在胶体硅表面产生硅悬挂键,生成新的缺陷中心；胶体物质捕获氚形成的氚化物(Si-T或Li-T)可能是锂陶瓷氚滞留的重要因素；运用一级反应动力学方程将Li4SiO4中子辐照缺陷的退火过程分解成快速和慢速两个过程：快退火活化能为0.18±0.02eV,与电子扩散和氢原子扩散有关；慢退火活化能为0.57士0.06eV,与氧回位造成E’心湮灭有关。比较Li4SiO4的辐照缺陷退火曲线和氚释放曲线,观察到缺陷湮灭过程快要结束时,氚释放速率才迅速递增,由此推测缺陷慢退火行为与氚释放行为存在一定关系。根据缺陷慢退火活化能与氧扩散活化能差异性,论文采用两种模型阐释了氧空位缺陷与氚的相互作用。一种模型延用传统的理论假设——“氧原子扩散复位造成E’心湮灭”,但氧回位产生-0.60eV的驰豫能。另一种模型根据辐照缺陷的慢退火活化能与羟基的扩散活化能在同一个能阈范围(0.56-0.58eV),推测辐照缺陷慢退火与羟基扩散、氧回位有关。采用D2+注入的方式模拟研究表面辐照缺陷对氚释放行为的影响。结果表明,3KeVD2+注入后,Li4SiO4表面状态遭到破坏,部分缺陷会捕获D形成Li-D、O-D,大部分D以非O-D存在。注氘后的热解吸谱与材料表面化学状态有关。以上研究结果只适用于低剂量的辐照实验,还不能延伸回答聚变包层中锂陶瓷增殖剂的氚释放和缺陷行为。更多还原

【Abstract】 Tritium breeder is considered as the key function material of the tritium breeding blanket in a fusion reactor. The tritium release behavior and the irradiation properties of tritium breeder in neutron fields are the important targets for the deuterium-tritium fuel cycle technology and the engineering design of the blankets. The preparation work of the solid tritium breeders, such as LiAlO2, Li2ZrO3, Li2TiO3and Li4SiO4, has been carried out for more than20years in China, but the properties of these materials are studied less, especially the tritium performance during irradiation. Therefore, the lack of the understanding about the tritium issues and irradiation effects can not meet the engineering requirements. According to the current research needs without fusion neutron source, out-of-pile annealing experiments are performed to study the tritium release behavior of the lithium ceramics which are irradiated in fission reactors in this work. The influencing factors, including the heating rate, the isothermal temperature and the purge gas, are investigated respectively to discuss the rate-controlling step during the tritium release process, y-ray, electron beam and neutron irradiation are used to produce defects in lithium ceramic pebbles, the changes of the microcosmic structure before and after irradiation are compared, and the characteristics of irradiation defects are detected by means of electron spin resonance (ESR). The annihilation kinetics of the irradiation defects in neutron-irradiated Li4SiO4is investigated using isochronal and isothermal annealing methods to accelerate the recover rate of the defects. The correlation between annihilation of the irradiation defects and tritium release is analyzed, and then the model of the mutual action between defects and tritium is established. The surface irradiation defects and the hydrogen isotopes desorption behavior of Li4SiO4are investigated by D2+implantation.Li4SiO4ceramic pebbles (-80%T.D), which are made of LiOH with freezing-shaping technology, show good tritium release performance in the low temperature region of500-800K. The apparent desorption activation energy of the bred tritium on the pebble surface is evaluated to be40.0±4.2kJ/mol based on the TDS experiments at different heating rates. The apparent desorption activation energy of tritium on the pebble surface was consistent with the diffusion activation energy of tritium in the crystal grains, indicating that tritium release was mainly controlled by diffusion process. The diffusion coefficients of tritium in the crystal grains at temperatures ranging from450K to600K are obtained by isothermal annealing tests, and the Arrhenius relation is determined to be D=1x10-7.0exp(-40.3×103/RT) cm2s-1. The tritium diffusivity of the Li4SiO4ceramic pebbles in this work is close to that of Li4SiO4crystal powder in the literature, implying that the diffusion of tritium through grain boundaries has little effect on the tritium release rate of Li4SiO4ceramic pebbles. The effect of irradiation fluence on tritium diffusivity in the samples is observed as reduction of the tritium diffusion coefficient with increasing the neutron fluence in the studied range. The bred tritium requires a high temperature region of750-1000K to be liberated from LiAlO2ceramic pebbles corresponding to the apparent desorption activation energy, which is evaluated to be128.7±28.6kJ/mol.The release forms of the bred tritium in lithium ceramics influenced by purge gas composition are investigated. He, He+0.1%H2and He+0.1%H2O are used as purge gas respectively in the TDS experiments. The experimental results show that the released pieces of tritium are co-controlled by the desorption reaction and the hydrogen isotope exchange reaction on the pebble surface, indicating that hydrogen addition to the purge gas plays small role on the released forms of the tritium for Li4SiO4because of its quick diffusivity and low desorption activity energy, but can significantly change the released form of the tritium for LiAlO2because of its slow diffusivity and high desorption activity energy. Catalytic metals loaded on the lithium ceramic pebbles can enhance the hydrogen isotope exchange reaction between tritium on the solid surface and hydrogen in the purge gas, and accelerates the recovery rate of the molecular tritium (HT) in the low temperature range.With respect to the irradiation effects of Li4SiO4ceramic pebbles, y-ray, electron beam and thermal neutron are used in a low dose range. The experimental conditions can not produce serious damage on the surface or in the body of the specimens, but induce some kinds of defect centers, which change the specimens’color. Electron spin resonance technique (ESR) was employed to analyze the paramagnetic defects in the specimens before and after irradiation. It is found that F+center is the principal paramagnetic defect existing in Li2O after irradiation, and there are E’-center, O--center and O2--center in ternary lithium ceramics after irradiation. The ESR spectra of Li4SiO4irradiated by un-neutron irradiation methods show similar characteristics with the spectrum of the specimen irradiated by thermal neutron when the irradiation dose is more than500kGy. This result provides a reference to study the irradiation effects of material exposed to neutron field.The annihilation kinetics of the irradiation defects in neutron-irradiated Li4SiO4is investigated with annealing methods to accelerate the reaction rates related to the defects. Trapped electrons and various kinds of defects recover via diffusion at a heating rate of5K/min from R.T. to798K, then the defects gradually annihilate. some holes are released from the recombination reactions of the E’s and PORs to enhance the production of the amorphous silicon centers in the present Li4SiO4when the annealing temperature is higher than798K. Tritium compounds (Si-T or Li-T) formed by colloids trapping tritium may be the considerable reason of tritium inventory in high temperature region. Basing on the first-order reactions, the annealing experiments of ESR show that there are fast and slow annihilation processes of the irradiation defects for neutron-irradiated Li4SiO4. The activation energies for the two processes are0.18and0.57eV respectively. The fast annihilation process may be attributed to diffusion of trapped electrons into defects, and the slow annihilation process may be attributed to the annihilation of E’-center via recovering oxygen.Isochronal annealing experiments of ESR for neutron-irradiated Li4SiO4are performed to contrast the annihilation process of irradiation defects and the tritium release process. It is observed that the tritium release rate speeds up when the annihilation process of the irradiation defects on the verge of ending, thus the correlation between the slow annihilation process and the tritium release is speculated. According the difference of activity energy of the slow process and the diffusivity activity energy of oxygen, two theoretical models are suggested to interpret the mutual action of oxygen vacancy and tritium. One model supposes that the recovering oxygen ions triggers the tritium release, but a relaxation energy of-0.60eV will be produced after the recover of oxygen ions. The other model considers that he slow annihilation process is related to the diffusion of hydroxyl radical and the recovery of oxygen owing to the same energy barrier of0.56-0.58eV for both the slow annihilation process and the hydroxyl diffusion.The effect of surface defects on the tritium release is simulated by D2+implantation. The chemical states and the deuterium desorption behavior of Li4SiO4implanted with3KeV D2+are investigated by in-situ X-ray photoelectron spectroscopy, the FT-IR spectroscopy and the thermal desorption spectroscopy. The experimental results show that deuterium implantation can change the chemical states on the surface of Li4SiO4, and produce many irradiation defects and binding groups, such as defects trapping D to form Li-D, therefore the thermal desorption behavior of molecular deuterium is related to the surface states. The un-O-D states of deuterium implanted into the sample convert to O-D states during the heating process, which has some relationship with the thermal desorption behavior of deuterium water (HDO). The above results are helpful to understand the tritium release behavior of lithium breeder affected by irradiation defects.The above results are suitable to the experiments with low dose irradiation, but can not extend to reflect the tritium release and defects behavior in the fusion blankets.更多还原