Degradation of thermal conductivity in tungsten alloys under irradiation

辐照下钨合金导热系数的降低

• 批准号: 2135304
• 金额: --
• 依托单位: University of Manchester
• 依托单位国家: 英国
• 项目类别: Studentship
• 财政年份: 2018
• 资助国家: 英国
• 起止时间: 2018 至 无数据
• 项目状态: 已结题
• 来源: https://gtr.ukri.org/projects?ref=studentship-2135304
• 关键词: Degradation; thermal; conductivity; tungsten; alloys

Tungsten-based alloys are currently the principal candidates for plasma facing components in fusion reactors due to their unique combination of properties such as neutron irradiation resistance, low sputtering yield, good high-temperature strength and thermal conductivity. The relatively high ductile-to-brittle transition temperature of tungsten, which even increases under neutron irradiation, is one of the main problems of using tungsten as plasma-facing material. Alloying tungsten with elements such as Ti, V or Ta improves the material's ductility. Unfortunately, our limited neutron irradiation data proves the degradation in thermal properties of pure W, and the effect of irradiation in thermal performance of those alloys remains unexplored. Changes in thermal conductivity due to irradiation damage is one of the more pressing areas of information required for the deployment of fusion reactor technology. The aim of the project is to detect changes in thermal conductivity in irradiated tungsten-based alloys, and to correlate those changes in conductivity to the lattice defects induced by radiation fields simulating those expected in the proximity of the hot plasma in fusion reactors. This project involves the use of intense ion beams to simulate the neutron damage that W alloys will be experiencing in future reactor conditions. Ion irradiation allows us to achieve the damage doses that will be experienced inside the reactor after several years of operation. The student will characterise the damaged areas of the material at the atomic-to-nanometer scale using advanced analytical electron microscopy and positron annihilation spectroscopy. The nature and concentration of radiation-induced lattice defects, such as dislocation loops or voids, will be correlated with variations in thermal conductivity. The student will have opportunities to learn fundamental radiation damage mechanisms to nuclear materials at an advanced level, to use upfront particle accelerators at large scale international user facilities, and to become an advanced user of electron and positron probe techniques.

钨基合金具有耐中子辐照、低溅射产额、良好的高温强度和导热性等独特的综合性能，是目前聚变堆中面向等离子体部件的主要候选材料。钨的韧脆转变温度相对较高，在中子辐照下甚至会升高，这是使用钨作为等离子体面对材料的主要问题之一。钨与Ti、V或Ta等元素的合金化提高了材料的延展性。不幸的是，我们有限的中子辐照数据证明了纯W的热性能的退化，并且辐照对这些合金的热性能的影响仍未被探索。辐射损伤引起的热导率变化是部署聚变反应堆技术所需信息的一个更为紧迫的领域。该项目的目的是检测辐照钨基合金的热导率变化，并将这些电导率变化与辐射场引起的晶格缺陷相关联，模拟聚变反应堆中热等离子体附近的预期缺陷。该项目涉及使用强离子束来模拟钨合金在未来反应堆条件下将经历的中子损伤。离子辐照使我们能够达到反应堆运行几年后将经历的损伤剂量。学生将使用先进的分析电子显微镜和正电子湮没光谱学在原子到纳米尺度上对材料的损坏区域进行分析。辐射诱导的晶格缺陷的性质和浓度，如位错环或空隙，将与热导率的变化相关。学生将有机会学习先进水平的核材料的基本辐射损伤机制，在大规模国际用户设施中使用前沿粒子加速器，并成为电子和正电子探测技术的高级用户。