Probing The Ductile-To-Brittle Transition in BCC Blanket Alloys

聚变反应堆内部的条件异常恶劣，包括高温和严重的中子辐射水平。在托卡马克反应堆中，与聚变等离子体直接相邻的是包层，这是一个至关重要的结构。该保护层不仅保护托卡马克的超导磁体和支撑结构免受高温和高能中子的影响，而且还将中子轰击它的动能转化为热量用于发电。所有候选材料的毯子是金属合金，拥有体心立方（BCC）晶体结构。具有这种结构的合金经历了从高温下的韧性机械行为（高韧性）到低温下的脆性机械行为（低韧性）的变化，见左图。这种韧脆转变（DBT）的温度随着辐照而升高，实际上降低了材料的断裂韧性并限制了部件的使用寿命。DBT温度的这种增加是核裂变反应堆共有的问题，在核裂变反应堆中，它也是寿命限制，并且是反应堆寿命延长的关键材料参数。典型地，DBT和转变温度使用宏观断裂测试（例如夏比冲击测试）来测量，其结果然后用于预测设备部件的寿命。然而，这种方法需要大量的材料，并且通常很难获得这种量的已经受辐射损伤的材料。这在聚变反应堆中是一个更大的问题，在那里根本不存在代表性的辐照材料，并且将使用新的合金，没有历史测试数据。本计画旨在利用高解析度数位影像相关技术（HRDIC），探讨BCC合金韧脆转变的基本原理。该技术由曼彻斯特大学首创，不仅需要很少的材料，而且能够提供有关材料内部变形模式的详细定量信息，原则上与DBT相关。该项目将测试HRDIC可用于测量合金中的DBT的论文，因为转变涉及变形行为的变化。目标包括：-发展HRDIC技术应用于BCC合金的实验方面（该技术迄今为止仅广泛用于具有不同晶体结构的合金）。- 对简单模型系统进行室温HRDIC研究（例如，纯Fe）在延伸到候选熔合合金（例如，EUROFER 97）。- 扩展HRDIC研究，以包括（i）在低温下和（ii）辐照后的变形研究。该项目将使用曼彻斯特大学和亨利罗伊斯研究所的先进表征设施。所有辐照均将在道尔顿坎布里亚工厂进行。如果成功，该项目将建立一种评估DBT的新方法，这不仅可以彻底改变我们评估用于聚变反应堆的候选合金的能力，而且还可以增强目前确定老化裂变反应堆结构完整性的方法。

Conditions inside fusion reactors are exceptionally hostile, involving both high temperatures and severe levels of neutron irradiation. Directly adjacent to the fusion plasma in a tokomak reactor is the blanket, a construction of critical importance. The blanket not only protects the tokomak's superconducting magnets and supporting structures from high temperatures and energetic neutrons, but also converts the kinetic energy of the neutrons bombarding it to heat for power generation. All the candidate materials for the blanket are metallic alloys that possess the body-centred cubic (BCC) crystal structure. Alloys with this structure undergo a change from ductile mechanical behaviour (high toughness) at high temperatures to brittle mechanical behaviour (low toughness) at low temperatures, see left-hand image. The temperature of this ductile-to-brittle transition (DBT) increases with irradiation, in effect decreasing the fracture toughness of the material and limiting the useful life of components. This increase in DBT temperature is an issue shared by nuclear fission reactors, where it is also life limiting and is the key material parameter in reactor life extension. Typically, the DBT and the transition temperature are measured using macroscopic fracture tests, such as the Charpy impact test, the results of which are then used to predict the life of plant components. However, such methods require large volumes of material, and it is usually very difficult to obtain such quantities of material that have undergone irradiation damage. This is an even bigger issue in fusion reactors, where representative irradiated material simply does not exist, and where new alloys will be used, for which no historical test data exists. This project aims to investigate the fundamentals of the ductile-to-brittle transition in BCC alloys using high-resolution digital image correlation (HRDIC). Pioneered at the University of Manchester , this technique not only requires very little material, but is able to provide detailed quantitative information about the pattern of deformation inside a material which is, in principle, related to the DBT. The project will test the thesis that HRDIC can be used to measure the DBT in alloys, since the transition involves a change in the deformation behaviour. Objectives include:- Developing experimental aspects of the HRDIC technique for application to BCC alloys (the technique has only been used extensively on alloys with different crystal structures so far). - Conducting room-temperature HRDIC investigations on simple model systems (e.g., pure Fe) before extending to candidate fusion alloys (e.g., EUROFER 97). - Extending the HRDIC studies to include investigations of deformation (i) at cryogenic temperatures and (ii) following irradiation. The project will use advanced characterisation facilities at the Univeristy of Manchester and the Henry Royce Institute. Any irradiations will be carried out at the Dalton Cumbria Facility. If successful, the project will establish a new method for assessing the DBT, which could not only revolutionise our ability to assess candidate alloys for use in fusion reactors, but may also enhance current methodologies of determining the structural integrity of ageing fission reactors.