Tritium removal from molten salt media in nuclear fusion and fission processes

熔融盐(如Flibe)作为聚变反应堆的包层材料越来越受到人们的关注。许多拟议的包层材料用于将能量转换为用于发电的热能，如果回收，还可以作为氚增殖材料，为聚变生命周期提供氚自给自足的潜力。熔盐作为储能介质也受到了极大的关注，因为它们具有固有的安全特性，这对核系统和非核系统都是有利的，但这些盐不断地产生氢的放射性同位素；在核聚变和裂变过程中，熔盐从Flibe熔盐中产生氚。这是一个重大的设计挑战，因为需要去除和分离氚，这样它才不会污染盐系统。因此，在熔盐系统中吸收氚等氢同位素，需要仔细考虑核裂变和聚变反应堆的下游氚管理系统。从熔融介质中去除氚在回收作为燃料以供未来部署聚变系统方面以及在报废、退役和废物管理中损害熔盐方面也是有意义的。因此，该项目将探索氚在熔融盐中的化学行为和传输，以便为清除策略提供依据，并研究提供高效、可靠和安全的氚回收以供再利用的方法。

Molten salts (e.g. FLiBe) are of growing interest as blanket materials for fusion reactors. Many of the proposed blanket materials that are used to convert the energy to heat for electricity generation, can also act as a tritium breeder material providing the potential of tritium self -sufficiency to the fusion life-cycle if recovered. Molten salts have also received significant attention for energy storage media as they possess inherent safety features that are advantageous in nuclear and non nuclear systems, however these salts continuously produce the radioactive isotope of hydrogen; tritium, from the FLiBe molten salt under nuclear operations across both fusion and fission. This presents a major design challenge, as tritium needs to be removed and separated so that is does not contaminate the salt system. Therefore the uptake of hydrogen isotopes such as tritium in molten salt systems, requires careful consideration of downstream tritium management systems in nuclear fission and fusion reactors. Tritium removal from molten media is also of interest both in terms of recovery as fuel for the future deployment of fusion systems and detritiation of the molten salt for end of life decommissioning and waste management. This project will therefore explore the chemical behaviour and transport of tritium in molten salts to inform remova strategies and to investigate methods to provide efficient, reliable, and safe recovery of tritium for reuse.