减压蒸馏过程中钍对氟化物熔盐回收的影响研究

Molten-Salt Reactor (MSR) is a design of an advanced reactor system from the GEN IV family working in thermal or epithermal neutron spectrum and using thorium or transuranium fuel in the form of molten fluorides. The MSR fuel cycle with integrated reprocessing represents one of the potential ways both for significant decrease of total amount of radioactive wastes for final deposition and for utilization of nuclear energy for electricity and heat production as effectively as possible. The most effective use of thorium is as nuclear fuel in MSR. To achieve the thorium-uranium cycle and recovery the nuclear material and carrier salt LiF-BeF2 from spent fuel are the major targets for MSR. Because of the difference volatility, carrier salt can be recovered using molten salt distillation method. It was observed that the evaporation rate was lower with the melts containing ThF4. However, the influence of thorium on the salt recovery efficiency was unknown yet..In this project, the distillation method will be established to evaluate the recovery efficiency of carrier salt. The evaporation behavior will be investigated in fluoride system containing ThF4. The evaporation rate and recovery ratio will be compared with the pure fluoride system. Moreover, the decontamination factor of fission products in condensate salt will be deduced when the mixture salt contains various salt, for example SrF2, NdF3, CsF and ThF4. The structure of thorium compound in molten salt will be further investigated by high temperature Raman, XRD, XAFS, molecular dynamics simulation methods and so on. An optical method, namely precipitation and distillation combination method，will further be performed to improve the purity of recovered salt when the thorium concentration reaches a very high level during the distillation process. The results will be helpful for the separation of recycling carrier containing thorium fluoride molten salt .

熔盐堆采用热容量大、传热性能好的氟化物熔盐作为冷却剂和燃料载体，无需制作燃料元件，被认为是最适合钍资源利用的反应堆型。核燃料出堆后成分复杂，包含U，Th，裂变产物和载体盐LiF-BeF2，提取有用的核燃料、分离裂变产物、回收载体盐，既可提高堆运行的经济学，又可以减少废盐量。常利用蒸馏法回收价格昂贵的Li-7，但钍的存在，尤其是处理过程钍富集到较高浓度时，对熔盐回收率、熔盐纯度等的影响不清楚。本项目旨在研究核燃料中钍对载体熔盐回收的影响。首先研究纯氟盐体系中，考察熔盐组成、钍浓度等条件下钍对载体盐蒸发行为的影响；以及含有模拟裂变产物的混合熔盐中，钍对熔盐中裂变产物去污的影响；进一步利用Raman，XRD，分子模拟、同步辐射XAFS等手段探讨不同浓度的钍在熔盐蒸馏富集过程中可能存在的微观结构，探索利用沉淀-蒸馏耦合法降低熔盐中钍浓度，提高钍、稀土去污因子和回收盐纯度的可行性，为真实含钍氟化物熔盐的载体盐回收提供思路。

熔盐堆采用传热性能良好的氟化物熔盐作为冷却剂和燃料载体，无需制作燃料元件，被认为是最适合钍资源利用的反应堆型。核燃料经反应堆运行后成分复杂，既包含为燃烧完全的U、Th，也有种类繁多的裂变产物，还有占主要组成的载体盐LiF-BeF2，提取有用的核燃料、分离裂变产物、回收载体盐，既可提高反应堆运行的经济学，又可以减少废物量。蒸馏法是利用组分间挥发性的差异的物理分离方法，近年来被广泛用于熔盐的回收和纯化。但钍的存在，尤其是处理过程钍富集到较高浓度时，对熔盐回收率、熔盐纯度等的影响不清楚。本项目研究氟化物核燃料中钍对氟化物载体熔盐回收的影响，首先在纯氟盐体系FLiNaK，FLiBe中，考察了熔盐组成、ThF4浓度等条件下钍对熔盐蒸发行为的影响.对于高浓度ThF4的体系，进行分段蒸馏，分析各个阶段残留盐、冷凝物组成，阐明蒸发规律。继而模拟含有裂变产物的混合熔盐中，在冷实验的基础上，利用白光中子源辐照ThF4和UF4，开展含辐照钍铀的示踪实验验证，进一步确定钍对熔盐中裂变产物去污的影响。结合实验结果，利用Raman，XRD，理论模拟、同步辐射XAFS等手段探讨蒸馏过程中稀土、钍的价态、配位结构、化学形态变化，揭示钍在熔盐中的存在形态及结构，推测ThF4的蒸发行为及与载体盐的相互作用机制。同时探索利用沉淀-蒸馏耦合法降低熔盐中钍浓度，提高钍、稀土去污因子和回收盐纯度的可行性，为真实含钍氟化物熔盐的载体盐回收提供理论依据。