Advanced study of tritium recovery from molten salt Flibe blanket in fusion reactor

Flibe （2LiF+BeF_2混合熔盐）在高磁场和高密度中子通量下具有良好的导热性和高的氚增殖比，有望成为螺旋型FFHR-2聚变反应堆的先进液体包层流体。但其缺点是熔盐中的氚由于产生氟化氚而难以处理，且氚的回收和封隔困难。因此，在聚变反应堆概念设计中的应用较少，因此对Flibe/氚化学的实验研究较少。在本研究中，我们通过实验研究了Flibe对氢同位素的渗透性以及浸没在Flibe中的Be棒对氢同位素的还原-氧化控制，这将为设计聚变反应堆提供重要的数据。本研究的主要目的是提出一种新的覆盖循环回路系统，该系统由Flibe管道、氚提取系统和…More热交换器组成，该系统可以控制氚的可接受泄漏率低于10 Ci/天，氚浓度低于可接受氚库存的1 ppm。通过本研究的实验和分析，得出以下结论：我们测定了氘通过封闭在双管Ni渗透容器中的熔融Flibe的渗透率。从氘在Flibe中的渗透性、扩散性和溶解度判断，氢同位素在Flibe中表现为H +离子。因此，观察到H^+和F^-离子之间的强相互作用，HF （DF或TF）是在非氧化还原控制的flibe中扩散的主要物质。我们通过Flinak（一种LiF+NaF+KF混合熔盐）测定了H_2氢，氢以分子形式存在于其中。这是因为没有与Flibe情况不同的四方网络。在Flinak中，F^-离子被定位在Li^+、K^+和Na^+离子附近。换句话说，在Flinak中有更少的自由F^-离子，因此，Flinak对应于氧化还原控制良好的flibe。将Be浸入其中，成功地对Flibe进行了氧化还原控制。这是因为存在于非受控Flibe中的自由F离子通过与溶解在Flibe中的Be的物理反应而转变为BeF_2。游离的F^-离子被氧化还原反应消耗，Flibe中的氢以分子形式存在。因此，当be用作中子倍增器时，氚将以分子形式存在于氧化还原控制良好的聚变反应堆包层中。为了使1GW核聚变反应堆的氚泄漏率低于可接受水平（10 Ci/天），我们需要一个由低渗透材料制成的复杂的氚回收系统。给出了氚高回收的氚透窗、鼓泡塔和喷雾塔的设计方程，并利用该方程对各装置进行了设计。少

Flibe (2LiF+BeF_2 mixed molten salt) is expected to be an advanced liquid blanket fluid for a helical-type DEMO fusion reactor called FFHR-2 because it has an ability of good thermal conductance and high tritium breeding ratio under high magnetic field and high-density neutron flux. However, it has disadvantages that it is difficult to handle tritium in the molten salt because of generation of tritium fluoride and the recovery and enclosure of tritium are difficult. Therefore, there was less application to conceptual design of fusion reactors and therefore less experimental studies on Flibe/tritium chemistry. In the present study, we investigated experimentally the permeability of hydrogen isotopes through Flibe and reduction-oxidation (redox) control by Be rods immersed in Flibe that will provide important data to design fusion reactors. The main purpose of the present study is to propose a new blanket circulation loop system composed of Flibe ducts, a tritium extraction system and a … More heat exchanger that can control to a rate lower than 10 Ci/day of acceptable tritium leak rate and a tritium concentration lower than 1 ppm of acceptable tritium inventory. The following findings were obtained from the experiment and analysis of the present study :1.We determined the permeability of deuterium through molten Flibe enclosed in a dual-tube Ni permeation vessel. Judging from the permeability, diffusivity and solubility of deuterium in Flibe, hydrogen isotopes are resent as a H^+ ion in Flibe. Therefore strong interaction between H^+ and F^- ions was observed, and HF (DF or TF) is a main species diffusing in non-redox controlled Flibe.2.We determined hydrogen H_2 through Flinak (a LiF+NaF+KF mixed molten salt), and hydrogen was present as a molecular form in it. This is because there is no tetragonal network of BeF_4^<2-> different from the Flibe case. F^- ions in Flinak were localized in the vicinity of Li^+, K^+ and Na^+ ions. In other words, there are much less free F^- ions in Flinak, and therefore Flinak corresponds to the well redox-controlled Flibe.3.Flibe was successfully redox-controlled by Be immersed in it. This is because free F- ions that were present in non-controlled Flibe were changed to BeF_2 by the reaction with Be dissolved in Flibe physically. The free F^- ions were consumed by the redox reaction and hydrogen in Flibe was present in a molecular form. Consequently, tritium will be present in a molecular form in well redox-controlled fusion reactor blankets when Be is used as a neutron multiplier.4.In order to maintain the tritium leak rate in a 1GW fusion reactor lower than the acceptable level (10 Ci/day), we need a sophisticated tritium recovery system made of low permeable materials. We provided design equations for tritium permeable window, bubbling tower and spray tower for high tritium recovery, and each apparatus was designed using the equations. Less