Effect of water adsorption on tritium release from Li4SiO4

As a ceramic material proposed for tritium breeding in a fusion reactor blanket, lithium orthosilicate (Li₄SiO₄) is being examined in view of the influence of water uptake on tritium release behavior. In this work, out-of-pile tritium release experiments were performed on Li₄SiO₄ samples that were transferred and stored under different moisture environments. The water content was measured on the samples that were treated in similar conditions. Effects of water adsorption on the chemical form and temperature of released tritium were investigated. It is found that with the water content increases, the gaseous tritium fraction decreases and the proportion of low-temperature desorption of HTO increases. The results of this study can be used later for engineering and design activities for fusion reactor blankets.     

基于蒙特卡罗方法的固态氚增殖剂聚变中子辐照损伤行为分析

实验包层模块(TBM)是聚变反应堆最重要的组件之一,作用是产氚和能量提取。锂陶瓷具有良好的化学稳定性、热机械性能、产氚性能以及可在更高温度下使用等特点,被认为是聚变堆包层最具吸引力的氚增殖剂材料。中国ITER-TBM设计方案采用了氦冷固态氚增殖剂(HCCB)TBM结构,其聚变环境下的辐照损伤行为可为中国HCCB TBM结构设计提供支持。针对固态氚增殖剂聚变中子辐照损伤问题,利用蒙特卡罗模拟,对比分析了Li_4SiO_4和Li_2TiO_3的中子辐照离位损伤和嬗变气体损伤。结果表明:在相同的服役时间下,Li_4SiO_4比Li_2TiO_3将产生更多的嬗变气体,且在高6 Li丰度情况下,其中子辐照损伤也更严重,会产生更高的损伤剂量和更大的损伤截面。但是,嬗变气体所造成的空位损伤Li_2TiO_3要比Li_4SiO_4严重;对两种陶瓷材料来讲,氦损伤效应均强于氚损伤效应。     

Recovery of Tritium in Room Air by Precious Metal Catalyst with Hydrophilic Substrate

The catalytic oxidation and adsorption method is considered to be a potential and reliable measure to recover tritium released into room air in fusion power plants. The activity of precious metal catalysts that are expected to be useful in recovery of tritium released into the room air is affected by moisture in the air, and tritium in the gas phase can be captured into the catalyst substrate not only through adsorption but also through isotopic exchange reaction. The simulation study on tritium behavior in the catalyst bed was carried out quantitatively on the basis of experimental results. It is confirmed by the simulation study that the installation of the preadsorption bed decreases water vapor before the gas is passed through the precious metal catalyst bed; this is an effective countermeasure against the deterioration of the catalytic oxidizing performance caused by moisture. It is also shown that large amounts of tritium can be captured by the catalyst itself when the preadsorption bed is introduced.     

Calculation of Breakthrough Curve of Multicomponent Hydrogen Isotopes Using Cryosorption Column

The effective tritium system should be designed to recover tritium from DT reactor blanket sweep gas in a form that is easy to transfer to the main fuel cycle. The cryosorption method using a porous adsorbent at the temperature of liquid nitrogen is one of the candidate processes for extracting tritium from the sweep gas. For designing of a cryosorption column for recovery of tritium from hydrogen-swamped helium sweep gas, it is necessary to predict the breakthrough curve for mixture of multicomponent hydrogen isotopes in helium. In this report, a method to calculate the breakthrough curve at cryosorption of multicomponent hydrogen isotopes on molecular sieve 5A, molecular sieve 4A or activated carbon at 77.4 K is presented.     

Release behavior of bred tritium from LiAlO2

The release behavior of bred tritium to the blanket purge gas is mainly controlled by such bulk phenomena as tritium forming reaction, diffusion of tritium in grain, interaction of tritium with irradiation defects, and absorption together with such surface phenomena as adsorption, isotope exchange reaction between molecular form hydrogen in purge gas and tritium on grain surface (isotope exchange reaction 1), isotope exchange reaction between water vapor and tritium on grain surface (isotope exchange reaction 2), and water formation reaction at addition of hydrogen. Following the observation of the present authors that the isotope exchange reaction 2 is much faster than the isotope exchange reaction 1, the release curve of bred tritium obtained at purge with humidified gas was used for estimation of the effective diffusivity of bred tritium in LiAlO₂. Then, the effective diffusivity of tritium in grain of LiAlO₂ is obtained as D_T = 2.5 × 10⁻⁷exp(−110 [kJ]/RT) [m²/s]. This equation gives the larger diffusivity than any other diffusivity presented so far because the mass transfer resistance at the grain surface is expected to be eliminated in the estimation procedure of this study.     

Out-of-pile tritium release study on Li4SiO4 pebbles from TRINPC-I experiments

Out-of-pile tritium release examinations of irradiated Li₄SiO₄ pebbles were performed in TRINPC-I experiments for evaluating material performance and verifying the system design. To generate tritium the specimens were irradiated with neutrons. Li₄SiO₄ pebbles were made by a freeze-drying method. In the experiments, concentrations of tritium in the form of tritium gas (HT + T₂) and tritiated water (HTO + T₂O) in the outlet streams of a reactor tube were measured separately with an ionization chamber and a liquid scintillation radiometer. The results show that the percentage of tritium gas (HT + T₂) and tritiated water trapped by the breeder pebbles were about 72% and 19% of totally released tritium, respectively. Thus, more tritium was released in the form of tritium gas in this work. In addition to tritium trapped by the breeder pebbles, the amount of free tritium was also measured by breaking on-line a quartz capsule containing Li₄SiO₄ pebbles, the percentage of which was 9% of totally released tritium. The temperature peaks of tritium gas mainly appeared at about 477 °C and 654 °C, while the temperature peak of tritiated water appeared at about 402 °C, under which most of tritiated water released.     

Studies of Ceramic Fuels with Use of Fission Gas Release Loop, (IV)

In-pile release of fission gas from sintered UC pellets in the presence of 8–230 ppm of water vapor in the He sweep gas was measured over the temperature range of 160°–1,000°C. A very complex release behavior was observed and the mechanisms of release were deduced from the manner in which the release depended on the decay constant. It was established that the release of short-lived fission gases during irradiation was controlled mainly by pseudo-recoil, while chemical reaction between UC and water vapor, as well as knock-out, appeared to contribute much more significantly in the case of the longer-lived fission gases. The release of fission gas after reactor shutdown was shown to be governed by the UC-H₂O reaction. The ratio of the release due to this reaction in reference to the total release was found to be dependent not only on the concentration of the water vapor but also on the amount present of the accumulated reaction products. Also, a discussion is given on the inordinately high release of ^135mXe observed at 600°C.     

Design and R&D progress of Korean HCCR TBM

Korea plans to test a Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in ITER. The HCCR TBM adopts a four sub-module concept considering the fabricability and the transfer of irradiated TBM for post irradiation examination. Each sub-module has seven-layer breeding zone, including three neutron multiplier layers packed with beryllium pebbles, three lithium ceramic pebble bed packed tritium breeder layers, and a reflector layer packed with graphite pebbles. Based on this configuration, neutronic and electromagnetic calculations were performed and their results were applied for the conceptual design of HCCR TBM that considers manufacturing feasibility. Also, a design and safety analysis of HCCR Test Blanket System (TBS) was performed using integrated design tools modifying nuclear system codes for helium coolant and tritium behavior evaluation. The Advanced Reduced Activation Alloy (ARAA) is being developed as a structural material. A total of 73 candidate ARAA alloys were designed and their out-of-pile performance was evaluated. The graphite pebbles as the neutron reflector were fabricated by using mechanical machining and grounding method with the surface coated with SiC. The hydrogen permeation characteristics of structural materials were evaluated using the Hydrogen PERmeation (HYPER) facility. The recent design and R&D progress on these areas are addressed in this paper.     

Diffusion analysis on the thermal release of tritium from a neutron absorbing material

Tritium released from neutron irradiated borosilicate glass was determined by a specially designed sampling system and a liquid scintillation counter at temperatures in the range of 200–700°C. It was found that the chemical form of tritium released was tritiated water (HTO, T₂O) for the most part. Tritium produced in the glass would react with oxygen to form OT and diffuse out by a similar mechanism as the molecular diffusion of water in glasses. The diffusion coefficient of tritiated water in borosilicate glass obtained is expressed by D (cm²/s) = 5.3 × 10⁻⁴ exp( −128 kJ/mol)/RT). It is concluded from the diffusion analysis that the greater part of tritium produced in a neutron absorber, which is made of borosilicate glass, would remain in the glass for a few years of irradiation.     

Release behavior of tritium from graphite heavily irradiated by neutrons

The release behavior of tritium formed in graphite has been investigated as a function of radiation damage by means of isochronal annealing of samples heavily irradiated by neutrons. The lithium impurities in graphite were estimated as the source of tritium formation. The main chemical form of released tritium was hydrogen accompanied by a small quantity of methane. No other hydrocarbons could be detected. Tritiated water was always measured, but the formation mechanism was experimentally confirmed as the secondary oxidation of released HT molecule. The release spectrum of tritium in isochronal annealing was shifted to a higher heating temperature with the increase of the neutron fluence received by the graphite crystal. A relationship was established between the amount of tritium released up to a certain temperature and the degree of graphitization of the sample.     

Estimation of the Tritium Behavior in the Pebble Type Gas Cooled Reactor for Hydrogen Production

In Korea, a nuclear hydrogen program has been established to develop and demonstrate mass production system for hydrogen generation. The objective of this study is to establish the evaluation procedure for predicting the tritium behavior in the 300 MWth Pebble type gas cooled reactor which is the one of the candidate reactors for nuclear hydrogen development and demonstration plant. The tritium generated by the fission reaction can be leaked to the helium coolant from the coated ceramic particles and fuel elements. The annual total release rate of the tritium is estimated as 0.47% from the fuel kernel to the helium coolant by the numerical method. Tritium attributed by ⁶Li existing as impurities in the reflector can be released to the helium coolant by the diffusion process and the total annual release rate of the tritium is estimated as 5.3% through the reflector to the helium coolant. Based on the Siverts' law, tritium permeation from the primary coolant to the hydrogen production system is also evaluated and the result is calculated as 76?0.23 Bq/g-H₂ with respect to the PRF (Permeation Reduction Factor= 10?1000) in case of the normal operation of the 300 MWth Pebble type reactor.     

Postirradiation Annealing of Boron Carbide Pellet Irradiated in Fast Breeder Reactor

This paper reports some irradiation effects and recovery behavior of neutron irradiated boron carbide pellets that were used as control rod elements in the Enrico Fermi Fast Breeder Reactor. Measurements were carried out on changes in lattice parameters, thermal expansion, helium release, elastic moduli and microstructure observations by annealing the irradiated pellets at elevated temperatures. The increase in unit cell volume of B₄C upon irradiation was found to be 0.22%. The recovery in lattice parameter began at around 500°C and completed at 1,000°C. It was found that the pellet showed a sharp increase in a dimensional change at about 700 to 800°C with a large amount of helium release, and the pellet which showed larger swelling released smaller amount of helium.     

中子辐照后6Li-Al合金靶片中氚的测量方法研究

建立了流气式系统捕集和化学转移法测量^6LiAl合金靶片在中子辐照过程中的渗漏氚和辐照后靶片中总氚量的方法。渗漏氚的捕集方法是：在流气式系统中，用含少量氢的惰性载气将渗漏氚载带出来，经高温催化氧化后被乙二醇鼓泡器捕集；靶片中的氚转移法是用NaOH溶液将合金靶片溶解，气相中的氚采用渗漏氚的捕集方法捕集，液相中的氚则蒸馏到馏分中。最后用液体闪烁计数器分别测量乙二醇鼓泡器和馏分中的氚量。测量结果与理论氚产量基本相符。     

硅酸锂球床传热传质特性与吹扫气体流动的DEM-CFD耦合模拟

氦冷固态增殖包层是中国聚变工程实验堆（CFETR）的3种候选包层概念之一,氚增殖球床是包层的核心部件,采用硅酸锂颗粒作为氚增殖材料。球床结构对氚在球床内的输运行为及流动和传热均有重要影响。本文基于离散单元法（DEM）生成了满足氚增殖球床填充率要求的随机堆积结构,通过CFD计算获取了球床结构下氚在吹扫气体内的等效扩散系数及吹扫气体的流动特性,包括速度分布、压力分布及进出口压降;开展了外加热流及有内热源两种工况下球床等效导热系数的模拟。计算结果表明,球床结构下氚在吹扫气体内的等效扩散系数为二元气体扩散系数的40%;受球床结构影响,球床内存在流动迟滞区,壁面出现流动加速;拟合得到Ergun方程的黏性阻力系数C1=87;有内热源工况下的球床等效导热系数低于外加热流工况下的球床等效导热系数。     

Tritium release experiments on Li4SiO4 pebbles from TRINPC-I experiments: Effects of water adsorption and hydrogen gas

Out-of-pile tritium release experiments under different water uptake contents and purge gas chemistry were performed on Li₄SiO₄. Water measurement was performed on samples under different experimental procedures. It was found that water was adsorbed on the sample during its transferring and storage process. A strong dependence of tritium release behavior on water uptake was determined. By doping H₂ in the sweep gas, the formation of water in orthosilicate was observed in addition to the isotope exchange reaction with H₂ gas. Thermal desorption peaks of the water formation reaction and H₂ isotope exchange reaction appeared at 668 °C and 463 °C, respectively, at ramping rate of 5 °C/min.     

氚在RAFM钢表面的吸附研究

采用气相吸附法研究了室温下RAFM钢表面对氚的吸附与释放行为,并使用316L钢、1Cr18Ni9Ti钢进行了对照实验。结果表明,RAFM钢表面的氚吸附与释放性质与316L钢、1Cr18Ni9Ti钢的非常相似,相同表面状态的样品,在相同实验条件下的吸附氚量相差不超过50%。可推测,未经深度除水处理的RAFM钢暴露于氚后,表面会形成富氚层,浓度远高于基体溶解氚,厚度不大于10 μm。表面氚的形态以化学吸附和物理吸附的氚化水为主,约占90%以上。室温下RAFM钢表面吸附的氚在干燥气氛中的释放非常缓慢,但遇水会因氚-水间的同位素交换而加速释放。     

Formation of water in lithium ceramics bed at hydrogen addition to purge gas

It has been proposed to make use of the isotope exchange reaction for enhancement of the tritium release rate from the blanket material by adding hydrogen isotopes to the purge gas. However, it is found by the present authors that formation of water in the lithium oxide bed occurs when hydrogen isotopes are introduced to the He sweep gas. The amount of water generated in various lithium ceramics beds at hydrogen addition to purge gas is discussed in this paper. The reaction rate of the water formation reaction is also discussed. The water formation reaction in each lithium ceramics bed is considered to be second order reaction which is affected by the concentration of active point as the oxygen supplier on the lithium ceramics and the concentration of hydrogen in the purge gas.     

堆内辐照在线产氚氦冷陶瓷氚增殖剂包层球床组件热工流体力学计算分析

在研究堆内进行辐照在线产氚试验是ITER计划专项国内配套研究项目之一。本工作主要针对研究堆内辐照氦冷陶瓷氚增殖剂包层（简称陶瓷）球床组件的试验技术要求，评估辐照陶瓷球床组件设计方案的可行性。通过对不同的陶瓷球床组件结构参数和组件在堆内的不同辐照位置，进行热工流体力学设计计算，得到满足要求的入堆辐照陶瓷球床组件设计方案。     

Catalytic Oxidation of Tritium in Wet Gas

Use of precious metal catalyst is recommended in the tritium recovery system because it can oxidize tritium at ambient temperature. The ability to operate at the ambient temperature without preheating and past cooling is a large advantage in the ease of system operation. It is observed in this study, however, that the catalytic oxidation characteristics of the precious metal catalysts are largely affected by the water vapor to such extent that almost no oxidation rate of tritium is expected in the wet gas. Effect of the water vapor on the oxidation rate is quantitatively discussed based on data obtained in this study and an emergency cleanup system from the room air with pre-adsorption bed is proposed.     

中国氦冷固态增殖剂实验包层模块材料研究进展

在未来核聚变反应堆中,为补充氚的消耗,需要在核聚变堆的包层中进行氚的在线增殖,以维持核聚变反应的持续进行。为验证这一关键技术,在国际热核聚变实验堆（ITER）上开展了ITER TBM计划（实验包层项目）。作为ITER计划成员方之一,中方以中国氦冷固态增殖剂实验包层模块（HCCB TBM）概念参与ITER TBM计划。HCCB TBM现今进入初步设计阶段,而材料的制备技术和性能数据是支撑其结构设计、安全分析和服役工况评估的基础。本文综述和分析了HCCB TBM结构材料低活化铁素体/马氏体钢（RAFM钢）与功能材料氚增殖剂和中子倍增剂的研究现状,并对这些材料下一步的研究方向进行了展望。