Adsorption and Desorption Behavior of Tritiated Water on the Piping Materials

The behavior of tritium on the surface of various piping materials must be investigated for establishment of the safety confinement technology of tritium or for development of the effective fuel handling technology in a D-T fusion reactor, because tritiated water or gaseous tritium is captured on the piping surface through adsorption or isotope exchange reaction. The present authors carried out the water adsorption and desorption experiments on 304 stainless steel, copper, and aluminum in the temperature range from 5 to 100°C and in the partial pressure range of water vapor between 11.8 and 198Pa using a breakthrough method and quantified the amount of water adsorbed and the overall mass transfer coefficients in adsorption and desorption of water. It was observed in this study that aluminum adsorbed more water than stainless steel or copper. It was also observed that the adsorption and desorption rates of water for three materials showed almost the same values. The breakthrough behavior of tritiated water in a 100 m pipe of stainless steel was also evaluated applying the results of this work. It is concluded that water adsorption and desorption reactions influence the behavior of tritiated water in the piping system under the condition where the partial pressure of tritiated water vapor is lower than several pascals.     

D+注入Li4SiO4的表面化学状态及其热解吸行为

锂陶瓷氚增殖剂的氢同位素行为是聚变堆固态产氚包层关心的重要课题。本文将3 keV D⁺注入Li₄SiO₄,采用X射线光电子能谱在线分析注氘前后材料表面的化学状态,同时采用热解吸谱（TDS）实验技术,研究注氘后Li₄SiO₄中氢同位素的热解吸行为。实验结果表明：D⁺注入会改变Li₄SiO₄表面的化学环境,产生多种辐照缺陷和化学键合状态;氘滞留量和热解行为受注氘时样品的温度影响较大,可在一定程度上预测产氚包层中氚的滞留行为。     

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.     

低温吸附法分离氦气流中氢同位素的实验研究

采用穿透曲线法研究了液氮温度下活性炭、碳分子筛和碳纳米纤维对氢同位素的吸附等温线和同位素效应。结果表明：活性炭和碳分子筛都是良好的吸附剂，碳纳米纤维对氢同位素的吸附量太低而不具备工程应用的可能性。实验还对吸附剂进行了改性处理，考察了改性吸附剂对氢同位素的吸附性能。     

Determination of the order of surface reactions in Li2O

An analysis of the different surface reactions taking place in Li₂O was performed in order to determine whether adsorption and desorption of tritium are first or second order reactions. Data from BEATRIX-II Phase I and CRITIC-I were used as basis for calculations.It was found that only second order adsorption/desorption on the surface of Li₂O can predict the tritium behavior observed experimentally.     

高温气冷堆用含硼碳对水蒸气的吸附特性研究

高温气冷堆中使用了大量的碳材料，其中含硼碳（BC）因其优秀的物理特性而被大量用于堆芯的结构材料。BC是一种典型的多孔材料，暴露于空气中会吸附水分等杂质，其对水蒸气的吸附和脱附特性将直接影响初装堆芯的除湿过程。通过动态吸附和脱附实验详细测定了BC的水蒸气吸附等温线，使用低相对湿度的吸附数据拟合BET方程，并得到高相对湿度下的BET预测值。研究结果表明，水蒸气在BC表面属于多层物理吸附，脱附曲线较吸附曲线有一定的滞后，水蒸气能深入进材料内部，这也导致在较高的相对湿度下很难达到吸附和脱附平衡。     

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.     

Effect of Lithium Enrichment on the Tritium Breeding Characteristics of Various Breeders in a Fusion Driven Hybrid Reactor

Selection of lithium containing materials is very important in the design of a deuterium–tritium (DT) fusion driven hybrid reactor in order to supply its tritium self-sufficiency. Tritium, an artificial isotope of hydrogen, can be produced in the blanket by using the neutron capture reactions of lithium in the coolants and/or blanket materials which consist of lithium. This study presents the effect of lithium-6 enrichment in the coolant of the reactor on the tritium breeding of the hybrid blanket. Various liquid–solid breeder couples were investigated to determine the effective breeders. Numerical results pointed out that the tritium production increased with increasing lithium-6 enrichment for all cases.     

Study on behavior of tritium in concrete wall

It is required to understand the tritium behavior in concrete for establishment of tritium safety technology of a fusion reactor or a tritium handling facility because the concrete wall is used as the final containment to prevent tritium release to the environment. This paper discusses about the effect of adsorption and diffusion of water and isotope exchange reaction between physically adsorbed water and chemically adsorbed water or structural water. It is known in this study that a large amount of tritium can be trapped to the concrete wall because cement paste has the nature of porous hydrophilic material.     

Thermal hydraulic simulation of reactor of HTR-PM based on thermal-fluid network and SIMPLE algorithm

This paper provides the model development and its verifications for the reactor thermal-hydraulic transient model for the High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM). A thermal-fluid network is constructed to simulate the complexity of the flow and heat transfer structure in the reactor of HTR-PM. SIMPLE algorithm is applied to solve the conservation equations of the thermal-fluid network to simulate the transient behavior of the high speed turbulent helium flow. A calculation process is proposed for coupling the high speed helium flow and the complex heat transfer structure. A FORTRAN code was developed based on the solution method and the thermal-fluid network. Several test cases including two steady states and a Control Rod Withdrawal accident are simulated by this code and the results are compared to those obtained by a safety analysis code, namely THERMIX. The good agreement between the two codes indicates that the proposed model and solution method based on SIMPLE algorithm is reasonable and applicable for simulating the thermal-hydraulic behavior in reactor of HTR-PM.     

Design aspects of the Chinese modular high-temperature gas-cooled reactor HTR-PM

The modular high-temperature gas-cooled reactor (MHTGR) has distinct advantages in terms of inherent safety, economics potential, high efficiency, potential usage for hydrogen production, etc. The Chinese design of the MHTGR, named as high-temperature gas-cooled reactor-pebble bed module (HTR-PM), based on the technology and experience of the HTR-10, is currently in the conceptual phase. The HTR-PM demonstration plant is planned to be finished by 2012. The main philosophy of the HTR-PM project can be pinned down as: (1) safety, (2) standardization, (3) economy, and (4) proven technology. The work in the categories of marketing, organization, project and technology is done in predefined order. The biggest challenge for the HTR-PM is to ensure its economical viability while maintaining its inherent safety. A design of a 450 MWth annular pebble bed core connected with steam turbine is aimed for and presented in this paper.     

Numerical Study on Tritium Behavior by Using Isotope Exchange Reactions in Thermochemical Water-Splitting Iodine—Sulfur Process

One potential problem in the hydrogen production system coupled with the high-temperature gascooled reactor (HTGR) is transmission of tritium from the primary coolant to the product hydrogen by permeation through the heat transfer tubes. Tritium accumulation in the process chemicals in the components of a hydrogen plant, a thermochemical water-splitting iodine-sulfur (IS) process, will also be a critical issue in seeking to license the hydrogen plant as a non-nuclear plant in the future. A numerical analysis model for tritium behavior in the IS process was developed by considering the isotope exchange reactions between tritium and the hydrogen-containing process chemicals, i.e., H₂O, H₂SO₄ and HI. The tritium activity concentration in the IS process coupled with the high-temperature engineering test reactor (HTTR), the HTTR-IS system, was preliminarily evaluated in regard to the effects of some indeterminate parameters, i.e., equilibrium constants of the isotope exchange reactions, permeability of tritium through heat transfer tubes, tritium and hydrogen concentrations in the secondary helium coolant, and the leak rate from the secondary coolant loop. The results describing how the tritium activity concentration changes with variations in these parameters and which component has the maximum tritium activity concentration in the IS process are described in this paper.     

氟盐冷却高温堆氚输运特性数值研究

氚的控制是限制氟盐冷却高温堆（FHR）发展的关键问题，欲实现氚的有效控制，首先需明确氚在熔盐堆一回路中的输运行为。本文阐明了氚在熔盐堆一回路中的输运特性，包括氚的产生及存在形态的分化、石墨对氚的吸附、氚在熔盐中的溶解与扩散以及氚在管壁材料中的渗透等。针对氚在熔盐堆一回路中的输运行为，建立了数学物理模型，基于FORTRAN语言开发了适用于FHR的氚输运特性分析程序TAPAS。通过将实验数据与程序计算结果对比，说明了TAPAS程序计算的合理性和准确性。利用TAPAS对模块化移动式氟盐冷却高温堆（TFHR）中氚的输运特性进行了分析。计算表明，TFHR的初始产氚率约为5.54×10^-8 mol/s，一回路中的氚主要以T₂形式存在，腐蚀反应主要发生在热管段入口处。反应堆运行25 EFPD（等效满功率天）后，石墨吸附氚达到限值。反应堆稳态运行时，T₂向管壁表面的渗透速率可视为常数，其值为8.35 μmol/EFPD。本研究可为FHR的研究设计和辐射防护提供参考。     

Overview of R&D activities on tritium processing and handling technology in JAEA

In JAEA, the tritium processing and handling technologies have been studied at TPL (Tritium Process Laboratory). The main R&D activities are: the tritium processing technology for the blanket recovery systems; the basic tritium behavior in confinement materials; and detritiation and decontamination. The R&D activities on tritium processing and handling technologies for a demonstration reactor (DEMO) are also planned to be carried out in the broader approach (BA) program by JAEA with Japanese universities. The ceramic proton conductor has been studied as a possible tritium processing method for the blanket system. The BIXS method has also been studied as a monitoring of tritium in the blanket system. The hydrogen transfer behavior from water to metal has been studied as a function of temperature. As for the behavior of high concentration tritium water, it was observed that the formation of the oxidized layer was prevented by the presence of tritium in water (0.23 GBq/cc). A new hydrophobic catalyst has been developed for the conversion of tritium to water. The catalyst could convert tritium to water at room temperature. A new Nafion membrane has also been developed by gamma ray irradiation to get the strong durability for tritium.     

Tritium diffusivity in crystal grain of Li2TiO3 and tritium release behavior under several purge gas conditions

It has been reported by the present authors that behavior of tritium release from solid breeder grain is consisted of diffusion in grain, tritium transfer at surface layer and surface reactions on grain surface such as adsorption or isotope exchange reactions. Tritium release curves estimated using the tritium release model gave good agreement with observed tritium release curves from Li₄SiO₄, Li₂ZrO₃ or LiAlO₂.

Tritium release behavior from Li₂TiO₃ under humid purge gas, dry purge gas and dry purge gas with hydrogen conditions is discussed in this study, tritium release curves using the release model that we proposed previously give a good agreements with experimental tritium release curves. Tritium effective diffusivity in the crystal grain of Li₂TiO₃ is also estimated in this study using a curve-fitting method applied to the release curves obtained under the humid purge gas condition. It is discussed that change of color of Li₂TiO₃ surface under hydrogen purge gas condition is observed and this phenomenon might affect tritium release behavior from Li₂TiO₃.     

Tritium release behavior from the graphite tiles used at the dome unit of the W-shaped divertor region in JT-60U

Release behavior of tritium from the graphite tiles used at dome top and inner dome wing in JT-60U was investigated by the thermal desorption method in dry argon, argon with oxygen and water vapor, or argon with hydrogen. It was found that approximately 20-40% of total tritium is left in graphite even after heating to the high temperature above 1000 °C in dry argon. The residual tritium could be removed by exposing the graphite tile to oxygen with water vapor or hydrogen at the high temperature above 1000 °C. The tritium retention of the dome top tile was quantified as 84-30 kBq/cm². The inner dome wing tile had a steep tritium distribution from 8 to 0.1 kBq/cm². It is observed that a measurable amount of tritium existed in the deep site of the graphite tile.     

Surface phenomena at fluid-structure interfaces and their effects on the migration of tritium in controlled thermonuclear reactors

Phenomena that are likely to have a significant effect on tritium migration through the structural materials of fusion reactors include Langmuir adsorption/desorption, isotopic swamping resulting from the presence of large quantities of protium and/or deuerium, passivation of metal surfaces due to impurity layers, and surface fluxing in the presence of reactive fluids lithium and some molten salts. These phenomena are considered in terms of their anticipated relative importance under a variety of conditions that could be encountered in fusion devices. In addition, the utility of the Ion Microprobe Mass Analyzer as a tool for studying surface characteristics is discussed, and results of some recent experimental studies of the effect of surface impurities on the hydrogen permeability of vanadium are presented.     

HTR-PM直流蒸汽发生器的建模与分析

球床模块式高温气冷堆核电站示范工程（HTR-PM）采用两座模块式高温气冷堆带一台汽轮发电机组的技术方案,为了开展其运行特性研究,清华大学核能与新能源技术研究院开发了针对HTR-PM的工程模拟机,其中螺旋管式直流蒸汽发生器的模型还需进一步完善。本文深入分析了螺旋管式直流蒸汽发生器的流动、换热规律,明确了蒸汽发生器一次侧和二次侧的流动与换热模型,通过对稳态工况中分布数据的详细分析,说明了模拟结果的正确性。为适应更多模块的高温气冷堆核电站的运行分析要求,通过网格划分方案的讨论与优化,在保证实时性的前提下,提高了蒸汽发生器中流动与换热模拟的准确性,为下一步采用工程模拟机开展其运行特性研究打下基础。     

Isotope exchange reaction between tritiated water and hydrogen on SiC

SiC has been considered as a primary candidate material for a first wall component in future fusion reactor because it has been claimed that SiC has excellent high-temperature properties, good chemical stability and low activation. However, the behavior of tritium on SiC has not been discussed yet. In this study, tritium trapping capacity on the surface of SiC was experimentally obtained at the temperature range of 25-800 °C in consideration of tritium trapping to the experimental system. The capacity, which was independent of the water vapor pressure in the gas phase and the temperature, was determined as about 10⁶ Bq/cm². The isotope exchange reaction rate between tritiated water in a gas phase and hydrogen on the surface was quantified at the temperature of 25, 500 and 700 °C in consideration of the behavior of tritium trapping at change of experimental condition by the numerical curve fitting method applying the serial reactor model. The reaction rate was observed to be constant as 3.48 × 10⁻⁵ m/s. Additionally tritium release behavior from the surface of SiC in water vapor atmosphere was predicted and compared with that for graphite and stainless steel.     

Role of microstructure and heat treatments on the desorption kinetics of tritium from austenitic stainless steels

The liquid scintillation counting of solid samples (LSC-SS technique) was successfully used to study the role of microstructure and heat treatments on the behavior of residual tritium in several austenitic stainless steels (as-cast remelted tritiated waste, 316LN and 321 steels). The role of desorption annealing in the 100-600 °C range on the residual amount of tritium in tritiated waste was investigated. The residual tritium concentration computed from surface activity measurements is in good agreement with experimental values measured by liquid scintillation counting after full dissolution of the samples. The kinetics of tritium desorption recorded with the LSC-SS technique shows a significant desorption of residual tritium at room temperature, a strong barrier effect of thermal oxide films on the tritium desorption and a dependance of the tritium release on the steels microstructure. Annealing in the 300-600 °C range allows to desorb a large fraction of the residual tritium. However a significant trapping of tritium is evidenced. The influence of trapping phenomena on the concentration of residual tritium and on its dependance with the annealing temperature was investigated with different recrystallized and sensitized microstructures. Trapping is evidenced mainly below 150 °C and concerns a small fraction of the total amount of tritium introduced in austenitic steels. It presumably occurs preferentially on precipitates such as Ti(CN) or on intermetallic phases.