Influence on moisture and hydrocarbons on conversion rate of tritium in catalytic reactors of fusion-DEMO detritiation system

Thoughtful consideration of abnormal events such as fire is required to design and qualify a detritiation system (DS) of a nuclear fusion facility. Since conversion of tritium to tritiated vapor over catalyst is the key process of the DS, it is indispensable to evaluate the effect of excess moisture and hydrocarbons produced by combustion of cables on tritium conversion rate considering fire events. We conducted demonstration tests on tritium conversion under the following representative conditions: (I) leakage of tritium, (II) leakage of tritium plus moisture, and (III) leakage of tritium plus hydrocarbons. Detritiation behavior in the simulated room was assessed, and the amount of catalyst to fulfill the requirement on tritium conversion rate was evaluated. The dominant parameters for detritiation are the concentration of hydrogen in air and catalyst temperature. The tritium in the simulated room was decreased for condition (I) following ventilation theory. An initial reduction in conversion rate was measured for condition (II). To recover the reduction smoothly, it is suggested to optimize the power of preheater. An increase in catalyst temperature by heat of reaction of hydrocarbon combustion was evaluated for condition (III). The heat balance of catalytic reactor is a point to be carefully investigated to avoid runaway of catalyst temperature.     

空气除氚系统中氚氧化催化剂的研究进展

在大型氚设施中空气除氚系统必不可少,通过气-水转换除去气态氚是目前应用最广泛也是最有效的工艺,过程中氧化催化剂至关重要。总结了气态氚的催化氧化研究进展、催化剂的催化性能及影响催化性能的主要因素。贵金属Pt和Pd在室温下对氚的转化效率接近100%,因而被广泛用于氚的催化氧化。通过负载分散载体、添加催化助剂、使用规整结构催化剂、设计新型的催化反应器能够进一步提高催化剂性能。以蜂窝状催化剂为研究热点的规整结构催化剂以其高比表面积和低压力降而显示出良好的催化性能,将它用于氚的催化氧化,是该领域的一个研究方向。氢、氘、氚在氧化过程中的同位素效应会影响除氚效率,需进行深入研究。     

Preliminary results from a detritiation facility dedicated to soft housekeeping waste and tritium valorization

Nuclear waste management has to be taken into account for fusion machine using tritium as fuel. Soft housekeeping waste (e.g. gloves, tissues, protective clothes, etc.) is produced during the whole life as well as during the dismantling of the reactor and is contaminated by tritium under reduced (HT) and oxidized (HTO) forms.In collaboration with ENEA, a lab-scaled facility has been built at CEA Cadarache for soft housekeeping waste detritiation and tritium valorization. The previously milled waste is placed in a reactor to be heated up to a temperature lower than the housekeeping melting point. A carrier gas is then injected in the detritiation reactor to remove tritium, thanks to the combined effects of temperature and carrier gas (type and feed flow). The tritiated gas exhausted from the detritiation reactor is then sent through a catalytic Pd–Ag membrane reactor (CMR) where tritium is recovered via isotopic exchange reaction and permeation phenomenon.Based on previous studies that have allowed defining the most efficient operating conditions for the detritiation process, this work presents the results obtained by the coupling of the detritiation facility with the CMR. Due to safety considerations, restrictions on the nature of the carrier gas were applied, rejecting air as the carrier gas even though air was the best candidate for the detritiation part of the process. The performance of the whole system was estimated by means of a parametric study on the influence of flow rates in the CMR and transmembrane pressure.     

101重水研究堆含氚轻水脱氚方案研究

101重水研究堆(HWRR)是中国第一座核反应堆,现已停堆进入退役准备期。其乏燃料水池和废水贮存罐中存有一定量的含氚轻水,含氚浓度较高,需进行脱氚处理。本文针对HWRR含氚轻水的处理量和含氚浓度,分别评价了3种含氚水脱氚方案:两种联合电解催化交换(CECE1和CECE2)工艺和水蒸馏(WD)工艺。结果表明,与WD相比,CECE工艺的塔径和塔高更小,CECE电解槽的能耗也较WD工艺的蒸发器稍低;两种CECE工艺相比,顶部进天然水的CECE2工艺更适合处理HWRR的含氚轻水。     

Design Study of Feasible Water Detritiation Systems for Fusion Reactor of ITER Scale

Two types of water detritiation systems have been designed for fusion reactors of ITER scale. One of the systems is a combination of WD (Water Distillation) and VPCE(vapor phase catalytic exchange) columns. The other is a combination of a WD column and a CECE(combined electrolysis catalytic exchange) column. Three water distillation columns are needed for the former system. The total height of the three columns is 106 m. The height of the water distillation and CECE columns for the latter system are 20 and 24m, respectively. These large water distillation columns result in the larger tritium inventory of the former system than for the latter system. However, there have been the results for the operation of the actual scale of the water distillation and VPCE columns. No demonstration test has been carried out for the CECE column. From these reasons, the WD+VPCE system should be the first candidate for the fusion reactor. The WD+CECE system is superior to the WD+VPCE system for the flexibility in design as well as the tritium inventory. It is desired to demonstrate the CECE column to develop the water detritiation system best suited to the fusion reactors.     

The Second International Symposium on Advanced Nuclear Energy Research

The diffusion of tritiated water vapor into concrete walls of a fusion reactor building is studied to evaluate the capability of the tritium containment of concrete materials.

First, depth profiles of tritiated water in concrete are calculated to evaluate the capacity of the tritium containment by the sound concrete without cracks, and a 0.5-m-thick concrete wall is sufficient to prevent tritiated water releases to the environment in a normal operation of a fusion reactor over 50 yr. Second, simulations of the cleanup of tritiated water in the concrete reactor hall atmosphere taking into account the soaking are performed. Concrete porosity should be decreased to shorten the cleanup time of the reactor hall atmosphere. Surface coatings on the concrete, which apparently decrease the surface porosity, are effective measures to prevent the diffusion of tritiated water vapor into concrete during accidental releases.     

联合电解催化交换系统的动态模型及理论计算

为探求联合电解催化交换系统各单元中氚浓度空间分布和动态变化的内在规律，建立了D／T体系的气-液两元模型。根据不同的催化剂传质性能，计算了为达到特定脱氚率和电解池浓缩倍数所要求的交换床总高度和进液位置。理论计算得到的氚在交换床上的空间分布趋势与文献报道的中试结果一致，电解池中的氚浓度随时间呈线性增长。     

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.     

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.     

Processing highly tritiated water desorbed from molecular sieve bed using PERMCAT

Tritium handling facilities use molecular sieve beds (MSB) to collect and recover tritiated water. After reaching the capacity limit of the MSB, the water is desorbed and decontaminated in a water detritiation system (WDS). In the case of highly tritiated water (HTW) absorbed into a MSB, an inherent safe option for processing is necessary due to the HTW specific properties. Ideally, HTW should be processed immediately in a continuous mode. With this in consideration, the water desorption process from a zeolite bed was developed and optimized in a dedicated non active facility. The results of this experiments were applied into the regeneration of a MSB previously loaded with HTW containing an activity of 1.9 × 10¹⁴ Bq kg^?1. The water was desorbed, by step increasing the temperature bed and fed by helium carrier gas into the PERMCAT for detritiation and tritium recovery. The processed water was collected in a dry MSB downstream of the PERMCAT. These initial studies successfully demonstrate the viability of the process. The obtained results of the preliminary study and the subsequent tests with tritium, will provide useful information for the design of tritium processes relying on MSB, such as the water processing foreseen for the test blanket modules in ITER.     

Development of a technical scale PERMCAT reactor for processing of highly tritiated water

In view of future fusion rectors fueled by deuterium–tritium mixtures, highly tritiated water (HTW) of up to 5.2·10¹⁶ Bq kg^?1 will be produced, during routine operation and scenarios as an accidental release of tritium into a glove box. Also in the solid breeder blanket concept, a non-negligible fraction of the tritium produced will be in the tritiated water fraction. To decontaminate HTW the PERMCAT using isotope swamping in a Pd/Ag membrane reactor has been identified as a robust and reliable solution. In order to investigate the decontamination of HTW at flow rates relevant for future fusion power plants, a technical scale, fully tritium compatible PERMCAT consisting of a bundle of finger-type membranes inserted in a single catalyst bed was developed. Nevertheless, it represents only one part of a PERMCAT cascade necessary to achieve the required performance to process HTW on technical scale. By improving the existing PERMCAT geometry using experimental data obtained from isotopic exchange between D₂O and H₂, the performance of the existing PERMCAT reactor was optimised. Based on the optimised geometry a new fully tritium compatible technical scale PERMCAT cascade comprising of two PERMCAT reactors in series was designed, manufactured and commissioned as presented in this paper.     

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.     

带自动控制功能的除氚系统

针对氚工艺尾气处理的需求和源项实际情况,根据化工原理和氚特性设计含自动控制功能除氚系统的主要部件、自动控制功能和初步性能测试,得出催化反应器、氚水吸附床等部件的结构尺寸,催化剂、干燥吸附剂的装填量等参数及控制软件界面;通过除氘和除氚实验初步测试了除氚系统的处理性能。结果表明,在循环处理模式下,1m3密闭容器中氘体积比6.0×10–4–2.8×10–2范围内时,35min内氘气浓度降低两个量级;5次对30L密闭容器内不同浓度的含氚气体处理,60min内对氚的去除效率均达到95%以上。     

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.     

Preliminary results from a detritiation facility dedicated to soft housekeeping waste

Nuclear waste management has to be taken into account for fusion machine in tritium experimentations. Soft housekeeping waste is produced during both operating and dismantling phases and is contaminated by tritium under reduced (HT) and oxidized (HTO) forms. At CEA Cadarache, a lab-scaled facility has been built for soft housekeeping detritiation. The tritiated gas exhausted from the process described above is foreseen to be treated by a tubular Pd–Ag membrane reactor, for gaseous tritium recovery. Since this membrane reactor uses hydrogen as swamping gas the compatibility toward explosive hazard has to be taken into account. Then, this work presents a double objective. A first study is presented in order to identify the best conditions for the declassification of soft housekeeping waste, without tritium recovery. Experiments carried out at 120 °C are not efficient enough and do not allow one to choose the most efficient carrier gas. Some other tests are being currently performed at higher temperatures (150 °C). Moreover, due to safety issues, the use of air has to be avoided during membrane reactor implementation phase. Preliminary results obtained with hydrogen hazard-free carrier gases are also presented.     

Tritium Safety-Related Studies at TPL of JAERI

Activities regarding tritium safety technology in the Tritium Process Laboratory (TPL) at Tokai Establishment of Japan Atomic Energy Research Institute are reviewed. Research and development of a new tritium removal system is being carried out by using a gas separation membrane which enable to make the ITER atmosphere detritiation system more compact and cost-effective. Techniques of gas flowing calorimetry and laser Raman spectroscopy are applied to develop new tritium accountancy methods. Studies of tritium-material interaction, such as plasma material interactions, radiochemical reaction of tritium in gas phase, radiolysis of tritiated water, and waste processing are being carried out under ITER/EDA and U.S.-Japan collaboration. Tritium release experiments for research of tritium behavior in confinements and environment and demonstration of safety related components are planned.     

Overview of the tritium system of Ignitor

Among the recent design activities of the Ignitor program, the analysis of the tritium system has been carried out with the aim to describe the main equipments and the operations needed for supplying the deuterium–tritium mixtures and recovering the plasma exhaust.

In fact, the tritium system of Ignitor provides for injecting deuterium–tritium mixtures into the vacuum chamber in order to sustain the fusion reaction: furthermore, it generally manages and controls the tritium and the tritiated materials of the machine fuel cycle. Main functions consist of tritium storage and delivery, tritium injection, tritium recovery from plasma exhaust, treatment of the tritiated wastes, detritiation of the contaminated atmospheres, tritium analysis and accountability.

In this work an analysis of the designed tritium system of Ignitor is summarized.     

99Tcm标记趋化肽类似物的制备

利用固相法合成趋化肽类似物fMLFK,并采用双功能螯合剂(HYNIC)技术合成了^99Tc^m标记的趋化肽类似物^99Tc^m-HYNIC-fMLFK,测定了标记物的标记率和放化纯度,考察了配合物的稳定性。结果表明,合成fMLFK的产率大于80％,纯度大于97％;质谱法测得的相对分子质量与理论值相符;^99Tc^m标记后,配合物的放化纯度大于95％,在24h内放化纯度无明显降低。     

φ6mm Pt-PTFE疏水催化剂的研制

在研制粒径为6mm的多孔PTFE疏水担体基础上,采用浸渍法研制了可在工程上应用的Pt PTFE疏水催化剂。室温下,在并流催化床上考察了该疏水催化剂的催化活性、疏水性能和催化剂上活性粒子的稳定性。结果表明:当氚浓度为153Bq/mL,氢气线速度为5.31cm/s和15.93cm/s时,水中氚的催化转化率分别为73.7%和69.6%。在用普通水浸泡并淋洗145d后,该疏水催化剂的催化活性和活性粒子(Pt)的含量无明显变化。     

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.