Tritium Recovery Research Articles

Overview of the TBM R&D activities in Japan

In Japan, development of Water Cooled Ceramic Breeder (WCCB) blanket is being performed as the primary candidate of Test Blanket Module (TBM) toward DEMO. Also development of high temperature Li–Pb blanket, Li/V blanket and Flibe blanket are being performed toward the advanced DEMO blanket. Regarding the WCCB TBM development, real scale First Wall (FW) was fabricated using Reduced Activation Ferritic Martensitic Steel (RAFMS), F82H and successfully tested by heat flux equivalent to ITER TBM condition, 0.5 MW/m 2, 80 cycles with the same coolant condition as DEMO. Also, real scale Side Wall (SW) and real scale breeder pebble bed structure are successfully fabricated. Furthermore, asemmbling test of the real scale FW plate mockup and SW plate mockup was performed to clarify the welding condition to form TBM box structure. All key technologies of blanket module fabrication has been addressed and almost achieved toward TBM. Regarding the development of liquid breeder blankets, all key technologies, such as, material compatibility and mass transfer, tritium recovery performance, insulation coating development, were covered, by using elementary experiments to convection loops of Li–Pb, Li and Flibe. High temperature heat exchange test up to 900 °C has been initiated by using Li–Pb test loop.

Microstructure control and deuterium permeability of erbium oxide coating on ferritic/martensitic steels by metal-organic decomposition

The development of a tritium permeation barrier is essential for building a tritium recovery system in fusion power plants. Toward realizing this application, the fabrication of erbium oxide coatings by metal-organic decomposition has been carried out on reduced activation ferritic/martensitic steels. The coated samples exhibit different surface structures after they are heat-treated under different conditions. A sample that is heat-treated in high-purity argon produces an oxide layer between the erbium oxide coating and the substrate. It is believed that the presence of this oxide layer causes defects in the coating and degradation of the samples during deuterium permeation measurements. The sample heat-treated in high-purity hydrogen with moisture exhibits a thinner oxide layer and improved stability during measurements. A 0.3-μm-thick coating yields a permeation reduction factor of 500–700 at 773–973 K, which is comparable to the coating deposited by the physical vapor deposition technique.

Mechanical design and first experimental results of an upgraded technical PERMCAT reactor for tritium recovery in the fuel cycle of a fusion machine

The PERMCAT process developed for the final clean-up stage of the Tokamak Exhaust Processing systems of the ITER tritium plant combines a catalytic reactor and a Pd/Ag permeator in a single component. A first generation technical PERMCAT has been successfully operated as part of the CAPER experiment at the Tritium Laboratory Karlsruhe for several years. Various alternative PERMCAT mechanical designs were proposed and studied on small-scale prototypes. An upgraded technical PERMCAT reactor was designed, manufactured and commissioned with deuterium. A parallel arrangement of finger-type membranes inserted in a single catalyst bed design was chosen to simplify the geometry and the manufacturing while improving the robustness of the reactor. The component has been designed and manufactured to be fully tritium compatible and also fully compatible with both process and electrical connections of the previous PERMCAT to be replaced. The new PERMCAT mechanical design is more compact and easy to manufacture. This PERMCAT reactor was submitted to functional tests and experiments based on isotopic exchanges between H 2O and D 2 to measure the processing performances. The first experimental results show decontamination factors versus flow rates better than all previously measured.

Isotopic exchange between hydrogen and deuterium in the process of permeating through Li 0.17Pb 0.83

The permeation process of hydrogen isotopes through Li 0.17Pb 0.83 has been investigated experimentally. We obtained the overall D permeation rates that take into consideration of the effects of the H–D isotopic exchange reaction on the surface and diffusion in the Li–Pb layer. It was proved that the rate-determining step was not the surface reaction but the diffusion in the Li–Pb bulk. We proposed a system of gas–liquid counter-current extraction using a packed column for continuous tritium recovery. The concentration profiles through the column and tritium extraction rates are calculated. It is estimated that the column height of 5 m is sufficient to achieve the tritium recovery ratio of 99.999%. This counter-current extraction method is easy to operate and efficient to recover tritium from Li–Pb.

Counter-current isotope swamping in a membrane reactor: The PERMCAT process and its applications in fusion technology

A PERMCAT reactor is a catalytic membrane reactor that combines a Pd/Ag membrane and a catalyst bed. It has been developed to ensure very high tritium recovery from the unspent fuel of fusion machines using deuterium tritium mixtures. The PERMCAT process takes advantage of simultaneously unlocking chemically bound tritium via heterogeneously catalysed isotope exchange reactions and removing tritium via its selective permeation through the membrane. The PERMCAT reactor operated in the counter-current isotope swamping mode allows a very low tritium activity at the outlet of the component to be maintained. Two main issues have been solved to achieve efficient and reliable PERMCAT operation. Firstly, the mechanical design has to cope with the elongation and deformation of the membrane resulting from thermal expansion and lattice parameter increase under operation with hydrogen. Secondly, the catalyst material has to be chosen in order to promote isotope exchange reactions while minimising the numerous side reactions that occur especially when the mixture contains carbon oxides. This paper presents a general overview of the R&D performed at the Tritium Laboratory Karlsruhe for PERMCAT technology. Technical solutions to solve both issues together with relevant experimental results including processing tests with tritium are discussed.

Study on Tritium Balance in a D-T Fusion Reactor

The tritium balance in a D-T fusion reactor is discussed in this paper comparing the amount of tritium consumed in the fueling cycle including the plasma vessel with the amount of tritium generated in the blanket system, using information reported so far. This comparison shows that the overall burning efficiency of tritium in the plasma vessel, the tritium loss ratio represented by tritium trapping in the redeposition layer of the plasma-facing material, and the recovery efficiency in the tritium breeding system play important roles in the tritium balance and that it may not be easy to maintain good tritium economy of a D-T fusion reactor if the proper combination of burning efficiency, tritium loss ratio, and tritium recovery efficiency is not obtained. The allowable limits for the overall tritium burning efficiency, for the tritium loss ratio in the fueling cycle, and for the recovery efficiency to secure the self-sustainable tritium system are also discussed in this paper.

Study on the fuel balance of a DT reactor

The tritium balance in a DT fusion reactor is discussed in this paper. It is pointed out in this discussion that tritium balance in a fusion reactor is largely influenced by such uncertainties as plasma vessel effect, module effect and recovery efficiency of bred tritium required to estimate the overall tritium recovery efficiency and such uncertainties as overall burning efficiency of tritium in the plasma vessel, tritium trapping ratio at the first wall, loss due to β-decay of tritium inventory and designated reactor base tritium breeding ratio required to decide ( β) overall to decide tritium balance.

Design of Gaseous Tritium Recovery System Applying Commercially Available Membrane-Type Dehumidifier

In order to realize the planned deuterium plasma experiments using the large helical device (LHD), the National Institute for Fusion Science (NIFS) is planning to install a system for tritium recovery from exhaust gas. While adopting typical tritium recovery systems, NIFS has also made plans for the development of a compact reduced-waste recovery system by applying a membrane-type dehumidifier. The applicability of a commercially available membrane dehumidifier has been evaluated experimentally, with the results indicating that such a membrane is feasible for practical application. Based on performance test results, the optimum specifications of the membrane dehumidifier are evaluated quantitatively and an actual tritium recovery system for the LHD has been designed.

Optimization for the tritium isotope separation factor and permeation by selecting temperature and thickness of the diffusion Pd–Ag alloy cathode

Large quantities of tritiated water will be produced in the controlled fusion reactors for power generation. To eliminate the concentrated tritiated water and for recycling tritium, an industrial electrolyzer was developed. The aim of this paper is to give the design of this electrolyzer and the results for optimization in diffusion and in isotopic exchange by selecting thickness of a thimble-shaped Pd–25%Ag cathode working at high temperature and current. In this process, the tritium recovery system is based on the principle of the tritium diffusion Pd–Ag cathode which produces very pure hydrogen isotopes from enriched tritiated water.

Research and development of the tritium recovery system for the blanket of the fusion reactor in JAEA

A water-cooling solid breeder blanket is a prime candidate for the blanket of the fusion reactor in Japan. In this case, the blanket tritium recovery system will be composed of three processes: tritium recovery from helium sweep gas as hydrogen, that as water vapour and tritium recovery from coolant water. The authors have proposed a set of advanced systems. For tritium recovery as hydrogen, an electrochemical hydrogen pump with a ceramic proton conductor has been proposed. The correlation between the proton concentration in the ceramic and the hydrogen gas pressure has been investigated to describe the pumping performance specifically. A ceramic electrolysis cell has been proposed to process the tritiated water vapour. The authors have developed a new electrode containing cerium oxide, and it has shown fairly good electrolysis efficiency. For tritium recovery from coolant water, reduction in the processing water by tritium concentration is necessary. The authors have proposed to apply the fixed-bed adsorption process of synthetic zeolite, and have developed new zeolite. It showed unique characteristics for water adsorption and desorption. The authors have determined the potential of these systems for the blanket of the fusion DEMO reactor.

R&D of a Li2TiO3 pebble bed for a test blanket module in JAEA

At JAEA, a test blanket module (TBM) with a water-cooled solid breeder is being developed. This paper presents recent achievements of research activities for the TBM, particularly addressing the pebble bed of the tritium breeder materials and tritium behaviour. For the breeder material, the chemical stability of Li2TiO3 was improved using Li2O additives. To analyse the pebble bed behaviour, thermomechanical properties of the Li2TiO3 pebble bed were assessed experimentally. To verify the pebble bed's nuclear properties, the activation foil method was proposed and a preliminary experiment was conducted. To reduce the tritium permeation, the chemical densified coating method was developed and the coating was attached to F82H steel. For tritium behaviour, the tritium recovery system was modified in consideration of the design change of the TBM.

Optimisation of operating parameters for plasma facing components detritiation

Within the framework of management of waste produced by fusion facility, the presence of high tritium content in large pieces is very specific. This question is very sensitive in the case of maintenance of plasma facing components (PFC). In order to reach several goals (tritium recovery, limitation of tritium content in the waste), several treatment methods have been analysed and thermal detritiation is selected as a reference process. In the case of ITER facility a tritium recovery station based on this technique is foreseen. In order to better understand which parameters have important influence on the efficiency of the treatment (detritiation factor), a recall of the process principle and of the main data concerning tritium behaviour in different materials is done. For example, low tritium diffusion coefficient in Be which is the material of the first layer on some components can limit drastically the process performance. In this paper a recall of the different process parameters is presented. Some important parameters are selected in order to lead a sensitivity study on adequate variation domains. It allows studying several scenarios of treatment and analysing the consequences of them. The main parameters of this study are the level of temperature, the duration of treatment, the renewal rate of the furnace atmosphere, the presence of Be or not. The main response to this sensitivity analysis is the detritiation factor, however some other consequences such as tritiated water flow rate and also tritium concentration in water are evaluated. Then the consequences of change of operating parameters are compared in order to identify what could be the most appropriate operating domain.

In-situ observation of hydrogen hot trapping from molten lithium with yttrium

Hydrogen permeation window is candidate not only for tritium recovery system from molten lithium, but also for hydrogen monitor for molten lithium because hydrogen permeation rate is dominated by hydrogen concentration. In this study, in-situ observation of deuterium hot trapping by yttrium in molten lithium was performed, measuring HD and D 2 permeation through an iron wall at 773, 823 and 873 K. Deuterium permeation showed good agreement with deuterium concentration in lithium and iron permeation wall was found to be reliable for deuterium monitoring for liquid lithium around a hundred wppm. As a result of in-situ observation, gettering rate for deuterium was 6 × 10 −5 mol/m 2/sec at 773 K, 5 × 10 −4 mol/m 2/sec at 823 K and 4 × 10 −4 mol/m 2/sec at 873 K. Oxidation or nitridation at the surface of yttrium may suppress gettering rate at 873 K.

Tritium production, recovery and application in Korea

Four CANDU reactors have been operating at the site of Wolsong Nuclear Power Generation in Korea. The Wolsong tritium removal facility was constructed to reduce the tritium levels in heavy water systems. This facility was designed to process 100 kg/h of tritiated heavy water feed and to produce 99% pure T 2. This recovered tritium will be made available for commercial applications. The initial phases on the tritium applications are made to establish the infrastructure and the tritium controls.

Recent results of R&D activities on tritium technologies for ITER and fusion reactors at TPL of JAEA

At Tritium Process Laboratory (TPL) of Japan Atomic Energy Agency (JAEA), tritium technologies for a fusion reactor have been carried out up to date. The design studies of Air Detrtiation and Vent Detritiation System (ADS/VDS) of ITER have been carried out in JAEA as a contribution of Japan to ITER. For the tritium processing technologies, our efforts have been focused on the R&D of the tritium recovery system of ITER test blanket system. A ceramic proton conductor has been studied as an advanced blanket system. A series of fundamental studies on tritium safety technologies for ITER and for fusion DEMO plants has also been carried out. The main R&D activities in this field are the tritium behavior in a confinement and its barrier materials, monitoring, accountancy, detritiation and decontamination, etc. Especially, for the fundamental studies for a DEMO plants, a part of the studies will be carries out a new facility at Rokkasho in Aomori in the Broader Approach (BA) program in coming 10 years. For this purpose, a design study of the facility at Rokkasho has first been started.

Recovery of Tritium Dissolved in Sodium at the Steam Generator of Fast Breeder Reactor

The tritium recovery technique in steam generators for fast breeder reactors using the double pipe concept was proposed. The experimental system for developing an effective tritium recovery technique was developed and tritium recovery experiments using Ar gas or Ar gas with 10-10000 ppm oxygen gas were performed using D2 gas instead of tritium gas. It was found that deuterium permeation through two membranes decreased by installing the double pipe concept with Ar gas. By introducing Ar gas with 10000 ppm oxygen gas, the concentration of deuterium permeation through two membranes decreased by more than 1/200, compared with the one pipe concept, indicating that most of the deuterium was scavenged by Ar gas or reacted with oxygen to form a hydroxide. However, most of the hydroxide was trapped at the surface of the membranes because of the short duration of the experiment.

Hydrogen Isotope Recovery Using a Cathode Water Vapor Feed PEM Electrolyzer

A critical component of tritium glovebox operations is the recovery of high value tritium from the water vapor in the glove box atmosphere. One proposed method to improve existing tritium recovery systems is to replace the disposable hot magnesium beds used to separate the hydrogen and oxygen in water with continuous use Proton Exchange Membrane Electrolyzers (PEMEs). This study examines radiation exposure to the membrane of a PEME and examines the sizing difference that would be needed if the electrolyzer were operated with a cathode water vapor feed instead of an anode liquid water feed.

Influence of Oxygen on Permeation of Hydrogen Isotopes Through Group 5 Metals

Group 5 metals (V, Nb and Ta) are potential candidates of tube material in vacuum permeator for tritium recovery from Pb-17Li liquid blanket system. From this viewpoint, the influence of oxygen on the surface reaction rates of hydrogen on V and Ta were examined in an ultra-high vacuum apparatuses at elevated temperatures, and the results were compared with the data on Nb acquired in a previous study. The surface reaction rates of hydrogen on V and Ta, and consequently permeation rates, decreased with increasing oxygen concentration in the bulk as previously observed for Nb. These observations were ascribed to the increase in surface oxygen coverage with increasing bulk oxygen concentration. The weakest influence of oxygen on hydrogen permeation rate was observed for V. The expected permeation rate through V under typical blanket conditions, however, was not necessarily high due to high oxygen solubility in V. The evaluation indicated that the highest permeation rate should be obtained with Nb under typical blanket conditions.

Tritium recovery system for Li–Pb loop of inertial fusion reactor

The best material for a wet wall and blanket of an inertial fusion reactor is selected among Li, eutectic alloys of Li–Pb, Li–Sn and a 2LiF + BeF 2 molten salt mixture called Flibe, judging from their chemical, nuclear and heat-transfer properties. Li 0.17Pb 0.83 is found to be the most promising one because of low Li vapor pressure, moderate melting temperature, good heat-transfer properties under the condition of a KOYO-fast circulation loop operated between 300 and 500 °C. A counter-current extraction tower packed with metallic rashig rings is proposed to extract tritium generated and dissolved in the Li–Pb eutectic alloy. Mass-transfer parameters when He and Li–Pb flow counter-currently through the tower packed with the rings are determined to satisfy the two necessary conditions of a self-sufficient tritium generation rate of 1.8 MCi/day and a target tritium leak rate of 10 Ci/day. It is found that the height of a tower to achieve the 99.999% recovery is comparatively low because of the promising property of a large equilibrium pressure of tritium. In order to mitigate the disadvantage of its high density, which needs a large pumping power, a porous packing material that keeps good contact between He and Li–Pb should be developed in the future. It is found experimentally that D 2 addition in He purge gas is effective to achieve a faster rate of tritium recovery from the Li–Pb flow. The rate-determining step of tritium permeation through a steam generator is determined as a function of a Li–Pb flow rate in a stainless-steel heat-transfer tube.

Improved characteristics of hydrophobic polytetrafluoroethylene–platinum catalysts for tritium recovery from tritiated water

This paper presents recent results concerning a new preparation method and characteristics of the hydrophobic catalysts used in hydrogen isotopes separation. The objectives of the paper are: (1) to assess the current status and find a new procedure for the preparation of a new improved hydrophobic catalyst; (2) to improve the characteristics and performance of platinum hydrophobic catalysts. As a result of review of references it was concluded that platinum is one of the most active and efficient catalytic metal and polytetrafluoroethylene is a very good wet-proofing agent in the hydrogen isotopes separation processes. A category of newly improved hydrophobic Pt-catalysts has been proposed, prepared and characterized and are now under testing. The novelty consists in adding of metallic oxides (titanium dioxide, zirconium dioxide) as a new binding and wet-proofing agents. The added metallic oxides play a catalytic role, too. The physico-structural parameters of the improved catalyst have been determined and are discussed. Preliminary tests concerning the catalytic activity and stability in tritium separation process have been carried-out. The new proposal is a promising idea to improve the performance of conventional hydrophobic Pt-catalysts.