K. Kobayashi

Improvement of tritium accountancy technology for ITER fuel cycle safety enhancement

In order to improve the safe handling and control of tritium for the ITER fuel cycle, effective in situ tritium accounting methods have been developed at the Tritium Process Laboratory in the Japan Atomic Energy Research Institute under one of the ITER-EDA R&D tasks. The remote and multilocation analysis of process gases by an application of laser Raman spectroscopy developed and tested could provide a measurement of hydrogen isotope gases with a detection limit of 0.3 kPa analytical periods of 120 s. An in situ tritium inventory measurement by application of a `self-assaying storage bed with 25 g tritium capacity could provide a measurement with the required detection limit of less than 1% and a design proof of a bed with 100 g tritium capacity.

Tritium Behavior on the Water-Metal Boundary for the Permeation into Cooling Water Through Metal Piping

Abstract In order to accumulate data on tritium transferred to cooling water of a fusion reactor, a series of experiments of tritium permeation into water jacket pressurized to 0.8MPa by He gas was performed through pure iron piping, which contained about 1 kPa of pure tritium gas at 423 K. Chemical forms of tritium permeated into water were monitored periodically under continuous purging water jacket by He. Observation of metal surface was also carried out periodically by SEM and XRD analysis. The actual tritium permeation rate was about 1/5 level of the calculated value. Even if surface oxide layer (magnetite, porous & fine layers) grew in the water boundary, tritium permeation rate to water was not changed drastically. On the other hand, hydrogen gas (HT) fraction of tritium permeated in water jacket decreased drastically with oxide layer growth. Furthermore, permeated species and amounts were not affected clearly by the dissolved hydrogen in water by purging 1% H2 in He.

Concentration profiles of tritium penetrated into concrete

Abstract Concentration profiles of tritium in cement paste, mortar and concrete were measured after exposure to tritiated water vapor for a given time. Tritium penetrated a distance of about 5 cm from the exposed surface during an exposure of 6 months. The model of tritium behavior in concrete materials reported by the present authors was developed in this study with the consideration of the effects of sand and aggregate on both the diffusion coefficient of tritiated water vapor and the isotope exchange capacity. Predictive calculations based on the tritium transport model were also carried out in some situations of tritium leakage. The results of the calculations show that a large amount of tritium will be trapped in the concrete walls, and the trapped tritium will be gradually released back to the tritium handling room over the time of months to years even after the decontamination of the room is completed.

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.

Observation of Tritium Distribution in Iron Oxide with Tritium Micro Autoradiography

Abstract To clarify the tritium permeation behavior, tritium distribution in iron oxidized in high temperature water was observed with tritium micro autoradiography. It was found that tritium was distributed homogeneously in the iron metal. However the oxide surface (magnetite) was found to contain a very low concentration of tritium. The inner layer of oxide could strongly effect the tritium permeation. From a comparison with the permeation experiment that had been reported in Ref. 1, it was suggested that tritium would mainly diffuse other path except the oxide lattice. According to the chemical form of tritium, which was released from iron surface into water, two assumptions were suggested. One is based on the different combination of tritium on the water-surface interface. The other is based on the oxidation mechanism.

Tritium Behavior Study for Detritiation of Atmosphere in a Room

ABSTRACT To construct the ITER with high safety and acceptability, it is very important to grasp the removal behavior of tritium happened to leak in the room, the final confinement barrier. In order to obtain data on tritium removal behavior from atmosphere in a room under the various conditions (humidity, ventilation flow rate), intentional tritium release experiments have been carried out with the Caisson Assembly for Tritium Safety Study (CATS) which consists of 12 m3 gas-tight box (Caisson) for the study of tritium behavior in large space. Effect of adding water vapor has also investigated for effective removal. When the tritiated water existed in the released tritium, residual contamination on the wall of the Caisson was detected under the various ventilation flow rate and it was found that it depended on the initial humidity in the Caisson. On the other hand, when the water vapor was added into the Caisson after found the residual contamination, the residual contamination was removed quickly on the wall of the Caisson. The adding water vapor into the Caisson, it was effective for the tritium removal. Analytical work have also progressed and analyzed tritium removal behavior became to be in good agreement with the experimental results by considering the adsorption and desorption reaction rate of tritiated water on the wall.

Development of a tritium fuel processing system using an electrolytic reactor for ITER

A system composed of a palladium diffuser and an electrolytic reactor has been proposed and then developed for the fuel cleanup system of ITER. The performance of the system has been studied in detail in a stand-alone test. A fuel simulation loop of ITER was constructed by connecting the fuel cleanup and hydrogen isotope separation systems developed; and the function of each system in the loop has been demonstrated. For tritium recovery from the exhaust gas during helium glow discharge cleaning of the vacuum chamber of ITER, a cryogenic molecular sieve bed system has been proposed and demonstrated.

Studies on the behaviour of tritium in components and structure materials of tritium confinement and detritiation systems of ITER

Confinement and the removal of tritium are key subjects for the safety of ITER. The ITER buildings are confinement barriers of tritium. In a hot cell, tritium is often released as vapour and is in contact with the inner walls. The inner walls of the ITER tritium plant building will also be exposed to tritium in an accident. The tritium released in the buildings is removed by the atmosphere detritiation systems (ADS), where the tritium is oxidized by catalysts and is removed as water. A special gas of SF6 is used in ITER and is expected to be released in an accident such as a fire. Although the SF6 gas has potential as a catalyst poison, the performance of ADS with the existence of SF6 has not been confirmed as yet. Tritiated water is produced in the regeneration process of ADS and is subsequently processed by the ITER water detritiation system (WDS). One of the key components of the WDS is an electrolysis cell. To overcome the issues in a global tritium confinement, a series of experimental studies have been carried out as an ITER RD (2) the effect of SF6 on the performance of ADS and (3) tritium durability of the electrolysis cell of the ITER-WDS. (1) The tritiated water vapour penetrated up to 50?mm into the concrete from the surface in six months exposure. The penetration rate of tritium in the concrete was thus appreciably first, the isotope exchange capacity of the cement paste plays an important role in tritium trapping and penetration into concrete materials when concrete is exposed to tritiated water vapour. It is required to evaluate the effect of coating on the penetration rate quantitatively from the actual tritium tests. (2) SF6 gas decreased the detritiation factor of ADS. Since the effect of SF6 depends closely on its concentration, the amount of SF6 released into the tritium handling area in an accident should be reduced by some ideas of arrangement of components in the buildings. (3) It was expected that the electrolysis cell of the ITER-WDS could endure 3 years operation under the ITER design conditions. Measuring the concentration of the fluorine ions could be a promising technique for monitoring the damage to the electrolysis cell.

Tritium transfer in porous concrete materials coated with hydrophobic paints

Abstract An experimental study on tritium transfer in porous concrete materials for the tertiary tritium safety containment is performed to investigate; (i) how fast tritium is transferred through porous concrete walls coated with or without a hydrophobic paint, and (ii) how well the hydrophobic paint coating works as a film protecting against tritium migrating through concrete. The experiment is comparatively carried out using two types of cement-paste and mortar disks with or without two kinds of paints. The results obtained here are summarized as follows: (1) Tritium transfer can be correlated in terms of the effective tritium diffusivity of DT=1.2x10-11 m2/s in porous cement. (2) Adsorbed or condensed liquid HTO itself is transferred only through pores in cement, and no tritium transfer path is present in non-porous sand. (3) Rates of tritium sorption and dissolution in cement and mortar coated with an epoxy-resin paint is correlated in terms of the diffusivity through the paint film of DT=1.0x10-16 m2/s. (4) The epoxy paint works more effectively as an anti-tritium diffusion coating than the acrylic-silicon resin paint. (5) The hydrophobic property of the silicon resin paint is deteriorated with elongating the contact time with H2O.

Recent activities on tritium technologies for iter and fusion reactors at jaea

Abstract The design studies of Atmosphere Detirtiation System (ADS) have been carried out in Japan Atomic Energy Agency (JAEA) as a contribution of Japan to ITER. The performance of ADS has also been investigated under accidental conditions such as fire and co-existing of a poison gas for catalyst like SF6. There is no degradation of Detritiation Factor (DF) under co-existing of CO or CO2 up to 20% as a simulated fire condition. However, only 0.1% of SF6 degrades the DF from more than 1000 to 50, following reduction of water by SF4 etc. (decomposition products of SF6) at 773K of catalyst bed. For the tritium processing technologies, our efforts have been focused on the R & D of the tritium recovery system of breeding blanket. In case of ITER Test Blanket Module, a cryogenic molecular sieve bed system was designed and demonstrated. Furthermore, electro-chemical pumping system using a proton conductor is also investigated to design more effective system. The durability of electrolysis cell for Water Detritiation System (WDS) has been investigated and it is expected that the cell can endure more than 3 years’ operation under the ITER WDS design condition. A series of fundamental studies on tritium safety technologies has been carried out as another major activity of JAEA for ITER and future fusion demo reactors. Tritium behavior in various confinement materials, tritium monitoring & accountancy, and detritiation were studied under collaboration programs with universities, using Caisson Assembly for Tritium Safety study.