K. Isobe

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.

Hydrogen isotope exchange in tungsten irradiated sequentially with low-energy deuterium and protium ions

Hydrogen isotope exchange in tungsten was investigated at various temperatures both after sequential exposure to low-energy deuterium (D) and protium (H) plasmas and after sequential irradiation with low-energy D and H ions. The methods used were thermal desorption spectroscopy, and the D(3He,p)4He nuclear reaction at 3He energies varied from 0.69 to 4.0 MeV, allowing the determination of the D concentration at depths of up to 6 μm. It was found that a major portion of the deuterium initially accumulated in the D-implanted W is released on subsequent exposure to H plasma or irradiation with H ions. Depth profiling of D without and with subsequent H implantation shows strong replacement close to the surface near room temperatures, but extending to all analyzable depths at elevated temperatures.

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.

Effect of Tritium and Dissolved Oxygen on Anodic Polarization of SUS304 Stainless Steel in Sulfuric Acid Solution

Since exotic corrosion of stainless steels in tritiated water can be expected, the anodic polarization of a SUS304 stainless steel sample in approximately 5 wt% sulfuric acid solution was performed at various concentrations of tritium and dissolved oxygen (hereafter DO) in the electrolyte. The inhibitory effect of tritium on the passivation could be observed with DO even at a tritium concentration in the electrolyte of as low as 2.2 kBq cm−3. This effect became more pronounced as the tritium concentration increased. It was suggested that the inhibitory reaction depending on tritium concentration would compete with the self-passivation depending on the DO concentration (hereafter [DO]), since it was found that there is a threshold [DO] for self-passivation at each tritium concentration.

Demonstration of fuel cleanup system consisting of electrolytic reactor and tubular reservoir tank for fusion reactors

ABSTRACT A Fuel cleanup system (FCU) that recovers fusion fuel (tritium and deuterium) from plasma exhaust mixture gas has been developed and demonstrated at the Tritium Process Laboratory (TPL) of Japan Atomic Energy Research Institute (JAERI). We have proposed a new closed loop FCU system built up by connecting the tubular reservoir tank, the electrolytic reactor and a palladium diffuser. In the electrolytic reactor, methane and water are converted at the same time by electrochemical reaction in gas phase oxidation and reduction to liberate hydrogen isotope as a form of elemental hydrogen. The long tubular reservoir tank that is designed to store and transfer the products gas in plug flow prevents from mixing with reactants for the successive repeat processing. With this tank, high overall decontamination factor of system can be obtained by small number of circulation. As the demonstration test, mixture gas consist of hydrogen isotopes, methane and He were processed in the closed loop FCU. The electrolytic reactor and the tubular reservoir tank worked as designed successfully, and the entire loop exhibited efficient impurity processing performance. The concentration of methane was observed to decrease sharply in every processing by the electrolytic reactor from 2.3% to less than 12ppm finally.

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.

Effect of tritium on corrosion behavior of chromium in 0.01 N sulfuric acid solution

Abstract The effects of tritium on the corrosion behavior of chromium were investigated in the present study, since it was suggested in the previous studies that the elution of chromium, which is one of the main constituent elements of passive layer of SUS 304 stainless steel, during passivation through further oxidation induced by oxidative radiolysis products would be the key reaction for the enhancement of the corrosion of SUS304 stainless steel induced by tritium. As the experimental results of the dependence of both dissolved oxygen and tritium concentration on the anodic behavior of chromium, it was found that the self-passivation of chromium induced by dissolved oxygen was inhibited in the tritiated solution, as found in the previous studies for SUS304 stainless steel. Therefore, it was considered that the elution of chromium by highly oxidative radiolysis products would induce a passivation inhibitory effect onto SUS304 stainless steel in a tritiated solution, resulting in an enhancement of the corrosion.

Tritium Safety Study Using Caisson Assembly (CATS) at TPL/JAEA

Abstract Tritium confinement is required as the most important safety function for a fusion reactor. In order to demonstrate the confinement performance experimentally, an unique equipment, called CATS: Caisson Assembly for Tritium Safety study, was installed in Tritium Process Laboratory of Japan Atomic Energy Agency and operated for about 10 years. Tritium confinement & migration data in CATS have been accumulated and dynamic simulation code was accumulated using these data. Contamination & decontamination behavior on various materials and new safety equipment functions have been investigated under collaborations with a lot of laboratories and universities.

Self-Decomposition Behavior of High Concentration Tritiated Water

Abstract In order to handle high-level tritiated water (HTO) safely, the self-decomposition behavior has been investigated as functions of tritium concentration (from 16 GBq/cm3 to 2 TBq/cm3) and storage temperature (269K ˜ 303K). The representative decomposition products such as H2 in the gas phase and H2O2 in the liquid phase were measured periodically, storing HTO in a leak-tight vessel. The effective production rate of H2 increased with tritium concentration, however, the normalized production rate by tritium decay, like effective G-value, decreased with tritium concentration. The effective production rate of H2O2 also increased with tritium concentration and the normalized one also decreased under consideration of its natural decomposition rate, though it thought that the almost H2O2 calculated by the reported G-value decomposed by extra stimulus in tritiated water. The effective production rates of H2 and H2O2 increased with temperature.