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Abstract: The permeation behavior of tritium through pure iron into the gas containing water vapor has been studied. It was found that the permeation rate of tritium was almost the same as that with no water vapor, and the major chemical form of tritium was HTO. It was also found that the permeation rate was decreased to ∼1/10 of the above value by plated with gold, and the major chemical form was HT. 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 (0.23GBq/cc). A set of data for an advanced chemical exchange column has been obtained as R&D on the tritiated water processing. The packing and catalyst were uniformly mixed into the advanced column. It has been proved that there was possibility that the advanced column has a large separation factor in comparison with the ordinary columns, where the packing and catalyst were alternately packed. Tritium durability tests were also carried out for the electrolysis cell of the chemical exchange column. The ion conductivity of the main material of the cell (Nafion) was hardly decreased even in the case of 1600kGy irradiation by gamma ray. [Copyright &y& Elsevier]

Abstract: In an ITER facility, about 3kg of tritium will be stored in about 50 hydride beds. The safe design and operation of tritium storage beds will be one of the most important points for enhancing overall safety in the facility. In the Tritium Process Laboratory at the Japan Atomic Energy Agency, many tritium storage beds with zirconium–cobalt (ZrCo) and uranium (U) have been used with/without self-accountancy measures, and 20 years of safe handling experience has been accumulated. From this experience, the key issues to be considered for the safety design are the effects of tritium decay, such as decay heat transfer and 3He behavior, which can cause an increase in temperature and pressure in the tritide vessel, along with the normal protection against over-temperature, over-pressure and leaks in the metal-hydride bed. Concerning the safety operation, the key issues are the hydrogenation–dehydrogenation cycle sequence under the requirements of the storage system and the emergency action, such as rapid hydrogen recovery and loss of normal cooling function. [Copyright &y& Elsevier]

Abstract: 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. [Copyright &y& Elsevier]

Abstract: The possibility of tritium self-sufficiency in a low aspect ratio (low-A) tokamak reactor with the outboard blanket only has been studied to simplify the inboard structure located in a high magnetic field. The tritium breeding ratio (TBR) of the outboard blanket increases by applying a neutron multiplier such as lead and beryllium to the inboard reflector. The effect in local TBR of a 40cm thick inboard reflector of lead is about 40% that of the inboard breeding blanket. The local TBR of a reactor with both the lead inboard reflector and outboard blanket using a Li2O breeder and a beryllium multiplier can be larger than 1.35 for aspect ratios less than 2.9. The local TBR of 1.35 satisfies the net TBR requirement of 1.05 where there is 78% coverage of the breeding zone over the total plasma-facing surface area. The result indicates there is a design solution achieving tritium self-sufficiency even in a low-A tokamak reactor with the outboard breeding blanket only, using an appropriate inboard reflector. Be12Ti and Li2TiO3 are recommended from the viewpoint of safety. In the combination, the aspect ratio is required to be less than 1.35 to achieve local TBR of 1.35. [Copyright &y& Elsevier]

Abstract: In order to establish the effective surface decontamination methods for the ITER construction materials, the so-called ‘soaking’ effect is important and more systematic data of tritium sorption and desorption for various materials are required under actually conditions. In order to obtain and summarize data on the amount of sorbed tritium on the various materials such as stainless steel (SS), acrylic resin, and epoxy paint, a series of tritiated water vapor exposure experiments have been carried out as a function of time. The materials were almost saturated with tritium within a period from several weeks to 1 month, and it was larger in the order of epoxy paint>acrylic resin=butyl rubber>SS304. The sorption rate and subsequent desorption rate for various materials were evaluated. [Copyright &y& Elsevier]

It is important to measure retentions and depth profiles of deuterium and tritium on plasma facing materials to understand plasma-wall interactions. The nuclear reaction analysis (NRA) has been established using a deuteron beam with an adjustable energy. By means of the NRA, depth profiles of deuterium and tritium on the surface of JT-60 divertor graphite tiles irradiated with DD burning were obtained. The NRA method showed that the average concentrations of the deuterium retained in the surface of two different were 1.3 × 1016 and 0.3 × 1016 cm-2. α-particles from the T(d, α)n reaction were measured from a dome unit tile of JT-60U W-shaped divertor and the measured activity of tritium was about 18 kBq/cm2. The peak energy of the α spectrum obtained with the NRA was about 1.2 MeV. This result implies that energetic tritons were implanted in the divertor graphite tile. [Copyright &y& Elsevier]

Abstract: Dust generated by plasma–wall interaction is a potential source of tritium retention in a fusion reactor. Evaluation of the dust accumulation in the entire vacuum vessel is required to estimate the total amount of tritium retention, but it was particularly difficult to measure for plasma-unexposed areas behind the PFC structures, i.e. “shadow areas”. Dust samples were collected at 3, 5 and 2–4 different toroidal locations on the first wall, divertor surface and the exhaust route under the divertor in JT-60U, respectively. On the tile surface, large mass area density was found at the inner divertor and baffle, in particular, upper tiles compared to the lower target tile where the thick deposition layers were produced. Mass area density was significantly increased at the shadow areas, i.e. under the divertor structure such as the divertor and baffle tiles and the divertor dome. It was found that the poloidal distribution is relatively symmetrical in the toroidal direction within a factor of three. In comparison with the previous collection just before major change of the plasma operations, dust accumulation was increased both at the exposed and shadow areas due to change in the operating conditions. [Copyright &y& Elsevier]

Abstract: The ventilation systems design for the ITER nuclear buildings ensures radioactive contamination is confined so that workers, the public and the environment are protected. Nuclear buildings are divided into confinement sectors which connect to the Heating, Ventilation and Air Conditioning (HVAC) system and detritiation system (DS). The Tokamak Complex DS provides centralized air purification for the building confinement sectors. A distributed arrangement of ventilation piping provides networks necessary for two key functions, these being Vent Detritiation (VD), to maintain sub-atmospheric pressure, and Air Detritiation (AD) to collect tritium released into the confinement sector. For the VD function, air extracted from the particular confinement sector is directed to the DS for processing prior to exhaust to the environment. This paper presents the configuration of the DS of the Tokamak Complex and addresses details of the design of the distributed piping network. Dynamic flow and pressure drop modelling has been applied to support the development of the system configuration and provide data for sizing the system and selecting components. Further design development is discussed in view of the safety requirements for operation of the system during design basis events such as earthquake or fire. [ABSTRACT FROM AUTHOR]

Abstract: The ITER Nuclear Buildings include the Tokamak, Tritium and Diagnostic Buildings (Tokamak Complex) and the Hot Cell and Low Level Radioactive Waste Buildings (Hot Cell Complex). The Tritium Confinement strategy of the Nuclear Buildings comprises key features of the atmosphere and vent detritiation systems and the heating, ventilation and air conditioning systems. The designs developed during the ITER EDA (engineering design activities) for these systems need to be adapted to the specific conditions of the Cadarache site and modified to conform with the regulatory requirements applicable to installations nucléaires de base (INB) – basic nuclear installations – in France. The highest priority for such adaptation has been identified as the Tritium Confinement of the Tokamak Complex and the progress in development of a robust, coherent design concept compliant with French practice is described in the paper. [Copyright &y& Elsevier]

Abstract: Erosion/deposition distribution and hydrogen isotope behavior in the JT-60U plasma-facing wall were investigated. Distribution of the tritium, which was produced by D–D nuclear reaction, was not correlated with erosion/deposition distribution. The tritium distribution can be explained by the distribution of high-energy tritium ion-implantation due to ripple loss. Deuterium distribution in the divertor region was different from the tritium distribution and not well correlated with the deposition. The highest D/C was ∼0.05 at the bottom of the outer dome wing, which is much less than that observed in other tokamaks. For the deuterium retention, at least two retention processes (ion-implantation and co-deposition) were found on the dome region. The systematic dust collection gave the small amount of dust (∼7g: 0.2mg/s production) in the whole vessel of JT-60U. Deposition was observed at the remote area of the outer divertor region. [Copyright &y& Elsevier]

This paper describes the design of the ITER tritium plant subsystems, layout in the tritium building and the construction plan. The tritium plant comprises tokamak fuel cycle processing systems, as well as tritium confinement and detritation systems. The plant processes tritiated gases received from the tokamak and other sources to produce the D, T gas streams for fuelling, and detritiates various waste streams including tritiated water before discharge to the environment. The plant has been designed to meet not only all anticipated plasma operation scenarios in the DD and DT phases with a wide range of burn pulse durations from short pulse (450 s) and long pulse (3000 s), but also safety requirements (minimization of equipment tritium inventory and environmental tritium release from different accidental events in tokamak and tritium processing subsystems, and reduction of workers’ tritium exposure, etc). [Copyright &y& Elsevier]

In order to reduce the tritium inventory in a fuel circulation system in fusion plants, a possibility of deuterium-tritium (DT) mixed neutral beam injectors with DT negative ion beams has been investigated by analyzing the beam optics of the DT-mixed negative ion beams for the first time. In a two-dimensional beam analysis, DT-mixed ion beams could be accelerated regardless of the DT current mixing ratio (T/D) from the identical acceleration geometry with the identical applied voltages by adjusting total space charge for the D and T components. However, in a three-dimensional beam analysis including magnetic fields, the DT-mixed ion beams were separated due to a mass dependence of the beam deflection by orthogonal magnetic fields of an electron suppression and transverse filter fields for negative ion production. In the case of the beamline for ITER, which requires beam deflection angle within ±1 mrad, the operational range for the DT-mixed ion beams was found to be the DT current mixing ratio of 0.7-1.5 and beam energy of 0.9-1.0 MeV. Although the operational range was not wide, the DT-mixed beam was found to be feasible in terms of the beam acceleration. These analyses also contribute to the study of the acceleration of other mixed ion beams such as impurity or heavy ions, which are useful techniques for diagnostic beams. [ABSTRACT FROM AUTHOR]

Abstract: An experimental study on tritium (T) transfer in porous concrete for the tertiary T safety containment is performed to investigate (i) how fast HTO penetrates through concrete walls, (ii) how well concrete walls contaminated with water-soluble T are decontaminated by a solution-in-water technique, and (iii) how well hydrophobic paint coating works as a protecting film against HTO migrating through concrete walls. The experiment is comparatively carried out using disks of cement paste which W(water)/C(cement) weight ratio is 0.6:1 with or without hydrophobic paints, and mortar disks which W/C/sand ratio is 0.6:1:2 with or without the paints. The hydrophobic paints tested in the present study are an epoxy polymer resin paint and an acrylic-silicon polymer resin one. After T exposure during specified time under a constant HTO vapor pressure in an acrylic box, the amount of water-soluble HTO on/in the disks is determined using a technique of H2O dissolution during specified time. The results obtained here are summarized as follows: (1) HTO penetration in porous concrete can be correlated in terms of the effective diffusivity. (2) Its value in porous cement without coating is 1.2×10−11 m2/s at 25°C. (3) HTO penetrates only through pores in cement, and there is no path for HTO transfer in non-porous sand. (4) Rates of sorption and dissolution of HTO in disks of cement and mortar coated with the epoxy resin paint are correlated in terms of the effective diffusivity through the paint film which value is D T =1.0×10−16 m2/s. The rate-determining step is diffusion through the paint. (5) The epoxy resin paint works more effectively as an anti-HTO diffusion coating. (6) Another acrylic-silicon resin paint does not work well as anti-HTO diffusion coating. This may be because the hydrophobic property of the silicon resin paint is deteriorated with elongating the contact time with H2O vapor or liquid. (7) The HTO uptake inside the epoxy paint is greater than that of the silicon one. (8) The permeation reduction factor (PRF) of HTO for the epoxy paint at steady-state is expected large, if HTO vapor only contributes to diffusion. However, when concrete surfaces coated with the epoxy paint are under wet conditions, the PRF value becomes smaller. All those results can be used to estimate the effect of HTO soaking in concrete walls in case of accidental T release in a fusion reactor room and to decontaminate wastes of tritiated concrete. [Copyright &y& Elsevier]

Abstract: The tritium permeation behavior through F82H steel from approximately pure tritiated water vapor diluted with He was investigated and the as-received and the permeation-experimented sample surfaces were analyzed by means of SEM, EDX and XRD. Extremely slow but obvious tritium permeation through F82H steel could be observed. It is indicated from the surface analyses that an oxide layer composed of magnetite and hematite is formed during the permeation experiments. The permeabilities estimated using the partial pressure of molecular tritium under the extreme assumption that the redox reaction among tritiated water vapor, magnetite and hematite with forming molecular tritium reaches equilibrium indicate that apparent tritium permeation using tritiated water vapor for the inlet side would be proceeded by molecular tritium generated by the redox reaction, although it is expected that actual tritium permeation would be derived from dissociated tritium atom indicated in transition state during the redox reaction. [Copyright &y& Elsevier]

Tritium and deuterium are expected to be the fuel for the first fusion power reactors. Being radioactive, tritium is a health, safety and environment concern. Room air tritium clean systems can be used to handle tritium that has been lost to the room from primary or secondary containment. Such a system called the Experimental Tritium Cleanup (ETC) systems is installed at the Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory. The ETC consists of (1) two compressors which draw air from the room, (2) a catalyst bed for conversion of tritium to tritiated water, and (3) molecular sieve beds for collection of the water. The exhaust from this system can be returned to the room or vented to the stack. As part of the US–Japan fusion collaboration, on two separate occasions, tritium was released into the 3000 m3 TSTA test cell, and the ETC was used to handle these releases. Each release consisted of about one Curie of tritium. Tritium concentrations in the room were monitored at numerous locations. Also recorded were the HT and HTO concentrations at the inlet and outlet of the catalyst bed. Tritium surface concentrations in the test cell were measured before and at a series of times after the releases. Surfaces included normal test cell equipment as well as idealized test specimens. The results showed that the tritium became well-mixed in the test cell after about 45 min. When the ETC was turned on, the tritium in the TSTA test cell decreased exponentially as was expected. The test cell air tritium concentration was reduced to below one DAC (derived air concentration) in about 260 min. For the catalyst bed, at startup when the bed was at ambient temperature, there was little conversion of tritium to HTO. However, once the bed warmed to about 420 K, all of the tritium that entered the bed was converted to HTO. Immediately after the experiment, surfaces in the room initially showed moderately elevated tritium concentrations. However, with normal ventilation, these concentrations soon returned to routine levels. The data collected and reported here should be useful for planning for the operation of existing and future tritium facilities. [Copyright &y& Elsevier]