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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 reducing the concentration of 14C produced in irradiation of reduced activation ferritic steels (RAFS) intended as structural materials in first wall and blanket of fusion power plants was investigated. During RAFS irradiation 14C is mainly produced from the more abundant of the two isotopes of nitrogen present in the steel, namely 14N. The method proposed consists in increasing the enrichment of the other isotope (15N). One-dimension transport calculations show that for a typical DEMO blanket configuration an 15N enrichment from the natural value (0.37%) to 95% was sufficient to keep the end of life 14C concentration in the RAFS below the limit (3.7×107 Bq/kg) fixed by the Japanese Nuclear Safety Commission for qualifying it as a low level material (LLM) which can be disposed by shallow land burial. [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 a fusion reactor, high-level tritiated water of more than one PBq/m3 will be generated and stored in various areas. The high-level tritiated water decomposes by itself and generates mainly hydrogen, becoming tritiated hydrogen peroxide water. In order to summarize safety requirements for long-term storage of high-level tritiated water, the characteristics, such as effective G-values of hydrogen and hydrogen peroxide, pH and Oxidation–Reduction Potential (ORP), have been investigated by storing tritiated water from 5TBq/m3 to 2EBq/m3 for 5 years. Because the effective G-value of hydrogen increases with decreasing tritium concentration, the storage tank for a wider range of tritium concentrations requires not only depression but, also, adequate processing of the cover gases. High-level tritiated water of more than 1PBq/m3 was acid and in some cases was in a corrosive condition; these characteristics have been maintained for 5 years. Therefore, the ORP–pH situation should be monitored periodically to avoid the corrosive condition. [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]

Abstract: Neutron transport calculations have been performed to evaluate the capability of metal hydrides and borohydrides as advanced shielding materials. Some hydrides are characterized by considerably higher hydrogen content than polyethylene and solid hydrogen. The hydrogen-rich hydrides show superior neutron shielding capability compared to the conventional materials. A one-dimensional cylindrical model was used to do the neutron transport calculation. The shields are located at the inboard and outboard sides, and they are 30 and 70cm in radial thickness, respectively. Mg(BH4)2, TiH2 and ZrH2 can reduce the thickness of the outboard shield by 23, 20 and 19%, respectively, compared to the combination of steel and water. Mixing Mg(BH4)2 with F82H produces large effects in the gamma-ray shielding. In the mixing ratio of 0.5, the gamma-ray flux was 300 times smaller than that of pure Mg(BH4)2, even though the fast neutron flux is 13% larger than that of pure Mg(BH4)2. In the neutron spectrum for Mg(BH4)2, the neutron flux in the energy range of less than 100eV drastically reduced as compared to those of other materials, due to the effect of boron. [Copyright &y& Elsevier]

Abstract: Construction has just begun in the JT-60 Super Advanced (JT-60SA) tokamak project, which is part of the Japanese domestic program and the Japan – EU international program “ITER Broader Approach”. All plasma facing components (PFCs) will be actively cooled because they are exposed to high heat flux for long duration. A lower single-null closed divertor with a vertical target (VT) will be installed when experiments begin. A divertor cassette integrated with coolant pipe connections will be used for remote handling maintenance. Each cassette covers a 10° sector in the toroidal direction. Modularized PFCs such as VTs, baffles, and domes will be assembled on divertor cassettes. Static structural analysis shows that the displacement and stress in the divertor module are generally small for determining the dead weight, coolant pressure, and electromagnetic forces (EMFs). However, the support structure of the outer baffle requires improvement because eddy currents generate concentrated overturning EMFs. [ABSTRACT FROM AUTHOR]

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]

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]

After the Fukushima Dai-ichi nuclear accident, a need for assuring safety of fusion energy has grown in the Japanese (JA) fusion research community. DEMO safety research has been launched as a part of Broader Approach DEMO Design Activities (BA-DDA). This paper reports progress in the fusion DEMO safety research conducted under BA-DDA. Safety requirements and evaluation guidelines have been, first of all, established based on those established in the Japanese ITER site invitation activities. The radioactive source terms and energies that can mobilize such source terms have been assessed for a reference DEMO concept. This concept employs in-vessel components that are cooled by pressurized water and built of a low activation ferritic steel (F82H), contains solid pebble beds made of lithium-titanate (Li 2 TiO 3 ) and beryllium–titanium (Be 12 Ti) for tritium breeding and neutron multiplication, respectively. It is shown that unlike the energies expected in ITER, the enthalpy in the first wall/blanket cooling loops is large compared to the other energies expected in the reference DEMO concept. Reference accident event sequences in the reference DEMO in this study have been analyzed based on the Master Logic Diagram and Functional Failure Mode and Effect Analysis techniques. Accident events of particular concern in the DEMO have been selected based on the event sequence analysis and the hazard assessment. [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: JT-60SA is a fully superconducting coil tokamak to be built under the framework of the EU-JA Broader Approach Agreement, and it aims to contribute to the complement of the ITER experiments and to the DEMO reactor design by the study of steady-state high-beta plasma experiments. The conceptual design of the JT-60SA tokamak and the peripheral systems has been carried out in close collaboration of EU and Japan from middle of 2006, and the JT-60SA Conceptual Design Report which involves an analysis of plasma physics and engineering design is summarized in May 2007. This paper intends to present an overall view of the JT-60SA project, and the latest design status of the key engineering issues such as superconducting magnets, divertor target, plasma heating devices, and magnet power supplies. [Copyright &y& Elsevier]

Abstract: The radiation shielding, nuclear heating, absorbed dose of the insulator and dose rate have been evaluated for the superconducting tokamak JT-60SA. The shielding structure consists of double wall using the 24mm thickness of SS316L filled with the 140mm thickness of borated water (40°C, 95% 10B enriched). The maximum nuclear heating of the winding at the inboard side of the superconducting TF coil was suppressed to be lower than 0.3mW/cc. The absorbed dose of the insulator on the in-vessel coil was estimated to be 8.5MGy for 10 years. The insulators are not necessary to be exchanged with 10 years operation on JT-60SA. The dose rate of JT-60SA site boundary is estimated to be 144μSv/3 months, and it satisfies the legal limit of 250μSv/3 months without any additional shielding performance for the JT-60 building. [Copyright &y& Elsevier]

Abstract: A new type of aid system for fusion reactor design, to which the virtual reality (VR) visualization and sonification techniques are applied, is developed. This system provides us with an intuitive interaction environment in the VR space between the observer and the designed objects constructed by the conventional 3D computer-aided design (CAD) system. We have applied the design aid tool to the heliotron-type fusion reactor design activity FFHR2m [A. Sagara, S. Imagawa, O. Mitarai, T. Dolan, T. Tanaka, Y. Kubota, et al., Improved structure and long -life blanket concepts for heliotron reactors, Nucl. Fusion 45 (2005) 258–263] on the virtual reality system CompleXcope [Y. Tamura, A. Kageyama, T. Sato, S. Fujiwara, H. Nakamura, Virtual reality system to visualize and auralize numerical imulation data, Comp. Phys. Comm. 142 (2001) 227–230] of the National Institute for Fusion Science, Japan, and have evaluated its performance. The tool includes the functions of transfer of the observer, translation and scaling of the objects, recording of the operations and the check of interference. [Copyright &y& Elsevier]

Abstract: Three options of fusion DEMO plant are proposed characterized by functions of the center solenoid (CS). The prime option uses a downsized CS, which does not provide sufficient V-s for plasma current ramp-up but supplies enough coil current for plasma shaping. This option produces a fusion output of 3GW with a major radius of 5.5m, aspect ratio of 2.6, normalized beta of 4.3 and maximum field of 16.4T. The estimated reactor weight is lighter than that of other conventional tokamak reactors, suggesting an economic advantage. The plant uses rather conservative technologies such as Nb3Al superconductor, water-cooled solid breeder blanket, low activation ferritic steel as the structural material and tungsten monoblock divertor plate. The design philosophy and key issues related to the constituent technologies of the plant are described in the present paper. [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]

Minimum sets of in-vessel components were installed for initial operation of JT-60SA in order to product and control the plasma and to know the plasma basic information, including inboard first wall, upper divertor, protection limiter, glow electrodes and magnetic sensors. All of magnetic sensors were installed within the required accuracy with in-situ measurement with laser tracker. The inboard first wall and the upper divertor were installed for the limiter configuration at the plasma initiation and the divertor configuration, respectively. They have carbon tiles as the first wall, which must be installed with accuracy of ± 1 mm in order to avoid the heat concentration. However, the vacuum-vessel wall had some deformation. Therefore, we measured in advance with laser tracker the position of the vacuum vessel where the bases of the inboard first wall and upper divertor would be welded. By using these position data and 3D CAD, we customized all bases. Owing to these procedures, we could install the first wall with required accuracy. Also, all other components were installed with required accuracy. [ABSTRACT FROM AUTHOR]