A cylindrical carbon pellet with a size of 1.0 mm^φ × 1.0 mm^L was injected into low-density (n_e =0.3−0.5 × 10¹³ cm^-3) neon-seeded NBI plasmas of the Large Helical Device (LHD). When the two NBI beam pulses were delayed just after the pellet injection, a large increment of central ion temperature up to5 keV was observed with the appearance of a peaked density profile (n_e0/⟨n_e⟩∼2.5) and enhanced toroidal rotation speed up to 35 km/s. Improvement of the ion transport is expected with the suggestion of a new operational scenario for confinement improvement in the LHD.

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We have constructed and operated an internal coil device with a high temperature superconductor. Three different types of Ag-sheathed Bi-2223 tapes are employed; i.e., a high critical current tape with a low silver ratio for the main HTS coil, a 0.3wt%Mn-doped one for the persistent current switch, and a 3at%Au-doped one for the coil-leads. Cold gas helium is provided by a GM refrigerator and supplied to the coil through a check valve, and the coil current is directly excited by the external power supply through removable electrodes. It took about 11 hours to cool the coil to 21 K from room temperature, and a nominal cable current of 118 A (overall coil current: 50 kA) was achieved. A decay time constant of the persistent current is a few tens of hours. Plasma experiments in a dipole configuration have been initiated.

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Low-energy ion beam technology is very interesting for material researchers. The advantages of this technology are the controllability of both the ion species and the energy of impinging ions, and the process in an ultra-high vacuum. However, the quality of the ion beam must be well considered during the designing process. Organometallic ion beam deposition which is a newly developed attractive technique for compound research is also described.

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Roles and requirements of the blanket system of the fusion power reactors are discussed from viewpoints of economics, fuel supply, generation system, maintenance, radioactive waste, and interaction with the plasma core. As the blanket system influences the cost of the fusion energy, the blanket system must be designed to minimize the fusion energy cost. Tritium breeding performance of the blanket is indispensable role to show the advantage of fusion energy on energy security. Material development for high temperature use under high neutron flux is one of the key issues of the generation system because the thermal efficiency depends on the coolant temperature of the blanket. Innovative maintenance technologies such as dividable superconducting coil system are very effective to make the fusion power reactor attractive. From viewpoints of natural resources and waste management, materials used in the fusion reactors should be recycled. Material selection is also of a large importance on the cost of radioactive waste disposal. Finally, it must be paid a careful attention that the design of the blanket system is inseparable from the achievement of a high performance plasma core.

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The main functions of the blanket in fusion power reactors are basically independent of the type of magnetic fusion reactor (tokamak, helical, etc.) and inertia fusion. However, from technical point of view, many candidate designs of blanket have been proposed depending on the particular reactor concepts. Their main features are characterized for the recent typical designs, and key issues are defined.

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This paper describes the status of blanket technology and material development for fusion power demonstration plants and commercial fusion plants. In particular, the ITER Test Blanket Module, IFMIF, JAERI/DOE HFIR and JUPITER-II projects are highlighted, which have the important role to develop these technology. The ITER Test Blanket Module project has been conducted to demonstrate tritium breeding and power generation using test blanket modules, which will be installed into the ITER facility. For structural material development, the present research status is overviewed on reduced activation ferritic steel, vanadium alloys, and SiC/SiC composites.

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Current topics concerning blanket technology are reviewed. In the chemical engineering/chemistry area, the qualitative and quantitative effects of mass transfer steps of tritium is important in the understanding of the behavior of bred tritium in the solid breeder blanket system. Such phenomena as adsorption, isotope exchange reactions, and water formation reaction at the grain surface produce profound effects on the behavior of the bred tritium in the blanket. Regarding the liquid system, the physical or chemical properties of Li, Li₁₇Pb₈₃ and Flibe as liquid blanket materials were compared. Some recent studies were introduced regarding tritium recovery from the liquid blanket materials, impurity removal from salts, ceramic coating of structural materials, and the vapor pressure of mixtures of metals or salts. Thermal hydraulic topics in relation to several candidate power reactor concepts are summarized. Emphasis is laid on the simultaneous removal of heat and tritium from the blanket and some aspects of forming effective power cycles are developed.

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Methods for heat transport analysis and heat transport simulation in toroidal plasmas are summarized on the basis of an energy balance equation. Joule heating, NBI heating, RF heating, and α heating are briefly explained. Among the energy loss mechanisms, the conduction loss and the radiation loss dominate in the core region and the peripheral region, respectively. In tokamaks, microturbulence causes anomalous transport, which is much larger than neoclassical transport. The other mechanisms of enhanced transport, sawtooth oscillation and magnetic island formation, are also shown.

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The time dependence of the current decay during the current quench phase of disruptions, which can significantly influence the electro-magnetic force on the in-vessel components due to the induced eddy currents, is investigated using data obtained in JT-60U experiments in order to derive a relevant physics guideline for the predictive simulations of disruptions in ITER. It is shown that an exponential decay can fit the time dependence of current quench for discharges with large quench rate (fast current quench). On the other hand, for discharges with smaller quench rate (slow current quench), a linear decay can fit the time dependence of current quench better than exponential.

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