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An interesting high-density operational regime with an internal diffusion barrier (IDB) has been extended to the helical divertor configuration in the Large Helical Device. The IDB is characterized...

The Large Helical Device (LHD) is a heliotron-type device employing large-scale superconducting magnets to enable advanced studies of net-current-free plasmas. The major goal of the LHD experiment ...

Transport characteristics of the stochastic magnetic boundary of the Large Helical Device (LHD) are investigated, based on three-dimensional Monte-Carlo Braginskii-type fluid model code, EMC3, coup...

Heliotron reactors have several features suitable for a fusion power plant, such as no need for current drive, no plasma current disruptions, suitability for steady-state operation, and a wide spac...

Fuel layering of a cryogenic target with a conical laser guide such as the FIREX target is complicated because of its non-spherical symmetry appearance. To simplify the layering, a foam layer is pl...

The divertor performance of LHD is studied for the two configurations, LID and HD. It is shown that the both divertor configurations play important roles for obtaining high performance plasmas in LHD : the large pumping capability of the LID to keep the low edge density in the IDB-SDC plasma, the large wetted area and the flexibility of strike point sweep of HD to reduce the power load on the divertor plates in long pulse operations. The possible effect of the ergodic layer on impurity retention in divertor is discussed by using the 3D edge transport modelling. It is found that the drag force exerted by the plasma flow can dominate over the thermal force, providing the impurity retention effect. The further changes needed to improve the current divertor configurations are discussed. New divertor designs for the future upgrade of LHD and for a LHD-type reactor are presented.

The fuel layering process of a cryogenic target for the Fast Ignition Realization EXperiment (FIREX) project has been studied. A foam shell method is proposed as a fuel layering technique for this target design. The difficulty of the fuel layering comes from the aspherical target symmetry. In the case of the foam shell method, liquid fuel is directly infiltrated into a foam shell though a fuel feeder and is soaked up into the foam layer by capillarity. The fuel is then solidified and an ideal cryogenic target is formed. To date, the cryogenic system for the demonstration of the fuel layering was fabricated and subsequently modified to improve its cool-down performance. A dummy foam target has been utilized to study the fuel layering process using H 2 instead of D 2 and DT fuels. Liquid H 2 is supplied into the shell through a feeder with a 20 μm inner tip diameter. The solid H 2 quantity remaining in the shell was controlled by regulating both H 2 pressure and target temperature during solidification.

One of the characteristics of the heliotron-type magnetic configuration is that it has an intrinsic divertor structure (helical divertor). Particle control using a helical divertor configuration, to achieve improved confinement and sustainment of steady-state high-performance plasmas, is a major experimental goal in the Large Helical Device (LHD), the largest heliotron-type superconducting device, and it needs to be demonstrated on the route to the design of the heliotron-type fusion reactor. The LHD scrape-off layer (SOL) in the intrinsic helical divertor configuration has a unique magnetic field line structure consisting of stochastic regions, residual islands, whisker structures, and laminar layers contrasting with the "onion-skin"-like magnetic field line structure in poloidal divertor tokamak SOLs. Since the first experimental campaign in LHD in 1998, studies aiming at understanding the edge plasma properties in the "open" helical divertor configurations have been conducted experimentally and theoretically. In this paper, the helical divertor studies in the LHD are reviewed, and the future experimental plan is shown.

The long-pulse plasma discharge experiment is an important experiment in the Large Helical Device, which has a superconducting coil system and the capability of steady-state operation. The experiment of long-pulse plasma discharge was carried out using mainly ion cyclotron range of frequencies heating. The maximum plasma duration is 31 min and 45 s, and the total injected heating energy reached 1.3 GJ. Swing of the magnetic axis is adopted as an effective method to scatter the local heat load on the dive rtor plate during the discharge. The plasma was terminated abruptly by the influx of metallic impurities accompanied by a spark in the vacuum vessel.

Confinement of energetic ions from neutral beam injection heating is investigated by changing the magnetic field configuration of the Large Helical Device from a classical heliotron configuration t...

The Large Helical Device (LHD) is the largest super-conducting heliotron type device with two continuous helical winding coils. The first plasma shot in the LHD was carried out in March, 1998, and since then, plasma confinement experiments have been performed. As a future plan, a DD plasma experiment, which is an experiment with deuterium (D) beam injection into D plasma, has been proposed by researchers. Conducting such an experiment is essential for the progress in magnetic fusion research. In DD plasma experiments, tritium and neutrons are generated as a result of nuclear fusion reaction. The maximum annual amount of tritium is about 370 GBq. We are planning to remove tritium with an efficiency greater than 99 %. We conducted a tritium safety analysis and found that for inhabitants around the LHD facility, the annual dose of exposure to tritium is less than 1x10 -7 mSv, which is a result of off gases and effluent liquid being released into the environment. However, to perform DD experiments, we need to obtain public consent, which is not easy. The National Research Institute for Fusion Science (NIFS) is open to the public, and we at the NIFS make every effort to communicate to the public and the municipal governments information on radiation safety. The present report shows results of preliminary safety analysis and Problems with obtaining public consent to future experiments of the LHD.

A comprehensive study of the divertor plasma flow (DPF) distribution in the Heliotron-E helical device has been carried out by means of collector arrays under various experimental conditions. A strong inhomogeneity of DPF distribution is observed in all the investigated regimes. This inhomogeneity is characterized by a significant asymmetry of plasma flows recorded by symmetrically displaced collectors. The degree of asymmetry in horizontal (in-out) and in vertical (up-down) directions depends on characteristics of magnetic configuration as well as on experimental conditions: method of plasma heating, injected power and power absorbed by the plasma, plasma density, gas puffing time, pellet injection, etc. A certain degree of asymmetry also remains after termination of the active phase of discharge, i.e ., at the stage of decaying plasma. By comparison of asymmetry indices at the active stages of discharge and at the stage of afterglow plasma, the contribution of trapped particles into the asymmetry of DPF...

A new diagnostic method to measure ion energy distribution using ion reflection on the solid surface, one of the processes of plasma-wall interactions, is proposed. It is well known that the energy distribution of reflected particles is related to the energy distribution of incident ions. This method is used in relatively Iow ion temperature region such as edge plasma in fusion experimental devices. Incident ions to target plate which is inserted in the edge plasma are mainly reflected as neutral particles, and time-of-flight method is utilized to measure energy distribution of reflected particles. Energy distribution of incident ions is deduced by fitting with calculation using Monte Carlo simulation code TRIM.SP. The sheath potential in front of the target can also be obtained with this method.

The Large Helical Device (LHD) is a non-tokamak type superconducting toroidal fusion apparatus. The LHD project contributes to the physics and technological developments exploring the currentless and steady-state plasmas, which have the high performance of temperature, density, confinement time and beta values being extrapolatable to the reactor condition. This paper reviews the major progress of the SC magnet technology attained from R&D programs, design and construction of LHD.