S. Inagaki

Progress summary of LHD engineering design and construction

In March 1998, the LHD project finally completed its eight year construction schedule. LHD is a superconducting (SC) heliotron type device with R = 3.9 m, ap = 0.6 m and B = 3 T, which has simple and continuous large helical coils. The major mission of LHD is to demonstrate the high potential of currentless helical-toroidal plasmas, which are free from current disruption and have an intrinsic potential for steady state operation. After intensive physics design studies in the 1980s, the necessary programmes of SC engineering R&D was carried out, and as a result, LHD fabrication technologies were successfully developed. In this process, a significant database on fusion engineering has been established. Achievements have been made in various areas, such as the technologies of SC conductor development, SC coil fabrication, liquid He and supercritical He cryogenics, development of low temperature structural materials and welding, operation and control, and power supply systems and related SC coil protection schemes. They are integrated, and nowadays comprise a major part of the LHD relevant fusion technology area. These issues correspond to the technological database necessary for the next step of future reactor designs. In addition, this database could be increased with successful commissioning tests just after the completion of the LHD machine assembly phase, which consisted of a vacuum leak test, an LHe cooldown test and a coil current excitation test. These LHD relevant engineering developments are recapitulated and highlighted. To summarize the construction of LHD as an SC device, the critical design with NbTi SC material has been successfully accomplished by these R&D activities, which enable a new regime of fusion experiments to be entered.

Energy confinement and thermal transport characteristics of net current free plasmas in the Large Helical Device

The energy confinement and thermal transport characteristics of net current free plasmas in regimes with much smaller gyroradii and collisionality than previously studied have been investigated in the Large Helical Device (LHD). The inward shifted configuration, which is superior from the point of view of neoclassical transport theory, has revealed a systematic confinement improvement over the standard configuration. Energy confinement times are improved over the International Stellarator Scaling 95 by a factor of 1.6 ± 0.2 for an inward shifted configuration. This enhancement is primarily due to the broad temperature profile with a high edge value. A simple dimensional analysis involving LHD and other medium sized heliotrons yields a strongly gyro-Bohm dependence (T E Ω ρ *-3.8 ) of energy confinement times. It should be noted that this result is attributed to a comprehensive treatment of LHD for systematic confinement enhancement and that the medium sized heliotrons have narrow temperature profiles. The core stored energy still indicates a dependence of T E Ω ρ *-2.6 when data only from LIED are processed. The local heat transport analysis of discharges dimensionally similar except for ρ * suggests that the heat conduction coefficient lies between Bohm and gyro-Bohm in the core and changes towards strong gyro-Bohm in the peripheral region. Since the inward shifted configuration has a geometrical feature suppressing neoclassical transport, confinement improvement can be maintained in the collisionless regime where ripple transport is important. The stiffness of the pressure profile coincides with enhanced transport in the peaked density profile obtained by pellet injection.

Extension and characteristics of an ECRH plasma in LHD

One of the main objectives of LHD is to extend the plasma confinement database for helical systems and to demonstrate such extended plasma confinement properties to be sustained in the steady state. Among the various plasma parameter regimes, the study of confinement properties in the collisionless regime is of particular importance. Electron cyclotron resonance heating (ECRH) has been extensively used for these confinement studies of LHD plasma from the initial operation. The system optimizations including the modification of the transmission and antenna system are performed with special emphasis on the local heating properties. As a result, a central electron temperature of more than 10?keV with an electron density of 0.6 ? 1019?m?3 is achieved near the magnetic axis. The electron temperature profile is characterized by a steep gradient similar to those of an internal transport barrier observed in tokamaks and stellarators. The 168?GHz ECRH system demonstrated efficient heating at densities more than 1.0 ? 1020?m?3. The continuous wave ECRH system is successfully operated to sustain a 756?s discharge.

MHD instabilities and their effects on plasma confinement in Large Helical Device plasmas

Characteristics of MHD instabilities and their impacts on plasma confinement are studied in current free plasmas of the Large Helical Device. Spontaneous L?H transition is often observed in high beta plasmas close to 2% at low toroidal fields (Bt ? 0.75?T). The stored energy starts to rise rapidly just after the transition accompanying the clear rise in the electron density but quickly saturates due to the growth of the m = 2/n = 3 mode (m and n: poloidal and toroidal mode numbers), the rational surface of which is located in the edge barrier region, and edge localized mode (ELM) like activities having fairly small amplitude but high repetition frequency. Even in low beta plasmas without L?H transitions, ELM-like activities are sometimes induced in high performance plasmas with a steep edge pressure gradient and transiently reduce the stored energy up to 10%. Energetic ion driven MHD modes such as Alfv?n eigenmodes (AEs) are studied in a very wide range of characteristic parameters (the averaged beta of energetic ions, ?b?, and the ratio of energetic ion velocity to the Alfv?n velocity, Vb?/VA), of which range includes all tokamak data. In addition to the observation of toroidicity induced AEs (TAEs), coherent magnetic fluctuations of helicity induced AEs (HAEs) have been detected for the first time in NBI heated plasmas. The transition of a core-localized TAE to a global AE (GAE) is also observed in a discharge with temporal evolution of the rotational transform profile, having a similarity to the phenomenon observed in a reversed shear tokamak. At low magnetic fields, bursting TAEs transiently induce a significant loss of energetic ions, up to 40% of injected beams, but on the other hand play an important role in triggering the formation of transport barriers in the core and edge regions.

Plasma confinement studies in LHD

The initial experiments on the Large Helical Device (LHD) have extended confinement studies on currentless plasmas to a large scale (R = 3.9 m, a = 0.6 m). Heating by NBI of 3 MW produced plasmas with a fusion triple product of 8 × 1018m-3keVs at a magnetic field strength of 1.5 T. An electron temperature of 1.5 keV and an ion temperature of 1.1 keV were achieved simultaneously at a line averaged electron density of 1.5 × 1019 m-3. The maximum stored energy reached 0.22 MJ with neither unexpected confinement deterioration nor visible MHD instabilities, which corresponds to β = 0.7%. Energy confinement times reached a maximum of 0.17 s. A favourable dependence of energy confinement time on density remains in the present power density (~40 kW/m3) and electron density (3 × 1019 m-3) regimes, unlike the L mode in tokamaks. Although power degradation and significant density dependence are similar to the conditions on existing medium sized helical devices, the absolute value is enhanced by up to about 50% from the International Stellarator Scaling 95. Temperatures of both electrons and ions as high as 200 eV were observed at the outermost flux surface, which indicates a qualitative jump in performance compared with that of helical devices to date. Spontaneously generated toroidal currents indicate agreement with the physical picture of neoclassical bootstrap currents. Change of magnetic configuration due to the finite β effect was well described by 3-D MHD equilibrium analysis. A density pump-out phenomenon was observed in hydrogen discharges, which was mitigated in helium discharges with high recycling.

High-ion temperature experiments with negative-ion-based neutral beam injection heating in Large Helical Device

High-Z plasmas have been produced with Ar and/or Ne gas fuelling to increase the ion temperature in Large Helical Device (LHD) plasmas heated with high-energy negative-ion-based neutral beam injection (NBI). Although the electron heating is dominant in the high-energy NBI heating, the direct ion heating power is significantly enhanced in low-density plasmas due to both an increase in the beam absorption (ionization) power and a reduction of the ion density in the high-Z plasmas. Intensive neon- and/or argon-glow discharge cleaning works well to suppress dilution of the high-Z plasmas with wall-absorbed hydrogen. As a result, the ion temperature increases with an increase in the ion heating power normalized by the ion density and reaches 10 keV. An increase in the ion temperature is also observed with the addition of centrally focused electron cyclotron resonance heating to a low-density and high-Z NBI plasma, suggesting improvement of the ion transport. The results obtained in the high-Z plasma experiments with high-energy NBI heating suggest that an increase in the direct ion heating power and improvement of the ion transport are essential to ion temperature rise, and that a high-ion temperature could be obtained as well in hydrogen plasmas with low-energy positive-NBI heating which is planned in the near future in the LHD.

Overview of long pulse operation in the Large Helical Device

The Large Helical Device is the worlds largest heliotron type helical system, with the plasma confining magnetic field being generated by only external superconducting coils. One of the main objectives of the LHD project is to sustain high temperature plasmas for a long time in steady state. The plasma vacuum vessel and the divertor are water cooled, and a heat load of 3 MW can be removed continuously. The NBI, ECH and ICRF heating systems, diagnostic instruments and data acquisition system are designed for long pulse operation. The present status of these systems and the recent experimental results of long pulse operation are reviewed. A steady state discharge with NBI was obtained for 35 s. The ECH discharge duration was extended to 120 s with a duty factor of 95%. Plasma sustainment by ICRF alone was achieved for 2 s. The performance of these long pulse operations is summarized.

LHD divertor experimental program

The LHD experiment has just begun. A scenario is presented for LHD divertor experiments. It includes development of LHD divertor components particularly efficient pumping system, local island divertor as a closed pumped divertor, simultaneous achievement of H-mode and radiative cooling (SHC operation) as an H-mode approach in the helical device, high temperature divertor plasma operation for enhancement of the energy confinement.

Improved plasma performance on Large Helical Device

Since the start of the Large Helical Device (LHD) experiment, various attempts have been made to achieve improved plasma performance in LHD [A. Iiyoshi et al., Nucl. Fusion 39, 1245 (1999)]. Recently, an inward-shifted configuration with a magnetic axis position Rax of 3.6 m has been found to exhibit much better plasma performance than the standard configuration with Rax of 3.75 m. A factor of 1.6 enhancement of energy confinement time was achieved over the International Stellarator Scaling 95. This configuration has been predicted to have unfavorable magnetohydrodynamic (MHD) properties, based on linear theory, even though it has significantly better particle-orbit properties, and hence lower neoclassical transport loss. However, no serious confinement degradation due to the MHD activities was observed, resolving favorably the potential conflict between stability and confinement at least up to the realized volume-averaged beta 〈β〉 of 2.4%. An improved radial profile of electron temperature was also achie...