S. Tanahashi

Overview of the Large Helical Device project

The Large Helical Device (LHD) has successfully started running plasma confinement experiments after a long construction period of eight years. During the construction and machine commissioning phases, a variety of milestones were attained in fusion engineering which successfully led to the first operation, and the first plasma was ignited on 31 March 1998. Two experimental campaigns were carried out in 1998. In the first campaign, the magnetic flux mapping clearly demonstrated a nested structure of magnetic surfaces. The first plasma experiments were conducted with second harmonic 84 and 82.6xa0GHz ECH at a heating power input of 0.35xa0MW. The magnetic field was set at 1.5xa0T in these campaigns so as to accumulate operational experience with the superconducting coils. In the second campaign, auxiliary heating with NBI at 3xa0MW has been carried out. Averaged electron densities of up to 6 × 1019m-3, central temperatures ranging from 1.4 to 1.5xa0keV and stored energies of up to 0.22xa0MJ have been attained despite the fact that the impurity level has not yet been minimized. The obtained scaling of energy confinement time has been found to be consistent with the ISS95 scaling law with some enhancement.

Initial physics achievements of large helical device experiments

The Large Helical Device (LHD) experiments [O. Motojima, et al., Proceedings, 16th Conference on Fusion Energy, Montreal, 1996 (International Atomic Energy Agency, Vienna, 1997), Vol. 3, p. 437] have started this year after a successful eight-year construction and test period of the fully superconducting facility. LHD investigates a variety of physics issues on large scale heliotron plasmas (R=3.9u200am, a=0.6u200am), which stimulates efforts to explore currentless and disruption-free steady plasmas under an optimized configuration. A magnetic field mapping has demonstrated the nested and healthy structure of magnetic surfaces, which indicates the successful completion of the physical design and the effectiveness of engineering quality control during the fabrication. Heating by 3 MW of neutral beam injection (NBI) has produced plasmas with a fusion triple product of 8×1018u200akeVu200am−3u200as at a magnetic field of 1.5 T. An electron temperature of 1.5 keV and an ion temperature of 1.4 keV have been achieved. The maximum s...

Large Helical Device (LHD) program

The largest superconducting fusion machine, Large Helical Device (LHD), is now under construction in Japan and will begin operation in 1997. Design and construction of related R&D programs are now being carried out. The major radius of this machine is 3.9 m and the magnetic field on the plasma center is 3 T. The NbTi superconducting conductors are used in both helical coils and poloidal coils to produce this field. This will be upgraded in the second phase a using superfluid coil cooling technique. A negative ion source is being successfully developed for the NBI heating of LHD. This paper describes the present status and progress in its experimental planning and theoretical analysis on LHD, and the design and construction of LHD torus, heating, and diagnostics equipments.

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.

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.

Overview of the Large Helical Device

The Large Helical Device (LHD) experiments have started after a construction period of eight years, and two experimental campaigns were performed in 1998. The magnetic field was raised up to 2.75 T at a magnetic axis position of 3.6 m at the end of the second campaign. In the third campaign, started in July in 1999, the plasma production with ECH of 0.9 MW and auxiliary heating with NBI of 3.5 MW have achieved an electron temperature of 3.5 keV and an ion temperature of 2.4 keV. The maximum stored energy has reached 0.75 MJ with an averaged electron density of 7.7×1019 m-3 by hydrogen pellet injection. The ICRF heating has sustained the plasma for longer than 2 s and the initial stored energy of the NBI target plasma has increased from 0.27 MJ to 0.335 MJ. The major characteristic of the LHD plasma is the formation of the temperature pedestal, which leads to some enhancement of energy confinement over the ISS95 scaling law. The confinement characteristic is gyro-Bohm and the maximum energy confinement has reached 0.28 s. The LHD has also shown its high potentiality for steady-state operation by realizing a 22 s discharge in the second campaign.

Experiments on NBI plasmas in LHD

Neutral beam injection (NBI) heating started in the second experimental campaign of the Large Helical Device (September to December 1998) by two tangential beam lines. With 100 keV hydrogen, the beam port through power of up to 3.7 MW was injected for 1 s typically. The energy confinement was systematically better than that predicted by the International Stellerator Scaling 95 up to a factor of 1.5. The temperature pedestal observed contributes to this enhancement. We have also demonstrated a long pulse discharge by NBI in the LHD. By injecting 0.7 MW of beam, a plasma with a density of 0.3 × 1019 m-3 was sustained for 22 s. A unique oscillating phenomenon of plasma quantities with a long time scale was observed in the long pulse discharge. One of the topics of NB discharge is that the plasma can be started up by NB alone. This technique is unique for helical systems that have a vacuum magnetic field confining high energy ions, and it is useful for helical systems to be free from the constraint of magnetic field strength that must coincide with the frequency required by electron cyclotron resonance heating (ECH).

The performance of ICRF heated plasmas in LHD

An ion cyclotron range of frequency (ICRF) heating experiment was conducted in the third campaign of LHD in 1999. 1.35 MW of ICRF power were injected into the plasma and 200 kJ of stored energy were obtained, which was maintained for 5 s by ICRF power only after the termination of ECH. The impurity problem was so completely overcome that the pulse length was easily extended to 68 s at a power level of 0.7 MW. The utility of a liquid stub tuner in steady state plasma heating was demonstrated in this discharge. The energy confinement time of the ICRF heated plasma has the same dependences on plasma parameters as those of the ISS95 stellarator scaling with a multiplication factor of 1.5, which is a high efficiency comparable to that of NBI. Such an improvement in performance was obtained by various means, including: (a) scanning of the magnetic field intensity and minority concentration, (b) improvement of particle orbits due to a shift of magnetic axis and (c) reduction of the number of impurity ions by means of titanium gettering and the use of carbon divertor plates. In the optimized heating regime, ion heating turned out to be the dominant heating mechanism, unlike in CHS and W7-AS. Owing to the high quality of the heating and the parameter range being extended far beyond that of previous experiments, the experiment can be regarded as the first complete demonstration of ICRF heating in stellarators.

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.

Experimental studies on NBI and ICRF heated plasmas in the large helical device

In the Large Helical Device (LHD), upgrading of the key hardware system since the last EPS conference has led to (1) higher Te (Te(0) = 4.4 keV at ne = 5.3×1018 m-3 and Pabs = 1.8 MW); (2) higher confinement (τE = 0.3 s, Te(0) = 1.1 keV at ne = 6.5×1019 m-3 and Pabs = 2.0 MW); (3) higher stored energy Wpdia = 880 kJ; and (4) the highest β value in helical devices (2.4% at B = 1.3 T). The energy confinement was systematically higher than that predicted by the International Stellarator Scaling 95 up to a factor of 1.6. Ion cyclotron range of frequencies (ICRF) power up to 1.3 MW was reliably injected into the plasma without significant impurity contamination and a plasma with stored energy of 200 kJ was sustained for 5 s by ICRF alone. Long pulse discharges greater than 1 min in duration have been successfully achieved with ICRF power alone and with neutral beam injection (NBI) power alone. With NBI heating an 80 s discharge was achieved with a heating power of 0.5 MW at 2.75 T. The electron density was maintained at around 1.6×1019 m-3 by controlled gas puffing. The central electron and ion temperatures were kept around 1.5 keV. With ICRF heating, a similar long pulse discharge was achieved for 68 s with a heating power of 0.85 MW. The sustained plasma parameters are: Wp~110 kJ, Te(0)~Ti(0) = 2.0 keV and ne = 1.0×1019 m-3. During these discharges, no increase in radiation power has been observed.