Nobuyoshi Ohyabu

Design Study for the Large Helical Device

The Large Helical Device (LHD) is a Heliotron/torsatron-type superconducting helical confinement fusion device. The design study is described. The goal of the LHD is to demonstrate high energy confinement and high beta in a helical device, which are necessary steps toward a helical reactor system.

The divertor plasma characteristics in the Large Helical Device

Divertor plasma characteristics in the Large Helical Device (LHD) have been investigated mainly by using Langmuir probes. The three-dimensional structure of the helical divertor, which is naturally produced in the heliotron-type magnetic configuration, is clearly seen in the measured particle and power deposition profiles on the divertor plates. These observations are consistent with the numerical results of field line tracing. The particle flux to the divertor plates increases almost linearly with the line averaged density. The high-recycling regime and divertor detachment, which are observed in tokamaks, have not been observed even during high density discharges with low input power. Both electron density and temperature decrease with increasing radius in the stochastic layer with open field lines, and at the divertor plate they become fairly low compared with those at the last closed flux surface. This means the reduction of pressure along the magnetic field lines occurs in the open field line region in LHD.

The divertor program in stellarators

Two significant problems that need to be solved for any future fusion device are heat removal and particle control. A very promising method to attack these problems in tokamaks and helical devices is the use of a divertor, providing a controlled interaction zone between plasma and wall. By carefully designing a divertor, conditions can be created in front of the divertor targets, which lead to a sufficient reduction of the power load on the targets by strong radiation redistribution. Any solution of course needs to allow for an energy confinement which is at least sufficient for the realization of a fusion reactor. Since energy confinement has been found to be strongly related to edge anomalous transport and edge plasma profiles, the ultimate aim is to find an integral solution which is optimum with respect to exhaust, heat load and energy confinement. Two different types of divertors are presently being investigated in helical devices: the `helical divertor and the `island divertor. So far divertor concepts have been investigated only in a few helical devices. Theoretical and experimental efforts have mainly concentrated on the suitability of divertor magnetic field structures, while detailed studies of the divertor plasma properties for the two types of divertor configurations have only recently begun. In the course of this exploration, a promising new high-density H-mode (HDH) plasma operational regime has been discovered on the Wendelstein stellarator W7-AS. It benefits from high-energy (up to twice the value of the International Stellarator Scaling ISS95) and low impurity confinement times, complemented by edge radiated power fractions of up to 90% in detached regimes. This allowed quasi-steady-state operation for up to 50 energy confinement times and so far was only constrained by machine operability.

Ergodic edge region of large helical device

Ergodicity in various vacuum magnetic configurations in large helical device was quantitatively estimated, using the Kolmogorov length as a measure. It is found that the edge electron temperature profile changes its gradient at the position where the ergodicity begins to increase several cm outside the last closed flux surface (LCFS). In the profiles no distinguished change of plasma parameters was seen at the LCFS, which suggests the existence of a region just outside the LCFS where the confinement performance is relatively as good as that in the closed region. The radial electron heat conductivity was also estimated using a simple energy balance equation and it was confirmed that the conductivity is strongly affected by the ergodicity. Using the relationship between ergodicity and electron heat conductivity, a new practical definition of the LCFS based on the experiments is proposed.

Review of initial experimental results of the PSI studies in the large helical device

The large helical device (LHD) is the largest heliotron type superconducting device. Its operation was started on 31 March 1998. Three experimental campaigns have been completed until the end of 1999. Wall conditioning mainly by cleaning discharges using ECRF or glow discharges worked well even without high temperature baking. The plasma production with ECRH and auxiliary heating with NBI and/or ICRF in the LHD configuration equipped with open helical divertor were well performed. The divertor material was SS316L in the first and second campaigns, and was replaced by the graphite in the third campaign. The influences of the different divertor materials were investigated. Our understanding of the edge and the divertor plasma has progressed. Long-pulse discharges 80 and 68 s heated by NBI (0.5 MW) or ICRF (0.9 MW) have been achieved, respectively. No severe limitation of the duration has appeared.

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.

Characteristics of Edge Magnetic Field Structure in LHD Heliotron

Results of numerical analyses of edge and divertor magnetic field structure are presented for the Large Helical Device (LHD). For the analyses, a field line tracing code has been developed to include the finite beta effect and cross-field particle diffusion, simulating the anomalous transport. Precise analyses of the helical divertor under finite beta conditions were performed for the first time. It is found that, in the magnetic axis shift operation from 3.6 m to 3.9 m, striking points of divertor legs shift more than 250 mm on target plates. Meanwhile very small shift less than 15 mm can be seen with increased beta up to ∼2%, in spite of the large magnetic axis shift more than 300 mm. Although the shift of striking point is small under finite beta conditions, distortion in the fine structure of a divertor leg is taken place, which may affect the edge particle transport during high beta discharges.

Behavior of helium gas in the LHD vacuum chamber

Abstract In general, helium gas does not remain on vacuum chamber walls because of its small activation energy. However, outgassing of helium gas from the walls has been observed in the LHD plasma vacuum chamber with a very long time constant after helium glow discharge cleaning (GDC), and absorption of helium atoms has been observed in plasma discharge experiments using helium gas. The helium partial pressure before the daily experiments was determined only by the gas species of the last GDC. No dependence has been found between the helium partial pressure and the species of the fueling gas of the last plasma experiment fueling gas. Considering that the outgassing rate of the helium gas is almost the same each morning after He GDC, the retention of helium atoms in the wall after the GDC is almost at the same level. The concentration of helium atoms in the wall before the daily experiments is estimated. The outgassing rate after the GDC is 2×10 −4 Paxa0m 3 /s and the concentration is 4.7×10 16 atoms/cm 2 . These results are of the same order as in another experiments. During helium gas plasma experiments, about a half of the amount of the inlet gas disappears with the missing particles remaining in the wall. The stainless steel wall, which is saturated with He GDC, may still have the capacity to trap high energy helium atoms. However, the energy dependence of trapping helium atoms presently is not clear.

Effect of Magnetic Ergodicity on Edge Plasma Structure and Divertor Flux Distribution in LHD

Strong nonuniformity in the helical divertor flux distribution is seen both in numerical calculations and in experiments. Flux profiles along the divertor trace on the vacuum vessel change according to vacuum magnetic configurations. Diverted particles choose one of two X-points as the escaping channel to divertor legs, which is determined by the local magnetic shear just inside the X-point. Ergodicity inside both X-points is also estimated quantitatively by a measure of Kolmogorov length for each magnetic configuration. The relationship between ergodicity and the choice of X-point for diverted particles is also discussed.

One-dimensional simulation on stability of detached plasma in a tokamak divertor

The stability of the radiation front in the scrape-off layer (SOL) of a tokamak is studied with a one-dimensional fluid code; the time-dependent transport equations are solved in the direction parallel to a magnetic field line. As the energy from the core into the SOL plasma is reduced, stable attached solutions change into stable detached solutions. Whenever such stable detached states are attained, the strong radiation front is in contact with or at a small distance from the divertor target. When the input energy is decreased further and total losses become larger than the input power, the radiation front starts to move towards the X-point, cooling the SOL plasma. In such a case, no solutions for the radiation front resting in the divertor channel are observed in our parameter space. This qualitatively corresponds to the results of tokamak divertor experiments which show the movement of the radiation front.