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 are planned 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 82.6 GHz ECH at a heating power input of 0.35 MW. The magnetic field was set at 1.5 T in these campaigns so as to accumulate operational experience with the superconducting coils. In the second campaign, auxiliary heating with NBI at 3 MW has been carried out. Averaged electron densities of up to 6 × 10 19 m-3 , central temperatures ranging from 1.4 IAEA-F1-CN-69/OV1/4 2 to 1.5 keV and stored energies of up to 0.22 MJ have been attained despite the fact that the impurity level has not yet been minimized. The obtained scarling of energy confinement time has been found to be consistent with the ISS95 scaling law with some enhancement.

Abstract. Using the island divertors (IDs) of W7-AS and W7-X and the helical divertor (HD) of LHD as examples, the paper presents a comparative divertor transport study for three typical helical devices of different machine-size following two distinct divertor concepts, aiming at identifying common physics issues/effects for mutual validation and combined studies. Based on EMC3/EIRENE simulations supported by experimental results, the paper first reviews and compares the essential transport features of the W7-AS ID and the LHD HD in order to build a base and framework for a predictive study of W7-X. The fundamental role of low-order magnetic islands in both divertor concepts is emphasised. Preliminary EMC3/EIRENE simulation results for W7-X are presented and discussed with respect to W7-AS and LHD in order to show how the individual field and divertor topologies affect the divertor transport and performance. For instance, a high recycling regime, which is absent from W7-AS and LHD, is predicted to exist for W7-X. The paper focuses on identifying and understanding the role of divertors for high density plasma operations in helical devices. In this regard, special attention is paid to investigating the divertor function for controlling intrinsic impurities. Impurity transport behaviour and wall-sputtering processes of CX-neutrals are studied under different divertor plasma conditions. A divertor retention effect on intrinsic impurities at high SOL collisonalities is predicted for all the three devices. The required SOL plasma conditions and the underlying mechanisms are analysed in detail. Numerical results are discussed in conjunction with the experimental observations for high density divertor plasmas in W7-AS and LHD. Different SOL transport regimes are numerically identified for the standard divertor configuration of W7-X and the possible consequences on high density plasmas are assessed. All the EMC3-EIRENE simulations presented in this paper are based on vacuum fields and comparisons with local diagnostics are made for low-ß plasmas.

The low to high confinement transition has been observed on the large helical device ͓A. Iiyoshi, A. Komori, A. Ejiri et al., Nucl. Fusion 39, 1245 ͑1999͔͒, exhibiting rapid increase in edge electron density with sharp depression of H ␣ emission. The transition occurs in low toroidal field ͑B t = 0.5-0.75 T͒ discharges and are heated by high power neutral beam injection. The plasma thus has a relatively high value ͑ϳ1.5% ͒ of the volume averaged ␤ value. The electron temperature and density profiles have steep gradients at the edge region which has high magnetic shear but is at a magnetic hill. Formation of the edge transport barrier leads to enhanced activities of the interchange type of modes with m =2/n =3 ͑m , n are the poloidal and toroidal mode numbers͒ in the edge region. At present, these magnetohydrodynamic activities limit the rise of the stored energy; the resultant increment of the stored energy remains modest. © 2005 American Institute of Physics. ͓DOI: 10.1063/1.1843122͔Since the discovery of rapid transition from the low ͑L mode͒ to high confinement regime ͑H mode͒ in the axisymmetric divertor experiment ͑ASDEX͒, 1 the low to high confinement ͑LH͒ transition has been observed for the past two decades in various tokamak configurations. These include the double-and single-null poloidal diverter configurations and also limiter configurations with circular and D-shaped cross sections. 2 The LH transition was also observed in stellarator or helical devices. These plasma are bounded by a limiter in the compact helical system ͑CHS͒ heliotron/torsatron 3,4 or by a limiter and/or an island chain in the Wendelstein 7-AS shearless stellarator. 5,6 Recently, the LH transition was observed also in spherical tori with very low aspect ratio. 7,8 Aside from high confinement, a universally observed signature of the transition was the formation of the edge pedestal and the edge transport barrier ͑ETB͒. Although many theoretical models of the transition in tokamaks and helical devices have been proposed, 9-11 the understanding of the LH transition mechanism and the formation of the edge pedestal is still insufficient. In particular, the magnetohydrodynamic ͑MHD͒ stability of a plasma with ETB has attracted much attention due to its impact on the possibility of sustaining a H-mode plasma with favorable divertor action at steady state. In tokamaks, the plasma is situated at a magnetic well which enhances the MHD stability of the plasma, especially in the edge region. The edge localized modes ͑ELMs͒ ͑Ref. 12͒ have variably been correlated with the stability of the ideal/resistive ballooning mode or kink/peeling mode. However, there is yet no complete understanding of the characteristics of the ELMs to allow their control during the operation of a reactor grade plasma.Therefore, achievement of the LH transition in large helical device ͑LHD͒, 13 which has a magnetic configuration different from those reported so far, provides insight to the underlying dynamics of the LH transitions and the accompanying behavior of the EL...

In the Large Helical Device (LHD), electron pressure profiles in gas-fueled high-density discharges tend to have a similar shape, as if these were frozen. This frozen profile is insensitive to variations in the magnetic field strength and moderate changes in the neutral beam heat deposition profile. At the same time, however, the absolute value of the electron pressure itself increases with the heating power, the electron density, and the magnetic field strength. In this study, a reference model for the electron pressure is proposed which consists of the frozen profile and parametric dependences derived from experimental observations. It is possible to define an operational regime where this typical profile appears by comparing the electron pressure profiles with this model. In the standard configuration, at which the maximum plasma stored energy in LHD has been obtained, the frozen profile appears in the plateau to the PfirshSchlüter regimes. As the collisionality decreases to the collisionless regime, the electron pressure becomes smaller than the prediction of the model and the deterioration is significant in the plasma core region. This tendency is enhanced in the configuration with the outward-shifted magnetic axis. The global energy confinement time, τ E , in the high-collisionality regime has a weaker density dependence together with the mitigated power degradation, scaling as τ E ∝ ne 0.28 P -0.43 (ne and P are the line-averaged density and the heating power, respectively), compared with the International Stellarator Scaling 95, where τ E ∝ ne 0.51 P -0.59 .

The progress of physical understanding as well as parameter improvement of net-current-free helical plasma is reported for the Large Helical Device since the last Fusion Energy Conference in Daejeon in 2010. The second low-energy neutral beam line was installed, and the central ion temperature has exceeded 7 keV, which was obtained by carbon pellet injection. Transport analysis of the high-Ti plasmas shows that the ion-thermal conductivity and viscosity decreased after the pellet injection although the improvement does not last long. The effort has been focused on the optimization of plasma edge conditions to extend the operation regime towards higher ion temperature and more stable high density and high beta. For this purpose a portion of the open helical divertors are being modified to the baffle-structured closed ones aimed at active control of the edge plasma. It is compared with the open case that the neutral pressure in the closed helical divertor increased by ten times as predicted by modelling. Studies of physics in a three-dimensional geometry are highlighted in the topics related to the response to a resonant magnetic perturbation at the plasma periphery such as edge-localized-mode mitigation and divertor detachment. Novel approaches of non-local and non-diffusive transport have also been advanced.