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Three-dimensional (3D) effects on the stochasticity of magnetic field lines in helical devices are investigated. The stochastization due to increased plasma pressure is an intrinsic property of the stellarator/heliotron because the pressure-gradient driven currents are the source of perturbation fields. The stochastization depends on the plasma pressure profile. In an advanced stellarator, the stochastization due to the 3D effects is smaller. However, magnetic islands in the edge are affected by 3D effects. 3D effects in a tokamak plasma are studied,too, and differences from the vacuum approximation are found (© 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

To understand the edge plasma transport in the complex magnetic field lines structure in the LHD heliotoron, the heat and particle fluxes profiles on divertor plates were investigated experimentally and theoretically by using EMC3-EIRENE code. The heat flux profiles on a divertor plate in different magnetic configuration were measured and the main channel of the heat transport was revealed to be the field lines with long connection lengths which connect the divertor plate and the stochastic field lines layer. A comparison of the heat and the ion fluxes profiles on the divertor plates between the experimental data and the results of the calculation using EMC3-EIRENE code was conducted. It was revealed that the change of the cross-field transport in the edge plasma region causes not only the change of the heat and particle flux profiles on divertor plates, but also the changes of the intrinsically non-uniform global heat and particle deposition profiles on the helical diveror. The cross-field transport seems to be larger in the higher edge electron temperature discharge in the experimental conditions in this paper (© 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

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.

Self-sustained detachment has been obtained in the Large Helical Device (LHD). Strong hydrogen gas puffing of ~200 Pa m3 s−1 after a density feedback phase detaches the plasma from the divertor plates with high reproducibility. High electron density of over 1 × 1020 m−3 is sustained without gas puffing until the heating beam stops and a high-density flat top for 2 s has been demonstrated. Throughout the self-sustained detachment phase, the minor radius of the hot plasma column shrinks to ~90% of the last-closed-flux-surface, which corresponds to the rational surface. This new state has been named the 'Serpens mode', for self-regulated plasma edge 'neath the last-closed-flux-surface. Global energy confinement of the Serpens mode is compared with the international stellarator scaling 1995 (ISS95) and the recently established scaling for high-density LHD plasmas (HD scaling), where shrinking confinement volume and shallow penetration of the heating beams are taken into account. Although the energy confinement of the Serpens mode seems deteriorated compared with ISS95, as in the case of high-density attached plasmas, it is consistent with the HD scaling. This suggests that the energy confinement properties of detached plasmas in LHD are similar to those in high-density attached plasmas.

In the Large Helical Device (LHD), two different divertor configurations, i.e. helical divertor (HD) and local island divertor (LID), are utilized to control the edge plasma. The HD with two X-points is an intrinsic divertor for heliotron devices, accompanied with a relatively thick ergodic layer outside the confinement region. Edge and divertor plasma behavior from low density to high density regimes is presented, referring to the divertor detachment. The effect of the ergodic layer on the edge transport is also discussed. On the other hand, the LID is an advanced divertor concept which realizes a high pumping efficiency by the combination of an externally induced magnetic island and a closed pumping system. Experimental results to confirm the fundamental divertor performance of the LID are presented.

Conditions for the excitation of Alfven eigenmodes (AEs) by energetic ions are investigated in neutral-beam-injection (NBI) heated plasmas of the Large Helical Device (LHD). This study is carried out in a wide parameter range of the beta values of the energetic ion components and the ratio of the energetic ion velocity to the Alfven velocity (up to with the assumption of classical slowing down and ). These ranges of parameters cover those predicted for the International Thermonuclear Experimental Reactor (ITER). During this experimental campaign of LHD, toroidicity-induced AEs (TAEs) with n = 1–5 (n being the toroidal mode number), global AEs (GAEs) with n = 0 and 1, and energetic particle modes (EPMs) were observed. The effect of the magnetic configuration on the TAE spectrum was also investigated. In magnetic configurations with relatively high magnetic shear, only TAEs with n = 1 and 2 were observed. On the other hand, TAEs with n up to 5 were observed in magnetic configurations with low magnetic shear. For two typical shots obtained in magnetic configurations characterized by different values of the magnetic shear, eigenfunctions of TAEs were calculated by using a global mode analysis code CAS3D3. The calculated results indicate that the eigenfunctions tend to be localized around the relevant TAE gaps. When the gap is located in the plasma core region (normalized minor radius ρ ≤ 0.4), the TAE tends to become a core-localized type. When the gap is in the outer region (typically 0.5 ≤ ρ ≤ 0.9) of the plasma, the TAE tends to (a) either become a global type having a radially extended structure if the magnetic shear is very weak in the core region inside the gap, (b) or become a gap localized type in the case of finite central magnetic shear. Transition of the eigenmode from the core-localized type with m ~2/n = 1 TAEs (m being the poloidal mode number) to the n = 1 GAEs (or cylindrical AEs) has been observed when the rotational transform at the core ι (0)/2π exceeds the specific value of ι(0)/2π = 0.4.

Hydrogen ice pellets have been injected using two different location configurations, namely outer port injection and coil side injection, in the large helical device. The behaviour of the pellet ablation has been observed using a fast camera, which possesses high spatial and time resolution. Striking toroidal deflection of the pellet trajectory is observed. The deflection is in the direction of tangential neutral beam injection. The toroidal velocity ultimately reaches 1000 m s−1 or more. The possibility of a rocket effect due to a unilateral ablation by the fast ions is discussed. The effective penetration depth of the pellet, which is measured by images, agrees with the prediction from the neutral-gas-shielding model. The penetration depth is compared with the measured deposition profile of the pellet.

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.

Significant progress has been made with ion cyclotron range of frequencies (ICRF) heating in the Large Helical Device. This is mainly due to better confinement of the helically trapped particles and less accumulation of impurities in the region of the plasma core. During the past two years, ICRF heating power has been increased from 1.35 to 2.7 MW. Various wave-mode tests were carried out using minority-ion heating, second-harmonic heating, slow-wave heating and high-density fast-wave heating at the fundamental cyclotron frequency. This fundamental heating mode extended the plasma density range of effective ICRF heating to a value of 1×1020 m−3. This use of the heating mode was its first successful application in large fusion devices. Using the minority-ion mode gave the best performance, and the stored energy reached 240 kJ using ICRF alone. This was obtained for the inward-shifted magnetic axis configuration. The improvement associated with the axis-shift was common for both bulk plasma and highly accelerated particles. For the minority-ion mode, high-energy ions up to 500 keV were observed by concentrating the heating power near the plasma axis. The confinement properties of high-energy particles were studied for different magnetic axis configurations, using the power-modulation technique. It confirmed that with the inward-shifted configuration the confinement of high-energy particles was better than with the normal configuration. By increasing the distance of the plasma to the vessel wall to about 2 cm, the impurity influx was sufficiently reduced to allow sustainment of the plasma with ICRF heating alone for more than 2 min.

The behaviour of high-energy ions accelerated by an ion cyclotron range of frequency (ICRF) electric field in the large helical device (LHD) is discussed. A better confinement performance of high-energy ions in the inward-shifted magnetic axis configuration was experimentally verified by measuring their energy spectrum and comparing it with the effective temperature determined by an electron slowing down process. In the standard magnetic axis configuration a saturation of the measured tail temperature was observed as the effective temperature was increased. The ratio between these two quantities is a measure of the quality of transfer efficiency from high-energy ions to a bulk plasma; when this efficiency was compared with Monte Carlo simulations the results agreed fairly well. The ratio of the stored energy of the high-energy ions to that of the bulk plasma was measured using an ICRF heating power modulation method; it was deduced from phase differences between total and bulk plasma stored energies and the modulated ICRF heating power. The measured high energy fraction agreed with that calculated using the injected ICRF heating power, the transfer efficiency determined in the experiment and the confinement scaling of the LHD plasma.

Intrinsic impurity behaviour in neutral beam heated LHD long pulse discharges with a discharge duration of 10 s was investigated. Spectroscopic and bolometric measurements showed a remarkable temporal increase of core radiation due to metallic impurities. Central impurity accumulation was found in a narrow plasma density region (e = 1–3×1019 m−3) for both density ramp-up discharges and constant density discharges. In the density ramp-up discharges, high-Z impurities were accumulated in the low-density region and diffused out from the core plasma in the high-density region. The impurity behaviour was compared with neoclassical predictions.

Energetic tail formation by ion-cyclotron-range-of-frequencies (ICRF) heating was studied in the large helical device (LHD), by measuring high-energy charge-exchange atom spectra with a specially developed natural diamond detector, which views the LHD plasma perpendicularly. Comparison of measured effective perpendicular temperature of ICRF-driven H+ minority ions with the classical Stix model indicates that perpendicular ions with energy at least up to 150 keV are well confined in the centre region of the LHD. However, deviation from the classical prediction was observed for high-energy perpendicular protons in the outer region of the plasma, indicating ripple-induced transport and charge-exchange losses. With perpendicular fast atom spectrum and flux measurements, small yet notable differences were detected co-injected and counter-injected ICRF-driven beam ion confinement during combined neutral beam injection (NBI) and ICRF heating. These results can be explained by the differences in the orbit topology and ICRF-induced drift.

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.

A mild pedestal has been observed in the large helical device (LHD) discharges. Broad profile pedestal leads to enhancement of the energy confinement. The shapes of the electron temperature profile in the gas puff divertor discharges are very close to parabolic profile, nearly being independent of density, input power and magnetic field. The plasma tends to preserve the temperature scale length (or local shape), ∇T/T even with a large island or during radiative collapse.

Positive radial electric field, Er, was observed at the edge region in the low-density NBI plasmas of the Large Helical Device (LHD). Ion thermal diffusivities, χi, are evaluated by the local transport analysis for two different density plasmas and for two cases of the magnetic axis position. In the standard configuration case, χi was reduced by large positive Er, while no reduction in χi was found in the inward shifted case. A certain level of suppression of the neoclassical ripple transport by Er seems to be confirmed since the neoclassical transport is considered to be more dominant in the standard magnetic configuration plasmas.

The experimentally optimized magnetic field configuration of the Large Helical Device [A. Iiyoshi et al., Nucl. Fusion 39, 1245 (1999)], where the magnetic axis is shifted inward by 15 cm from the early theoretical prediction, reveals 50% better global energy confinement than the prediction of the scaling law. This configuration has been investigated further from the viewpoints of high energy particle confinement and magnetohydrodynamic (MHD) stability. The confinement of high energy ions is improved as expected. The minority heating of ion cyclotron range of frequency was successful and the heating efficiency was improved by the inward shift. The confinement of passing particles by neutral beam injection was also improved under low magnetic field strength, and there could be obtained an almost steady high beta discharge up to 3% in volume average. This was a surprising result because the observed pressure gradient exceeded the Mercier unstable limit. The observed MHD activities became as high as beta but they did not grow enough to deteriorate the confinement of high energy ions or the performance of the bulk plasma, which was still 50% better than the scaling. According to these favorable results, better performance would be expected by increasing the heating power because the neoclassical transport can also be improved there.

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.

Neutral beam injection heated discharges at the density limit in the Large Helical Device [O. Motojima, H. Yamada, A. Komori et al., Phys. Plasmas 6, 1843 (1999)] are terminated with asymmetric radiative collapse (ARC) exhibiting several properties in common with the MARFE (multifaceted asymmetric radiation from the edge) phenomenon: (1) A highly poloidally asymmetric radiation profile which is stronger on the inboard side. (2) This asymmetry is well correlated with the signal from the multichord interferometer. (3) Moreover, evidence from several diagnostics at different toroidal locations supports the possibility that ARC may be toroidally symmetric. However in contrast to MARFE, ARC is only observed in the period just prior to the quench of the plasma.

The confinement characteristics of large net-current-free plasmas heated by neutral-beam injection have been investigated in the Large Helical Device (LHD). A systematic enhancement in energy-confinement times from the scaling derived from the medium-sized heliotron/torsatron experiments have been observed, which is attributed to the edge pedestal. The core confinement is scaled with the Bohm term divided by the square root of the gyro radii. The comparative analysis using a dimensionally similar discharge in the Compact Helical System indicates gyro-Bohm dependence in the core and transport improvement in the edge region of LHD plasmas.

The generation and transport of impurity particles in a boundary plasma are analysed on the basis of a single particle motion together with a Monte Carlo method, where the Langevin-type equation of an impurity particle is used. The impurity generation and transport in a scrape-off layer are modelled after the Large Helical Device (LHD). The screening efficiency of the SOL for the impurity penetration into a core plasma, impurity redeposition on the first wall and the other effects are analysed from the statistics of traces of many impurites.

The Large Helical Device (LHD) is a heliotron/torsatron-type confinement device (B = 4 T, R = 3.9 m) equipped with a helical divertor. In the LHD configuration, the plasma region is shifted inwards approximately a third of the plasma minor radius relative to the center of the two pairing helical coils, thus making the distance between the coil center and the edge plasma small, 328 mm on the small major radius side of the torus. Within this space, many components must be installed, such as the superconducting helical coil (163), its coil can (45), the thermal shield (30), the vacuum gap (35), the plasma vacuum vessel (15), the first wall (25) (the numbers in the parentheses are allocated radial space in mm for the respective components). Effective edge plasma control by the divertor requires a space of more than 15 mm between the plasma and the first wall. To meet the above space requirement, the thickness of the helical coil is designed to be as small as possible, yet the coil current density and maximum field strength are within the limits of reliable coil operation.

"It is found that the remnant island structure created by n/m=1/1 resonant magnetic perturbation field in the stochastic magnetic boundary of the Large Helical Device (LHD)[A. Komori et al.,Nucl. Fusion 49, 104015 (2009)] has a stabilizing effect on formation of radiating plasma, realizing stably sustained divertor detachment operation with the core plasma being unaffected. The data from the several diagnostics, (profiles of electron temperature and density, radiation and temporal evolution of divertor particle flux)indicate selective cooling around X-point of the island and thus peaked radiation there, which is stabilized outside of the last closed flux surface throughout the detachment phase. The vacuum ultraviolet spectroscopy measurements of high Z impurity (iron)emission shows significant decrease during the detachment, indicating core plasmadecontamination. The results from the three-dimensional (3D) edge transport code, edge Monte Carlo 3D (EMC3)[Y. Feng et al., Contrib. Plasma Phys. 44, 57 (2004)]-EIRENE [D. Reiter et al.,Fusion Sci. Technol. 47, 172 (2005)] show similar tendency in the radiation pattern. The island sizeand its radial location are varied to investigate the magnetic topology effects on the detachment control. The divertor particle flux and neutral pressure exhibit intermittent oscillation as well asmodification of recycling pattern during the detachment, which are found to reflect the island structure."

A steady-state H-mode plasma with frequent ELMs (edge localized modes) was obtained by applying magnetic fields with high poloidal mode numbers resonant in the plasma edge. The induced ELMs limit the density and impurity accumulation normally observed during ELM-free H-mode. An EML coil with numerous resonant modes ( m 10) in the plasma edge appears the most effective. This method of control is extended to higher auxiliary heating power by increasing the magnitude of the magnetic perturbations.

The anomalous particle transport in a tokamak core is believed to be linked to the advection of magnetically trapped electrons alone, owing to the passing electrons maintaining a thermal equilibrium along the field lines. Surprisingly, in nonlinear numerical studies, the radial flux of passing electrons rivals that of the trapped ones. The strong interaction of passing electrons and electric fluctuations is mediated by long tails of the modes along the magnetic field, which are generated by the passing electrons in the first place.

The confinement characteristics of large net-current-free plasmas heated by neutral-beam injection have been investigated in the Large Helical Device (LHD). A systematic enhancement in energy-confinement times from the scaling derived from the medium-sized heliotron/torsatron experiments have been observed, which is attributed to the edge pedestal. The core confinement is scaled with the Bohm term divided by the square root of the gyro radii. The comparative analysis using a dimensionally similar discharge in the Compact Helical System indicates gyro-Bohm dependence in the core and transport improvement in the edge region of LHD plasmas.

Recent progress in the heating capability in the large helical device [O. Motojima et al., Phys. Plasmas 6, 1843 (1999)] has allowed the highest average beta value (4.1%) obtained in the helical devices, and enables exploration of magnetohydrodynamics (MHD) stability in this beta region. MHD activities in the periphery are found to become stable spontaneously from the inner region to the outer region when the averaged beta value exceeds a threshold, and then a flattening of the electron temperature profile is observed around the resonant surface. Such a flattening can be formed externally by producing an m/n=1/1 magnetic island, and the complete stabilization of the m/n=1/1 mode is demonstrated by the moderate island width. In addition, attempts to control peripheral plasmas are also performed by using a limiter and a local island divertor utilizing the m/n=1/1 island, to improve plasma confinement and, especially, to stabilize pressure-driven modes in the present study. The stabilization of peripheral MHD modes is obtained with both approaches, and this indicates that these are available to the production of higher-beta plasmas without edge MHD activities.

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 (1999)], exhibiting rapid increase in edge electron density with sharp depression of H_alpha emission. The transition occurs in low toroidal field (B_t = 0.5?0.75 T) discharges and are heated by high power neutral beam injection. The plasma thus has a relatively high value (~1.5%) of the volume averaged beta 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 numbers) 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.

It is proposed to expand the outermost flux surfaces in tokamaks to provide a large volume for radiative cooling. The radiative power at the boundary is enhanced by the effects of plasma flow as well as by a volumetric factor, and the resultant edge cooling and reduced heat load on the limiter may significantly retard impurity generation.

An analytic derivation is presented of a condition for thermal stability of a radiatively cooled tokamak boundary. Application of these results to the expanded boundary configuration shows that such plasmas are thermally stable over the parameter regime of interest.

An experiment is performed to study the magnetic energy dissipation and ion heating in a magnetic neutral point discharge which is related to both an explosive energy dissipation in space plasma phenomena and the ion heating in thermonuclear fusion research.A plasma current sheet is created by exciting a plasma current along the magnetic neutral line of a two dimensional quadrupole magnetic field. A sudden anomalous increase of resistivity is observed which results in fast energy dissipation and strong plasma heating. The sudden increase of the plasma resistivity appears earlier and its magnitude increases as, i) the plasma density is decreased, ii) the discharge plasma current is increased or iii) the external quadrupole magnetic field is increased. When the sharp rise of the resistivity appears, the electron and ion temperatures rise abruptly up to several keV range Within 0.5μsec and about 10% of the initial energy Stored in the capacitor (for the plasma current excitation) is converted into ion thermal energy. This strong and efficient ion heating cannot be explained by a simple classical process and the role of turbulent waves should be important.

Externally applied magnetic fields are used on the Texas Experimental Tokamak (TEXT) to study the possibility of controlling the particle, impurity and heat fluxes at the plasma edge. Fields with toroidal mode number n = 2 or 3 and multiple poloidal mode numbers m (dominantly m = 7) are used, with a poloidally and toroidally averaged ratio of radial to toroidal field components 〈|br/Bo〉 ≅0. 1%. Calculations show that it is possible to produce mixed islands and stochastic regions at the plasma edge (r/a ≥ 0.8) without affecting the interior. The expected magnetic field structure is described and experimental evidence of the existence of this structure is presented. The edge electron temperature decreases with increasing 〈|br/Bo〉, while interior values are not significantly affected. The implied increase in edge electron thermal diffusivity is compared with theoretical expectations and is shown to agree with applicable theories to within a factor of three.

An experiment is performed to study the interaction of a high power microwave pulse with a magnetized dense plasma. A delayed emission appears after the complete termination of a pumping microwave pulse when the frequency of the pumping wave is equal to the harmonics of the cyclotron frequency and is near the plasma frequency. Three types of the delayed mission (type I, II, III) are observed, depending on the experimental conditions. Type I delayed emission is characterized by the frequency spectrum with sharp peaks at cyclotron harmonics (mfce m= 1, 2, ..), type II with broad peaks between (m+1/2) fce and (m + 1) fce and type III with broad peak between 0.8 fce and fce. Main properties of these phenomena are explained in terms of anomalous power absorption at cyclotron harmonic resonances and excitation of electron plasma waves by an anisotropic electron velocity distribution. Each type of the delayed emission can be triggered artificially by a sudden decrease of cold back ground plasma density and is suppressed by an increase of cold plasma. It can be also triggered by a pulsed electron beam.

It is proposed to expand the outermost flux surfaces in tokamaks to divert the heat flux emerging from the plasma core. The approach previously discussed in detail is to deform the poloidal field configuration for expansion of the boundary. A simpler approach is discussed here which does not require any coil system or power supply, an attractive feature for reactor application.

It is found that the magnetic fluctuation amplitudes abruptly decrease at L-H transitions and increase at H-ELM transitions (ELM is the edge localized mode). No other plasma parameter change precedes the magnetic signals, suggesting that the fluctuations are responsible for the increased edge transport in regimes with poor confinement. The fluctuation amplitude during L- and ELM-phases is largest at the points where the separatrix intersects the divertor plates. This is interpreted as evidence for the existence of line tied image currents in the scrape-off plasma.

In DIII-D H-mode plasmas, large amplitude modulation of steep edge electron pressure gradients are observed during the occurrences of giant edge-localized modes (ELM's). For a considerable range of plasma currents (1 to 2 MA), the experimental pressure gradient before a giant ELM is close to the theoretically predicted marginal stability limit for the first regime of the ideal ballooning mode. These results, together with observations showing that ELM's originate in the region of bad curvature, strongly suggest that ELM's are triggered by ballooning mode instabilities.

A new field-shaping coil concept is proposed which allows satisfactory plasma shaping and control and is compatible with some stringent reactor design requirements. The proposed field-shaping coil system has two coil layers. The inner coil layer, situated immediately outside the first wall, is very close to the plasma, thus significantly improving both MHD equilibrium and stability. The role of the outer coil layer,is to minimize the current in the inner coil layer, i.e. to reduce the power required to maintain the plasma configuration.

A series of initial resonant helical divertor (RHD) experiments have been carried out in ohmically and auxiliary heated JIPP T-IIU plasmas. Disruptive and MHD instabilities make the interpretation of the RHD results difficult but an apparent increase in the energy confinement time is observed when the helical magnetic perturbation is applied. This may be due to the suppression of MHD activity or to a reduction in the edge convective heat losses. Magnetic island effects have been observed on the floating potential of a Langmuir probe array and energy scrape-off layer widths have been measured with and without helical perturbations during ICRF operation. Basic pump limiter data is presented including ion temperatures and C4+ impurity profiles. Energy confinement times are reported in ohmically and NBI heated discharges.

From a condition for thermal stability of the plasma boundary in tokamaks, an upper limit on the density which is consistent with the experimental observations has been derived.

The improved energy confinement of the H-mode is invariably coupled to improved particle confinement, In other experiments, increased particle confinement resulted in impurity accumulation and severe power losses. In DIII-D, similar to the results of other experiments, H-mode confinement is associated with improved particle confinement. But in contrast to other devices, in most DIII-D H-mode plasmas there is no indication of impurity accumulation. In high current shots the impurity content of plasma during the H-mode is lower than the L-mode phase. Two factors most responsible for the low impurity content of DIII-D H-mode plasmas are the nearly flat or hollow electron density profiles and the presence of giant Edge Localized Modes (ELMs). Giant ELMs also cause large amplitude modulations in particle and power flow to the divertor target plates, resulting in large scale changes in the parameters of the the scrape-off layer plasma. Our analysis of temperature and density profiles during ELMs show that particle loss during ELMs is primarily due to perpendicular transport across the field lines.

The feasibility of a hot plasma injector using a divertor based on a magnetic neutral point discharge is studied experimentally. An intense plasma current magnetic neutral point discharge is induced along a circular magnetic neutral line of the divertor, and the plasma is heated and injected inward to the central region. The plasma diffusing along the magnetic field lines is measured by a time-of-flight method using a pair of electrostatic double probes, and it is shown that the average energy of the diffusing plasma ions is 200 eV and the total amount of injected particles is 6 × 1016 particles per shot. The possible use of the divertor injector in a fusion reactor is discussed.

It is proposed to expand the outermost flux surfaces in tokamaks to divert the heat flux emerging from the plasma core. The expanded flux surfaces provide a large volume for radiative cooling. The radiative power at the boundary is enhanced by the effects of plasma flow as well as by a volumetric factor, and the resultant edge cooling and reduced heat load on the limiter may significantly retard impurity generation. Furthermore, it seems to be compatible with reactor engineering requirements.

The authors' data indicate that the L-mode to H-mode transition in the DIII-D tokamak is associated with the sudden reduction in anomalous, fluctuation-connected transport across the outer midplane of the plasma. In addition to the reduction in edge density and magnetic fluctuations observed at the transition, the edge radial electric field becomes more negative after the transition. They have determined the scaling of the H-mode power threshold with various plasma parameters; the roughly linear increase with plasma density and toroidal field are particularly significant. Control of the ELM frequency and duration by adjusting neutral beam input power has allowed us to produce H-mode plasmas with constant impurity levels and durations up to 5 s. Energy confinement time in ohmic H-mode plasmas and in deuterium H-mode plasmas with deuterium beam injection can exceed saturated ohmic confinement times by at least a factor of two. Energy confinement times above 0.3 s have been achieved in these beam-heated plasmas with plasma currents in the range of 2.0 to 2.5 MA. Local transport studies have shown that electron and ion thermal diffusivities and angular momentum diffusivity are comparable in magnitude and all decrease with increasing plasma current.

The size of a magnetic island driven by an error field or a helical coil depends strongly upon the plasma response aiming at the maintenance of equilibrium. Sample calculations for tokamaks within the context of a linear theory show that the island size is greatly augmented when the tearing mode is close to a marginally stable state.

A phenomenological model has been developed to clarify the role of the boundary configuration in the heat transport of the H-mode regime. We assume that the dominant mechanism of heat loss at the edge of the plasma is convection and that the diffusion coefficient (D/sub edge/) at the edge of the plasma increases rapidly with plasma pressure, but drops to a low value when the temperature exceeds a certain threshold value. When particle refueling takes place without time delay, as in the case of a limiter discharge, the unfavorable temperature dependence of the D/sub edge/ prohibits even a modest rise of the edge temperature. In a divertor discharge, the particles lost from the closed surface are kept away from the edge region for a time comparable to or longer than the energy transport time in the edge region. Thus, rapid increase in the heat flux allows an excursion of the edge temperature to a higher value thereby reaching the threshold value of the H-transition.

Associated with the transition from ion root to electron root, an electron internal transport barrier (ITB) appears in the large helical device, when the heating power of electron cyclotron resonance heating exceeds the threshold power. The incremental thermal diffusivity of electron heat transport ${\ensuremath{\chi}}_{e}^{\mathrm{i}\mathrm{n}\mathrm{c}}$ in the ITB plasma is much lower than that in the plasma with the heating power below the threshold, and the thermal diffusivity ${\ensuremath{\chi}}_{e}$ decreases with increasing of heating power [$d{\ensuremath{\chi}}_{e}/d(P/{n}_{e})l0$] in helical ITB plasmas.

With use of a multimachine pedestal database, essential issues for each regime of ELM types are investigated. They include (i) understanding and prediction of pedestal pressure during type I ELMs, a reference operation mode of a future tokamak reactor; (ii) identification of the operation regime of type II ELMs, which have small ELM amplitude with good confinement characteristics; (iii) identification of the upper stability boundary of type III ELMs for access to the higher confinement regimes with type I or II ELMs; (iv) understanding the relation between core confinement and pedestal temperature in conjunction with the confinement degradation in high density discharges. Both scaling and model based approaches for expressing pedestal pressure are shown to roughly scale the experimental data well and could be used to make initial predictions for a future reactor. q and delta are identified as important parameters for obtaining the type II ELM regime. A theoretical model of type III ELMs is shown to reproduce the upper stability boundary reasonably well. It is shown that there exists some critical pedestal temperature below which the core confinement starts to degrade. It is also shown that it is possible to obtain improved pedestal conditions for good confinement in high density discharges by increasing the plasma triangularity.

Radial profiles of ion temperature and plasma flow are measured at the n/m = 1/1 magnetic island produced by external perturbation coils in the Large Helical Device. The sheared poloidal flows and sheared radial electric field are observed at the boundaries of the magnetic island, because the poloidal flow vanishes inside the static magnetic island. When the width of the magnetic island becomes large, the flow along the magnetic flux surface inside the magnetic island appears around the O point in the direction which reduces the shear of the poloidal flow at the boundary of the magnetic island.

The stabilizing effects of finite‐width poloidal limiters, toroidal limiters, and general mushroom limiters are examined for high‐beta finite resistivity tokamak plamas in free boundary. When the plasma pressure and resistivity are small, a poloidal limiter is effective in reducing the growth rate even with a small limiter size, while a toroidal limiter requires a large size for a comparable effect. As the plasma pressure or resistivity increases, a toroidal limiter becomes more effective in reducing the growth rate than a poloidal limiter of the same size. A small optimized mushroom limiter might have a stabilizing effect similar to a conducting shell.