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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)

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.

Density and power scan experiments have been carried out at various magnetic field strengths on the Large Helical Device (LHD), to investigate the temperature and magnetic field dependences of the thermal diffusivity. In the moderate density regime, the thermal diffusivity shows gyro-Bohm-like parameter dependences. In the high-density and low-temperature regime, on the other hand, the thermal diffusivity increases with the square root of the local electron temperature and decreases with a 1.2 power of the magnetic field strength. The turning point temperature where the weak temperature dependence changes to the gyro-Bohm type is also found to increase with a 0.8 power of the magnetic field strength. Based on the experimental results, dimensionally correct scalings for the energy confinement time, τE, and the plasma stored energy for the high-density LHD plasmas have been derived. Compared with the gyro-Bohm model (τE ∝ (n/P)3/5, where n and P denote the density and the heating power, respectively) and the international stellarator scaling 1995 (τE ∝ n0.51P−0.59), our scaling has a weaker positive dependence on the density together with the mitigated power degradation as τE ∝ (n/P)1/3, owing to the weak temperature dependence of the thermal diffusivity.

A local island divertor (LID) experiment has begun on LHD, with the aims of controlling edge recycling and improving the plasma confinement. The fundamental divertor functions of the LID have been demonstrated in the recent experiments. From the particle flux profile measurements on the LID head it was found that the particles diffusing out from the core region are well guided along the island separatrix to the LID head. Owing to the closed configuration around the LID head, evidence of the high efficient pumping was observed, together with a strong capacity to screen impurities. The first results of edge modeling using the EMC3-EIRENE code are also presented.

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.

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.

In LHD discharges, the NBI heated plasmas are terminated in two ways: (a) thermal decay (TD) after the termination of NBI and (b) radiative collapse (RC) during the NBI heating. The basic characteristics of the TD and RC discharges are compared. It is found that the decay and collapse of the plasma are mainly governed by the heating power and the plasma density. The critical density c for the collapse of RC plasmas is similar to the scaling laws obtained in other helical devices, i.e. c∝(PB/V)0.5, where P, B and V denote heating power, magnetic field and plasma volume, respectively. Moreover, measurements using multichannel bolometric diagnostics indicate that the total radiation profiles in TD and RC plasmas are usually inboard-outboard symmetric and asymmetric, respectively, at the end of the discharge. In RC discharges, the total radiation profile develops in several phases. Before the onset of the thermal instability (TI), the radiation profile is rather symmetric, while after that, the radiation profile evolves from being symmetric in the initial period towards being asymmetric eventually with high radiation on the inboard side. Corresponding variations are shown in the time evolutions of the density and temperature profiles, and a substantial contraction of the plasma column is observed immediately after TI onset. The spatial and temporal coincidence of the asymmetries in the radiation, density and temperature is similar to that observed with multifaceted asymmetric radiation from the edge (MARFE) in tokamaks. But, unlike MARFEs, the asymmetric radiation (AR) in LHD is rather transient since it appears just before the end of RC discharges. The underlying cause for the development of radiation asymmetry was investigated and compared with existing instability models. The result suggests that the high inboard radiation is a manifestation of an enhanced local thermal instability, and the AR results from asymmetric developments of TI on the inboard-outboard sides during the final stage of RC discharges.

The energy confinement time of net current-free plasmas in the Large Helical Device (LHD) is comparable to ELMy H-mode with q = 4.5 on the scale of IPB98(y,2). Broad temperature with high edge pressure, which is of robust nature in LHD, recalls the pedestal structure in tokamak H-mode. Basic parameter dependence is, however, weak gyro-Bohm in both edge and core regions. This suggests that one physical model could explain the observed profile. The edge temperature is proportional to the square root of heating power and inversely proportional to the square root of edge density. This means that edge pressure and its gradient are not limited and can be increased by heating power.

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.

During the first two years of the LHD experiment the following results have been achieved: (i) higher Te (Te(0) = 4.4 keV at ne = 5.3 × 1018 m-3 and Pabs = 1.8 MW); (ii) higher confinement (τE = 0.3 s, Te(0) = 1.1 keV at ne = 6.5 × 1019 m-3 and Pabs = 2.0 MW); (iii) higher stored energy, Wpdia = 880 kJ at B = 2.75 T. High performance plasmas have been realized in the inward shifted magnetic axis configuration (R = 3.6 m) where helical symmetry is recovered and the particle orbit properties are improved by a trade-off of MHD stability properties due to the appearance of a magnetic hill. Energy confinement was systematically higher than that predicted by the International Stellarator Scaling 95 by up to a factor of 1.6 and was comparable with the ELMy H mode confinement capability in tokamaks. This confinement improvement is attributed to configuration control (inward shift of the magnetic axis) and to the formation of a high edge temperature. The average beta value achieved reached 2.4% at B = 1.3 T, the highest beta value ever obtained in a helical device, and so far no degradation of confinement by MHD phenomena has been observed. The inward shifted configuration has also led to successful ICRF minority ion heating. ICRF powers up to 1.3 MW were reliably injected into the plasma without significant impurity contamination, and a plasma with a stored energy of 200 kJ was sustained for 5 s by ICRF alone. As another important result, long pulse discharges of more than 1 min were successfully achieved separately with an NBI heating of 0.5 MW and with an ICRF heating of 0.85 MW.

Semiconductor detectors based on natural diamonds have been installed on the Large Helical Device (LHD) heliotron to measure the energy distribution of charge exchange fast neutral particles from different viewing angles. Advantages of a natural diamond detector (NDD) are (1) very compact size, (2) relatively easy handling, and (3) high energy resolution. Although NDDs are sensitive to visible light, vacuum ultraviolet, and soft x rays, unfavorable pulses produced by such radiation were greatly reduced by choosing an appropriate stainless steel shield in this experiment. In LHD, the time-resolved energy distribution of counter-going beam ions and ion cyclotron range of frequency-produced energetic ions have been successfully obtained by means of an NDD. The performance of NDDs as a neutral particle analyzer and its good suitability to LHD plasmas were demonstrated throughout this work.

Irregular magnetic fields resonate with the rational surface of the magnetic confinement systems, form magnetic islands and ergodic layers, and destruct the plasma confinement. To avoid this confinement destruction the requirement of an accuracy of 10 −4 in the magnetic field is adopted as the magnetic-accuracy design criterion for the LHD machine. Following this criterion the width of the undesirable magnetic island is kept less than one tenth of the plasma radius. The irregular magnetic field from the superconducting (SC) helical and poloidal coils is produced by winding irregularity, installing irregularity, cooling-down deformations and electromagnetic deformations. The local irregularities such as feeders, layer connections, adjacent-conductor connections of the coils also produce an error field. The eddy currents on the supporting shell structure of SC coils, the cryostat, etc. are also evaluated. All irregular effects are analyzed using Fourier decomposition and field mapping methods for the LHD design, and it is confirmed that the present design of the superconducting coil system satisfies the design criterion for these field irregularities.

A simple method is proposed to suppress the breaking of magnetic surfaces which occurs in l = 2 heliotron/ torsatron equilibria. With this method, fairly high beta equilibria with clearly nested magnetic surfaces can be realized without changing other physical properties such as the rotational transform profile and the well depth.

In LHD discharges a significant enhancement of the global energy confinement has been achieved for the first time in a helical device with an edge thermal barrier, which exhibits a sharp gradient at the edge of the temperature profile. Key features associated with the barrier are quite different from those seen in tokamaks: (i) almost no change in particle (including impurity) transport, (ii) a gradual formation of the barrier, (iii) a very high ratio of the edge temperature to the average temperature, and (iv) no edge relaxation phenomenon. These features are very attractive in applying the thermal barrier to future reactor grade devices.

A simple model of energy confinement in tokamaks is proposed. In this model, the temperature and current profiles are determined by tearing mode stability considerations. As a result, the global energy confinement becomes strongly dependent on the boundary energy transport. A higher edge temperature ensures a higher core temperature and hence a higher global energy confinement.

The goal of the Compact Ignition Tokamak (CIT)d program is to provide a cost-effective route to the production of a burning deuterium-tritium plasma, so that alpha-particle effects may be studied. A key issue to be studied in the CIT is whether alpha power behaves like other power sources in affecting tokamak plasma confinement. The program is managed by the Princeton Physics Laboratory and includes broad community involvement. Guidelines for the preliminary design effort have been provided by the Ignition Technical Oversight Committee in discussion with the tokamak community. The reference design is a tokamak with a high filed (10 T), high current (10 MA), poloidal divertor, and liquid-nitrogen-cooled coils. It is a small, high-power-density device of the type proposed by Bruno Coppi (MIT). It has a major radius of 1.23 m, a minor radius of 0.43 m, and plasma elipticity of 1.8. This paper reviews the aims of the program and the basis for the physics guidelines. The role of the CIT in the longer-term tokamak program is briefly discussed. 23 refs., 9 figs., 1 tab.

The first experimental test of the ergodic magnetic limiter concept is being conducted on the TEXT device. Modular coils have been installed on TEXT allowing the external generation of resonant helical field perturbations. Preliminary results are presented.

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 instrumented scoop limiter probe is being operated on TEXT to test the concept of limiter cooling and improved particle removal efficiencies using an externally-applied resonant magnetic field perturbation (the resonant helical divertor concept). Cooling of the limiter face has been demonstrated for limiter positions ranging from r L = 29.0 cm inward to r L = 25.5 cm (the Text primary poloidal hoop limiter radius r a = 27.0 cm). Pressure rises in the limiter throat of approximatel 40% are observed under optimized conditions. Interchangable limiter heads with thicknesses of 1.0 cm and 0.3 cm have been used to examine particle ducting into the scoop aperture. Experimental results are discussed along with observations of the limiter floating potential, H α recycling emissions, pressure measurements, and edge density and temperature measurements.

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.

Application of a resonant helical magnetic field to the tokamak boundary may provide substantial improvement in performance. This configuration could establish the radiative boundary for handling high heat flux and minimizing impurity generation. Also described is the potential for improved confinement with properly selected resonant perturbations.

The DIII-D high and low beta stability results have established the basic feasibility of the divertor and H-mode configurations up to elongation 2.0 for next generation tokamaks. The 6.8% beta T achieved has already exceeded projected operating requirements of next generation devices. beta T > 6% has been sustained for 800 ms. Stability calculations and patterns of MHD mode behavior suggest a central expanding zone of ballooning instabilities leads ultimately to unstable m/n=2/1 modes which cause beta collapse or disruption. The pressure gradient at the plasma edge just reaches the first regime ballooning limit prior to ELMs. ECRH has proven effective for generating H-mode, sawtooth suppression, and ELM suppression.

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.

An idea is suggested which leads to a significant reduction in the required number of toroidal field coils and to a dramatic improvement in access for tokamak maintenance and diagnostics. It is found that a two or four coil system may be practical. The basic concept is presently being used for designing RST (Radio Frequency Driven, Steady-State Tokamak) of General Atomic Company.

Using a neutral-beam injection power of 3.4 M W, volume-averaged toroidal betas of up to βT = 4.5% have been obtained in low-toroidal-field, low-qψ, vertically elongated discharges in the Doublet III tokamak. This level of βT is above the minimum level required for a tokamak reactor, thus demonstrating that reactor level values of βT are possible in a tokamak device. The observed enhancement of βT with vertical elongation lends confidence in the design of future devices which rely on vertical elongation.

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 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.

The Ergodic Magnetic Layer Experiment on ohmically heated Text demonstrates that a small resonant helical field (m/n = 7/2 or 7/3) with fractional amplitude r/BT of about 10−3 creates a stable ergodic magnetic layer and substantially modifies the boundary heat flow. Field line tracings accurately map the observed perturbed limiter heat load patterns. The level of the intrinsic impurities is reduced, consistent with a lower edge electron temperature, as measured by the electron cyclotron emission detector.

資料番号: SA1001635000

In SIII-D, good H-mode confinement has been obtained with up to 6 MW of neutral-beam injection (NBI). In this report we summarize the essential features of divertor operation for DIII-D H-mode plasma. Briefly, our measurements show that: Large edge density and temperarture gradients are obtained with scale lengths of a few cm at the midplane; In the steady state, approximately 10-20% of the input power is deposited on the divertor plates in a region that is peaked on the separatrix and extends radially over )approx equal) 2-4 cm at the outboard intercept. More power (2:1) reaches the outboard intercept than the inboard; Low electron temperature (less than or equal to20eV) and high density (n/sub e//sup div/ )approx equal) n/sub e//sup main/ greater than or equal to 3-10 )times)10/sup 19/ m/sup -3/ at the divertor plates indicate that DIII-D obtains a high-recycling divertor during H-mode. DEGAS neutral transport modeling is consistent with these results; The duration of good H-mode confinement is limited by Edge Localized Modes (ELMs) rather than increasing impurity radiation; The dominant impurities, C and Ni are concentrated in the edge plasma (ra greater than or equal to 0.5) for I/sub p/ greater than or equal to 1 MA;more » and Global impurity concentrations fall markedly with increasing plasma current. The remainder of this report contains a brief description of the DIII-D tokamak and the relavant diagnostics, a summary of the divertor plasma parameters obtained during the H-mode and the heat loads on the diverotr target plates. The results of measurements of the central radiation and impurity content using data from an EUV spectrometer and a 21 channel bolometer array are discussed, and finally a discussion of the implication of these results for the CIT and ITER tokamak designs is given. 8 refs., 8 figs« less

The heating systems for the new large helical device which has been designed as a joint effort among Japanese universities are reported. The conceptional design of the heating devices and basic hardware structures are discussed. ECH heating system (112 GH z ) with maximum power of around 10MW is proposed to produce and heat the net current-less plasma. For the further heatings, NBI (20 MW, 100 kV, H 0 ) and ICRF (9 MW, 30–90 MH z ) heating systems have been considered. High energy particle loss, which is a fairly serious problem in helical configurations (stellarator/heliotron), have been analyzed numerically. Monte-Carlo calculation shows that, for tangential injection, the beam energy over 100 kV is necessary to obtain the thermalization efficiency of above 60% inside the half radius. On the fast-wave (ICRF) heating, theoretical analysis of the wave field and velocity distribution function, which includes the loss cone effect, has been developed. In the case of 3 He minority heating, the orbit loss problem becomes small than in the case of proton minority heating. From these theoretical analysis and hardware considerations, the required thermalized power of around 20 MW can be obtained by ECH, NBI and ICRF heatings.

New stellarator experiments have been launched since the last compilation of the international stellarator database in 1995. Parameter regimes have been extended by Large Helical Device (LHD), and a variety of improved modes have been found since then. The revision of the international stellarator database has been initiated, driven by these emerging interests and by the requirements for a reactor assessment. Some provisional issues are discussed. An understanding of configurational effects is a prerequisite to the derivation of a unified scaling. Differences in magnetic geometry are influential in characterizing energy confinement. The results from the magnetic axis and elongation scans in LHD are highlighted. Comparison with tokamak confinement is also addressed. The revision of the database is in progress, and this paper is an interim report.

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.

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.