Search

"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.

Neoclassical transport in the banana regime has been analysed with the DKES (Drift Kinetic Equation Solver) code for the Large Helical Device (LHD). It is found that in the 1/ν regime the diffusion coefficients change by one order of magnitude for various LHD configurations, depending on the structure of the helical magnetic ripple. The neoclassical transport calculated with the DKES code is quantitatively in good agreement with a multi-helicity theory formulated by Shaing and Hokin (1983). Incorporating the multi-helicity effect into the diffusion coefficient, the authors propose an interpolation formula for the diffusion coefficient in the 1/ν and ν regimes. When the ion temperature is increased at a fixed density of n = 1020 m-3, the ions undergo a transition from 1/ nu neoclassical transport to the nu regime when Ti > 3 keV, with a radial electric potential e comparable to the ion temperature (e/Ti ≈ 1). For the optimized configuration, the ion thermal diffusivity has a maximum value of ≈ 3.5 m2/s at a minor radius r/a ≈ 0.5. The bootstrap current has also been studied and the results have been compared with the theory. At the collisionless limit with a moderate radial electric potential of e/Ti ≈ 1, the DKES calculations evaluated for various LHD configurations support the theoretical formula given by Shaing and Callen (1983). For collision frequencies between the plateau regime and the banana regime, where the analytical theory is not applicable, the bootstrap current might become larger (by a factor of about two) than in the collisionless limit, depending on the radial electric field

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.

On DIII-D, periods of improved particle and energy confinement have been observed in low q, low BT divertor discharges with Ohmic heating alone. The Ohmic H-mode has characteristics similar to those of the H-mode produced by auxiliary heating. In the Ohmic H-mode the energy confinement time can have values near those predicted by Neo-ALCATOR scaling and a factor of two above those of non-H-mode Ohmic discharges at similar densities. These observations indicate that the H-mode is not limited to an improvement in confinement over that of the L-mode in auxiliary heating discharges.

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.

With Large Helical Device (LHD) and Wendelstein 7-X (W7-X), the development of helical devices is now taking a large step forward on the path to a steady-state fusion reactor. Important issues that need to be settled in these machines are particle flux and heat control and the impact of divertors on plasma performance in future continuously burning fusion plasmas. The divertor concepts that will initially be explored in these large machines were prepared in smaller-scale devices like Heliotron E, Compact Helical System (CHS), and Wendelstein 7-AS (W7-AS). While advanced divertor scenarios relevant for W7-X were already studied in W7-AS, other smaller-scale experiments like Heliotron-J, CHS, and National Compact Stellarator Experiment will be used for the further development of divertor concepts. The two divertor configurations that are being investigated are the helical and the island divertor, as well as the local island divertor, which was successfully demonstrated on CHS and just went into operation on LHD. At present, on its route to a fully closed helical divertor, LHD operates in an open helical divertor configuration. W7-X will be equipped right from the start with an actively cooled discrete island divertor that will allow quasi-continuous operation. The divertor design is very similar to the one explored on W7-AS. For sufficiently large island sizes and not too long field line connection lengths, this divertor gives access to a partially detached quasi-steady-state operating scenario in a newly found high-density H-mode operating regime, which benefits from high energy and low impurity confinement times, with edge radiation levels of up to 90% and sufficient neutral compression in the sub-divertor region (>10) for active pumping. The basic physics of the different divertor concepts and associated implementation problems, like asymmetries due to drifts, accessibility of essential operating scenarios, toroidal asymmetries due to symmetry breaking error fields, etc., are discussed.

The time-of-flight neutral particle measurement system1 has been installed on the 10-O port at the 2nd campaign. It has the capability of two-dimensional measurement with a larger viewing angle by using the pivot on the rectangular port, which is set inside of the 10-O mother port. The movable stage with the vertical scan mechanism rotates on the fan shaped stainless rails around the pivot. The sight lines of −2 to 32 degrees against the 10-O port surface makes the angle between the sight line and the magnetic axis of the standard configuration (3.6 m) of 95 to 35 in the 4th campaign. Therefore the analyzer can observe high-energy particles with these pitch angles. The higher scan speed of 0.17 degree/second than that in the last campaign (0.05 degree/s) can be obtained in this campaign. Precise spatial distribution of the high-energy particles can be expected because it enables the continuous scanning in the long discharge.

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.

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 local island divertor (LID) experiment was performed on the Compact Helical System (CHS) to demonstrate the principle of the LID. It was clearly demonstrated that the particle flow is controlled by adding a resonant perturbation field to the CHS magnetic configuration, and is guided to the back of an m/n = 1/1 island which is created by the perturbation field. The particles recycled there were pumped out with a pumping rate in the range from a few percent to about 10%. As a result, the line averaged core density was reduced by a factor of about 2 in comparison with non-LID discharges at the same gas puffing rate. In addition to the demonstration of these fundamental divertor functions, a modest improvement of energy confinement was observed, which could be attributed to the edge plasma control by the LID.

We have executed nonlinear magnetohydrodynamic simulations in a heliotron-type configuration with a large pressure gradient to reveal the nonlinear dynamics of a collapse phenomenon. The simulation results reproduce the qualitative characteristics of the experimental observation on the so-called core density collapse events in the Large Helical Device plasma with the super-dense core profile. A long-term nonlinear behaviour on the event, including the flushing mechanism of the core pressure, is clarified. The simulation result shows the linear growth of the ballooning-like resistive instability modes with the intermediate poloidal wavenumbers. The growth of the modes are eventually saturated, and the system experiences the energy relaxation in about 1 ms. It should be noted that the linear mode structures are localized in the edge region, whereas the core pressure rapidly falls as the system reaches the relaxed state. Such coexistence of the edge perturbation and the core collapse is consistent with the experimental observations. The lost pressure forms a wide base in the peripheral region. The core pressure is, on the other hand, remarkably reduced at a certain period, although it had well withstood the disturbance before it. The most salient feature on this period is the disordering of the magnetic field structure. The system keeps the nested-flux-surface structure well at the beginning, whereas part of them are abruptly lost in this period. Such a situation can induce a flattening of the pressure profile along the reconnected field lines. By checking the place where the plasma loss due to this mechanism occurs, such plasma outlets are found to be located mainly on the disordered region. Thus, one can conclude that the core collapse can be caused by the disturbance of the magnetic field.

The dynamics of the magnetic island structure in the plasma are investigated in plasmas with a wide range of beta and collisionality. The perturbed magnetic field is diagnosed by a toroidal array of flux loops installed in the vacuum vessel on the Large Helical Device (LHD). It is found that the magnetic island grows with beta at relatively low beta values. In contrast, when the beta exceeds a critical value, the sign of the perturbed magnetic field suddenly reverses and its strength saturates to the magnetic field perturbation required to cancel the external perturbation. This suggests spontaneous healing of the magnetic island.

High-ion-temperature (exceeding 5 keV) hydrogen plasmas have been successfully produced in the Large Helical Device [Iiyoshi et al., Nucl. Fusion 39, 1245 (1999); Motojima et al., Nucl. Fusion 47, S668 (2007)] with the ion heat confinement improvement in the core region. The experimental ion heat diffusivity at the core region is found to be almost independent of the ion temperature, T_i (even decreasing as T_i increases). The neoclassical (NC) ripple transport is suppressed by the ambipolar radial electric field, E_r (

In reduced recycling discharges using a local island divertor in the Large Helical Device [O. Motojima, H. Yamada, A. Komori et al., Phys. Plasmas 6, 1843 (1999)], a stable high-density plasma develops in the core region when a series of pellets is injected. A core region with ~5×10^20 m^?3 and temperature of 0.85 keV is maintained by an internal diffusion barrier (IDB). The density gradient at the IDB (r/a~0.6) is very high, and the particle confinement time in the core region is ~0.4 s. Because of the increase in the central pressure, a large Shafranov shift up to ~0.3 m is observed. The critical ingredients for IDB formation are a strongly pumped divertor to reduce edge recycling, and multiple pellet injection to ensure efficient central fueling. No serious magnetohydrodynamics activity and impurity accumulation have been observed so far in this improved discharge.

Characteristics of magnetic islands are investigated by magnetic diagnostics in the L arge Helical Device (LHD). The structure of the magnetic island with m/n = 1/ 1( where,m and n are poloidal and toroidal mode number, respectively) can be estimated from the perturbed magnetic field appearing when a magnetic island changes. To measure the toroidal profile of the perturbed magnetic field δb1 originating from the plasma, a toroidal array of magnetic flux loops is set up in the LHD. The toroidal profile of δb1 is then spatially Fourier decomposed to determine the amplitude of the n = 1c omponent, δb n=1 1 and its phase, φn=1 whic hc orrespond the change of the island width and the toroidal position of the X-point of the island, respectively. Therefore, the information about the magnetic island structure can be obtained from δb n=1 1 and φn=1 .I ncase the island width becomes larger than the seed island, measurements show that δb n=1 1 is non-zero and φn=1 is temporally constant. A non-zero δb n=1 1 can also be observed when the island width becomes smaller than the seed island. In this case, the angle φn=1 shifts by about π[rad] compared with the increasing case and the δb n=1 1 is limited to a certain value whic hc orresponding to the magnetic field suppressing the seed island.

Achieving steady-state plasma operation at high plasma temperatures is one of the important goals of worldwide magnetic fusion research. A high temperature of approximately 2 keV, and steady-state plasma-sustainment operation of the Large Helical Device (LHD) [O. Motojima, K. Akaishi, H. Chikaraishi et al., Nucl. Fusion 40, 599 (2000)] is reported. High-temperature plasmas were created and maintained for more than 30 min with a world record injected heating power of 1.3 GJ. The three-dimensional heat-deposition profile of the LHD helical divertor was modified and during long-pulse discharges it effectively dispersed the heat load using a magnetic-axis swing technique developed at the LHD. A sweep of only 3 cm of the major radius of the magnetic axis position (less than 1% of the major radius of the LHD) was enough to disperse the divertor heat load. The modification of the heat-load profile was explained well by field-line tracing. The steady-state plasma was heated and sustained mainly by hydrogen minority ion heating using ion cyclotron range of frequencies. The operation lasted until a sudden increase of radiation loss occurred, presumably because of wall metal flakes dropping into the plasma. The sustained line-averaged electron density was approximately 0.7?0.8×10^19 m^?3. The average input power was 680 kW, and the plasma duration was 31 min 45 s. This successful long operation shows that the heliotron configuration has a high potential as a steady-state fusion reactor.

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.

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.

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.

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.

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.

The concept of merging tokamak plasmas is proposed. It has potential applications for solving the problems of pseudo steady-state tokamak operation, refuelling, impurity removal and plasma heating. The currents in the two merging tokamak plasmas are in the same direction so that the columns attract each other. A model for the merging process is given.

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.

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.

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.

Discharges with improved energy confinement are produced in Doublet III when the major recycling location is shifted away from the main plasma by forming an expanded boundary (XB) divertor configuration with a separatrix to limiter separation Δ > 1.5 cm. Monte Carlo calculations indicate that over a wide range in density the divertor plasma effectively limits the penetration of recycled neutrals to the main plasma. This results in large particle recycling rates at the divertor plates and a reduction of particle recycling near the main plasma. This divertor function apparently plays an important role in obtaining improved energy confinement. Particle containment, as inferred from variations in Dα emission near the main plasma, is found to decrease with n e and BT, and increases rapidly (faster than quadratic) with IP whereas energy confinement is relatively insensitive to variations in n e and BT and increases linearly with IP. This improved energy confinement with the XB divertor configuration is suggestive of particle confinement playing an important role. However, in view of their rather different parametric dependences, the correlation between τP and τE, if any, remains unclear.

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.

The ergodic magnetic limiter coils on TEXT have been reconfigured to produce the primary helical perturbation resonance at m = 7 n = 3 . The experiments continue to demonstrate that the weak resonant perturbations modify the edge conditions in keeping with model predictions. We observe a reduction in the intrinsic impurity levels accompanying the helical current pulse, presumably the result of a reduction in the electron temperature in the edge. Heat follows the perturbed field lines to the limiter, generating heat load patterns which reflect the geometry of a magnetic island-limiter intersection. A strong spatial modulation of the electron density in the scrape-off-layer also reflects the helical mode structure.

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.

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.