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A new 320 GHz solid-state source interferometer is installed in the Heliotron J helical device to explore the physics of high-density plasmas (ne > 2–3 × 1019 m−3, typically) realized with advanced fueling techniques. This interferometry system is of the Michelson type and is based on the heterodyne principle, with two independent solid-state sources that can deliver an output power of up to 50 mW. A high time resolution measurement of

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The Pa α line (1875.13 nm) in the near-infrared (NIR) region was evaluated to apply Stark broadening of the line spectrum to the electron density measurement of the small-pellet ablation cloud in Heliotron J, a medium-sized helical-axis heliotron device. Pa α is three-to-four times broader than the visible H β line (486.13 nm) for the same electron density. Using a portable NIR spectrometer, preliminary proof-of-concept experiments determined the marginal density, below which the broadening was undetectable. The lower detection density limit can be decreased using a narrower entrance slit or a denser grating.

We have investigated the effect of uniaxial and biaxial strain on the binding energy of Ge 2p, Ge 3d core level and the top of the valence band edge using x-ray photoelectron spectroscopy. The strain was measured by using a Raman spectrometer. From these results, we succeeded in the observation of the effect of uniaxial and biaxial strain on the binding energy and the top of the valence band edge.

Volume recombination of C4+ and e? into C3+ is observed for the first time in detached divertor plasmas with an X-point MARFE. The recombination zone is located around the X-point, and the electron temperature and density are evaluated to be 6.3?eV and 7.8 ? 1020?m?3, respectively. In this zone, the volume recombination flux is larger by two orders of magnitude than the ionization flux of C3+. However, the radiation power due to the recombination process is only 2% of the total radiation power, measured by a bolometer. In contrast, the radiation power due to the excitation process from the ground state of C3+ by electron collision dominates the total radiation power.

An imaging bolometer has been realized for the first time on a tokamak (JT-60U). The imaging bolometer utilizes a 7 cm × 9 cm × 2.5 μm gold foil and an Omega/Micron infrared (IR) camera by Indigo/FLIR (128 × 164 pixels, 70 mK, 30 fps). The imaging bolometer has an array of 12 (toroidal) × 16 (poloidal) (192 in total) channels, a noise equivalent power density of greater than 350 μW/cm2 and a frame rate of 30 fps. The field of view of the imaging bolometer is semi-tangential covering the divertor for more than 100° toroidally in addition to the entire poloidal cross-section. The data clearly shows examples of the radiation changing from strong divertor radiation to core dominated radiation in agreement with the resistive bolometer data.

Modification for long pulse operation up to 65 s with a neutral beam heating power of 12 MW (30 s) has provided an opportunity for studies on plasma-wall interaction on a time scale of tens of seconds. During a long pulse, ELMy H-mode discharge, the net wall-pumping rate, evaluated by a particle balance analysis, gradually decreases, and subsequently, becomes zero in ~20 s after the neutral beam injection starts. This situation is interpreted as 'wall saturation'. Wall saturation is observed in repetitive long pulse discharges. Although the outgas due to the increase in the divertor plate temperature and increase in the static retention with pulses also contribute to the particle balance, the dynamic retention process dominates the particle balance and determines the wall saturation. Under the condition of wall saturation, the plasma density continuously increases due to insufficient divertor-pumping rate. This density rise results in detachment of the outer divertor plasma, followed by an X-point MARFE. Results from the calculation of a 2-dimensional plasma transport code indicate that higher divertor-pumping efficiency by a factor of 1.5 is required to avoid the detachment. Note that no negative effects on impurity behaviour such as carbon bloom, or significant dilution of the core plasma has been observed even in the case of the total input energy reaching 350 MJ.

Variation of particle absorption at the first wall has been investigated in long-pulse (~30?s) ELMy H-mode discharges on JT-60U. Quantitative analysis of particle balance indicated that particle inventory at the first wall was globally saturated with a time scale of 10?15?s after several long-pulse discharges. To understand mechanisms of the global wall saturation, distribution of a local wall saturation time on the first wall was calculated from the ion and neutral particle fluxes to the first wall evaluated using a Monte-Carlo neutral particle transport code. The local wall saturation time was estimated to be shorter than 1?s at the divertor plates and the divertor dome, ~10 s at the lower half of the baffle plates and ~100?s at the main chamber wall, respectively. This result suggested that the divertor plates, the divertor dome and the lower half of the baffle plates were saturated in a single discharge. On the other hand, the main chamber wall was not saturated in a single discharge. Based on the above result, a model of the global wall saturation was proposed, where dynamic and static inventory regions are defined depending on the wall temperature.

Compatibility of advanced tokamak plasmas with high density and high radiation loss has been investigated in both reversed shear (RS) plasmas and high βp H-mode plasmas with a weak positive shear on JT-60U. In the RS plasmas, the operating regime is extended to high density above the Greenwald density (nGW) with high confinement (HHy2 > 1) and high radiation loss fraction (frad > 0.9) by tailoring the internal transport barriers (ITBs). With a small plasma-wall gap, the radiation loss in the main plasma (inside the magnetic separatrix) reaches 80% of the heating power due to metal impurity accumulation. However, high confinement of HHy2 = 1.2 is sustained even with such a large radiation loss in the main plasma. By neon seeding, the divertor radiation loss is enhanced from 20% to 40% of the total radiation loss. In the high βp H-mode plasmas, high confinement (HHy2 = 0.96) is maintained at high density ( ) with high radiation loss fraction (frad ~ 1) by utilizing high-field-side pellets and argon (Ar) injection. The high is attributed to the formation of strong density ITB. Strong core-edge parameter linkage for confinement improvement is observed, where the pedestal pressure and the core plasma confinement increase together. The measured radiation profile including contributions from all impurities in the main plasma is peaked, and the central radiation is ascribed to the contribution from Ar accumulated inside the ITB. Impurity transport analyses indicate that the Ar density profile, twice as peaked as the electron density profile, which is the same level as that observed in the high βp H-mode plasma, can yield an acceptable radiation profile even with a peaked density profile in a fusion reactor.

JT-60U's diagnostic systems have been using CAMAC systems for their control and data acquisition. But the transient recorders used to store the diagnostic data are rather old and lack in memory expansion. Therefore, for a long pulse discharge of more than 15 s, data sampling rate must be reduced to cover the whole discharge period. The acquired data has, however, low time resolution and is sometimes insufficient for detailed analysis. To solve this problem, the transient recorders should be replaced with those having larger memory, but without modification to the existing data processing infrastructure. For this purpose, we have developed a new data acquisition system with SPARC CPU on VMEbus and the associated software, to retain the functionality of CAMAC control. We have applied this system to the existing bolometric diagnostics. In this paper, detail of design will be described.

Fusion performance of reversed shear discharges with an {ital L}-mode edge has been significantly improved in a thermonuclear dominant regime with up to 2.8 MA of plasma current in the JT-60U tokamak. The core plasma energy is efficiently confined due to the existence of persistent internal transport barriers formed for both ions and electrons at a large minor radius of r/a{approximately}0.7 near the boundary of the reversed shear region. In an assumed deuterium-tritium fuel, the peak fusion amplification factor defined for transient conditions involving the dW/dt term would be in excess of unity. {copyright} {ital 1997} {ital The American Physical Society}

DIII-D has successfully operated with an all-graphite first wall, including the first observations of VH-mode without boronization. A major goal of this, and other recent upgrades, was to control impurity influxes and hydrogenic fueling. Graphite tiles were carefully preconditioned, first by ex situ preparation and then by baking and helium glow conditioning. No deuterium or hydrogen was used until tokamak operation commenced. With the all graphite wall, both impurity and deuterium influxes during tokamak discharges were lower than previous boronized discharges; central nickel impurity line radiation, NiXXV and NiXXVI, was an order of magnitude lower than previous discharges during the ELM free beam heated phase. The effect of reduced particle fueling on plasma performance, particularly H- and VH-mode discharges, will be presented.

A regime of improved H-mode energy confinement, VH-mode, is obtained in the DIII-D tokamak with adequate vessel conditioning. The improved confinement in VH-mode is consistent with the extension of the region of high E*B velocity shear turbulence suppression zone further in from the plasma boundary. The energy confinement enhancement in VH-mode can be limited by ELMs, localized momentum transfer events, or operation at high heating power or low q. Energy confinement enhancement improves with increasing triangularity of the plasma cross section and is independent of elongation. The termination of VH-mode is associated with an edge localized kink mode which is destabilized by both the large edge pressure gradient and edge current density.

Second-harmonic electron cyclotron current drive (ECCD) experiments have been performed in the stellarator/heliotron (S/H) device, Heliotron J. A focused Gaussian beam is injected with the parallel refractive index, N ∥, ranging from −0.05 to 0.6. The EC driven current is estimated by excluding the bootstrap current from the total current. The experimental results show that the EC driven current is determined by the local magnetic field structure where the EC power is deposited. The maximum EC driven current is attained around N ∥ = 0.5 when the EC power is deposited at nearly the top of the magnetic ripple. A large increase in electron cyclotron emission (ECE) signals is observed when the EC current is driven, indicating an important role for high-energy electrons in the ECCD. The experimental results for N ∥ and B dependence agree with a ray-tracing simulation using a parallel momentum conservation model in which the trapped particle effect is included.

Recent lower hybrid current drive (LHCD), and heating (LHH) experiments on JT-60 are reported. The current drive product of neRpIRF approximately 12.5*1019 m-2 MA was achieved at the LH power of approximately 4.5 MW, and the CD efficiency, the energy confinement, the global power balance and the heat load on divertor plates were investigated in high power LHCD plasmas. Nearly steady state H-mode discharges were found during LHCD with two different frequency injections. Sawtooth suppression in NB heated plasmas by LHCD have shown an improvement in confinement near the plasma center. Parametric instabilities in LH heating experiments were significantly reduced by increasing the plasma current, and the stored energy increased linearly with heating power of up to approximately 9 MW at ne approximately 7*1019 m-3 and Ip=2.75 MA. Parametric instabilities near the plasma edge in the ion heating regime were also reduced in peaked density plasmas produced by pellet injection and LH waves increased the central plasma pressure at ne(0) > 1.4*1020 m-3.

In this paper we present the design of an upgraded version of the imaging bolometer on JT-60U. The 2.5 micron gold foil is replaced with a 5 micron Ta foil for improved rigidity. The vacuum window will be enlarged from 58 mm (ZnSe) to 100 mm (sapphire). The enlarged window and a 3.8 m IR periscope using CaF2 optics and an aluminum mirror will improve the optical throughput. The periscope will permit a more advanced IR camera (FLIR/Indigo Phoenix-mid) (3-5 microns, 256 times 360 pixels, 345 Hz, 25 mK) to be mounted at a location of lower neutron flux and lower magnetic field in a shield box consisting of 2 cm of soft iron (against magnetic field), 2 cm of lead (against secondary gammas), and 10 cm of polyethylene (against neutrons). Through the combination of improved optical throughput using the periscope and the improved sensitivity of the IR camera we expect an improvement in the sensitivity of the imaging bolometer by at least a factor of 50 compared to the previous version. This improvement in sensitivity is used to increase the number of channels from 12 (toroidal) times 16 (poloidal) to 24 times 40 by reducing the aperture size from 5 mm times 5 mm to 2 mm times 2.5 mm, respectively. This imaging bolometer represents a reactor relevant prototype for those planned for KSTAR, LHD, JT-60SA and proposed for ITER.

The calibration technique of the imaging bolometer foil gives confidence in the measured values of the plasma radiation power. In this paper we improve the accuracy of the calibration of the imaging bolometer foil by using an IR camera (FLIR/SC500) (60 Hz, 320 x 240 pixels, 7.5-13 mum) in a calibration laboratory (low noise environment) with a close-up lens, which provides a spatial resolution of ~0.09 mm which is ~7 times better than the previous in situ calibration using a different IR camera (Indigo/Omega) (30 Hz, 160 x 120 pixels, 7.5-13.5 mum). In addition the foil material is changed for the development of the imaging bolometer. Tantalum (Ta) is offered for improved sensitivity, foil strength, and lower neutron cross-section compared to the previously used gold foil of the imaging bolometer for the JT-60U tokamak. In this paper we report the local foil properties (the thermal conductivity (constant), and the thickness) of a single graphite-coated gold foil (a nominal thickness of 2.5 mum) for LHD compared with those of a single graphite-coated tantalum foil (a nominal thickness of 5 mum) for JT-60U by calibration with a He-Ne laser.

Progress, since the ITER Physics Basis publication (ITER Physics Basis Editors et al 1999 Nucl. Fusion 39 2137-2664), in understanding the processes that will determine the properties of the plasma edge and its interaction with material elements in ITER is described. Experimental areas where significant progress has taken place are energy transport in the scrape-off layer (SOL) in particular of the anomalous transport scaling, particle transport in the SOL that plays a major role in the interaction of diverted plasmas with the main-chamber material elements, edge localized mode (ELM) energy deposition on material elements and the transport mechanism for the ELM energy from the main plasma to the plasma facing components, the physics of plasma detachment and neutral dynamics including the edge density profile structure and the control of plasma particle content and He removal, the erosion of low- and high-Z materials in fusion devices, their transport to the core plasma and their migration at the plasma edge including the formation of mixed materials, the processes determining the size and location of the retention of tritium in fusion devices and methods to remove it and the processes determining the efficiency of the various fuelling methods as well as their development towards the ITER requirements. This experimental progress has been accompanied by the development of modelling tools for the physical processes at the edge plasma and plasma-materials interaction and the further validation of these models by comparing their predictions with the new experimental results. Progress in the modelling development and validation has been mostly concentrated in the following areas: refinement in the predictions for ITER with plasma edge modelling codes by inclusion of detailed geometrical features of the divertor and the introduction of physical effects, which can play a major role in determining the divertor parameters at the divertor for ITER conditions such as hydrogen radiation transport and neutral-neutral collisions, modelling of the ion orbits at the plasma edge, which can play a role in determining power deposition at the divertor target, models for plasma-materials and plasma dynamics interaction during ELMs and disruptions, models for the transport of impurities at the plasma edge to describe the core contamination by impurities and the migration of eroded materials at the edge plasma and its associated tritium retention and models for the turbulent processes that determine the anomalous transport of energy and particles across the SOL. The implications for the expected performance of the reference regimes in ITER, the operation of the ITER device and the lifetime of the plasma facing materials are discussed.

The effect of magnetic field configurations on plasma flow velocity was investigated by measuring the parallel flow velocity (v∥) using a charge-exchange recombination spectroscopy during neutral beam injection in three different toroidal mirror configurations of Heliotron J: high, standard and reversed mirror configurations. The magnetic ripple strengths, γ, for these mirror configurations were γ = 0.073 m−1, 0.031 m−1 and 0.027 m−1, respectively, at the normalized averaged minor radius ρ = 0.07. The magnetic ripple strength is defined as γ = {〈(∂B/∂l)2/B2〉}1/2, where 〈...〉 is the flux surface averaged value and l is the length along the magnetic field line. At ρ = 0.07, the parallel flow velocity in the high mirror configuration (v∥ ~ 4 km s−1) was 2–3 times smaller than those in the standard and reversed mirror configurations (v∥ ~ 10–12 km s−1). An anticipated interpretation is that the difference in the neoclassical damping force contributes to the difference in v∥ among the three mirror configurations.

This paper describes the measurement of the density fluctuation using beamemission spectroscopy in Heliotron J, having the non-symmetrical helical-magnetic-axis configuration. In order to optimize the sightlines, the numerical calculations are carried out to estimate the spatial resolution and the observation location. When a tangential neutral beam is used as diagnostic one, suitable sightlines from the newly installed diagnostic port are selected whose spatial resolution Δρ is less than ±0.07 over the entire plasma region. Modification of the interference filter and the detection systems enables us to measure the radial profile of the density fluctuation. Each of the three coherent modes due to the fast-ion-driven magnetohydrodynamic instabilities has different radial structure of the density fluctuation., Contributed paper, published as part of the Proceedings of the 19th Topical Conference on High Temperature Plasma Diagnostics, Monterey, California, May 2012.

This paper describes the application of the beam emission spectroscopy (BES) to Heliotron J, having the nonsymmetrical helical-magnetic-axis configuration. The spectral and spatial profile of the beam emission has been estimated by the numerical calculation taking the collisional excitation processes between plasmas (electrons/ions) and beam atoms. Two sets of the sightlines with good spatial resolution are presented. One is the optimized viewing chords which have 20 sightlines and observe the whole plasma region with the spatial resolution Δρ less than ±0.055 using the newly designed viewing port. The other is 15 sightlines from the present viewing port of Heliotron J for the preliminary measurement to discuss the feasibility of the density fluctuation measurement by BES. The beam emission has been measured by a monochromator with a CCD camera. A good consistency has been obtained between the spectral profiles of the beam emission measured by the monochromator and the beam emission spectrum deduced by the model calculation. An avalanche photodiode with an interference filter system was also used to evaluate the signal-to-noise (S/N) ratio of the beam emission in the present experimental setup. The modification of the optical system is being planned to improve the S/N ratio, which will enable us to estimate the density fluctuation in Heliotron J.

An improved tomographic algorithm in the scheme of Tikhonov-Phillips regularisation method, has been employed for multi-pixel bolometric measurements in order to get as much information as possible while keeping the assumptions to a minimum. The effects of finite detector size have been taken into account with a full three dimensional treatment of the detector geometry. The application of tomographic imaging was implemented to a two-array AXUVD (Absolute eXtreme UltraViolet photodiode) camera on the Large Helical Device (LHD) and a two-dimensional infrared imaging bolometric (IRVB) pinhole camera on JT-60U. Pertinent examples of the results are presented both to illustrate the analysis techniques and to demonstrate the wealth of physics which can be studied.

We obtained the local foil properties of the JT-60U imaging bolometer foil (a single graphite-coated gold foil with an effective area of 9 × 7c m 2 and a nominal thickness of 2.5 µm) such as the thermal diffusivity, κ ,a nd the product of the thermal conductivity, k ,a nd the thickness, t f ,b y calibrating some parts of the foil. Calibration of the foil was made in situ using a He-Ne laser (∼27 mW) as a known radiation source to heat the foil. The thermal images of the foil are provided by an infrared (IR) camera (microbolometer type). The parameters are determined by finite element modeling (FEM) of the foil temperature and comparing the solution to the experimental results. In this work we apply this calibration technique to investigate the spatial variation of the foil parameters. Significant variation in the local temperature rise of the foil due to local heating by the laser beam indicates a spatial variation of the foil parameters κ, k and t f .T his variation is possibly due to nonuniformity in carbon coating and/or the thickness of the foil.

The infrared imaging video bolometer (IRVB) in JT-60U includes a single graphite-coated gold foil with an effective area of 9 × 7 cm 2 and a thickness of 2.5 μ m . The thermal images of the foil resulting from the plasma radiation are provided by an IR camera. The calibration technique of the IRVB gives confidence in the absolute levels of the measured values of the plasma radiation. The in situ calibration is carried out in order to obtain local foil properties such as the thermal diffusivity κ and the product of the thermal conductivity k and the thickness t f of the foil. These quantities are necessary for solving the two-dimensional heat diffusion equation of the foil which is used in the experiments. These parameters are determined by comparing the measured temperature profiles (for k t f ) and their decays (for κ ) with the corresponding results of a finite element model using the measured HeNe laser power profile as a known radiation power source. The infrared camera (Indigo/Omega) is calibrated by...

Emphases in JT-60 experiments are placed on (1) lower-hybrid (LH) current drive characteristics with a multi-junction type launcher, and (2) the confinement study with combination of neutral beam injection LH current drive and pellet injection. The new multi-junction LH launcher provides a sharp N/sub /// spectrum with high directivity for N/sub ///=1-3.4. The current drive efficiency and the radial distribution of high energy electron production show clear correlation with injected N/sub ///: the current drive efficiency has the maximum at low N/sub ///( approximately 1.3) while flattening of plasma current is more effective in large N/sub ///. A broad radial distribution of high energy electron current and approximately 30% reduction in sawtooth inversion radius were obtained by high N/sub /// ( approximately 2.5) LH injection. To fully suppress the sawtooth activity, low N/sub /// ( approximately 1.3) injection was found to be more effective. Improved energy confinement has been obtained with hydrogen pellet injection. Energy confinement time was enhanced up to 40% relative to usual gas fuelled discharges. The discharge has a strongly peaked electron density profile with ne(0)/(ne) approximately 5 and ne(0) approximately 2.0*1020 m-3. The improved discharges are characterized by a strongly peaked pressure profile within the q=1 magnetic surface, and degrades when a large sawtooth recovers or the pressure gradient may reach a critical value. When large (3 mm, 4 mm) and fast (2.2 km/s) pellets were injected, 30% energy confinement improvement was obtained even during the NB heating of 14 MW. Further investigations of IDC characteristics have been made. The oxygen impurity lines from the main plasma and the main radiative loss drop first. Then the plasma stored energy starts to rise. The particle recycling is reduced around the main plasma, and is localized in the neighborhood of the X-point with a time lag of approximately 0.2 sec. Eventually the discharge shows a significant remote radiative cooling power at the divertor region.

Initial ohmic heating experiments in JT-60 were performed for a three month period of April-June 1985. A maximum plasma current of 1.6 MA was obtained for both divertor and limiter discharges. Low-q discharges of qeff=2.5 and high density discharges of 4.8*1019 m-3 in line-averaged density were obtained in the divertor configuration. In divertor discharges radiated loss from the main plasma can be kept at 20-30% of the ohmic input.

Essential divertor functions — density and energy control — have been investigated in the JT-60 divertor with metal walls. Neutral pressure in the divertor chamber increases in proportion to n2e and high recycling state for particle exhaust is realized. At ne ≤ 6 × 1019m−3, wall pumping by the divertor plates is dominant in particle exhaust. However, at ne ≥ 6 × 1019m−3, particle exhaust by active divertor pumping systems becomes essential. Because of the effectiveness of the divertor, radiation loss in the main plasma is reduced to 5–10% of the absorbed power and impurity concentrations are significantly suppressed at very low level (Zeff = 1.5-2.0). In the experiments with graphite walls, short periods of H-mode phases were found in the outside X-point divertor discharges although improvement in energy confinement is limited to within 10%. In limiter discharges, combination of high vessel temperature (210–300 °C) and low current limiter discharges (1.5 MA) without gas puffing is effective as a wall conditioning in obtaining reproducible high density discharges. The maximum parameters, Ip = 3.2 MA, ne = 1.2 × 1020m−3, qeff = 2.2 at Bt = 4.8T, a Murakami parameter of 7.5, and a stored energy of 3.1 MJ have been attained. The maximum density is limited by the occurence of the MARFE. During high power neutral beam heating, enhanced carbon influx was observed possibly due to chemical sputtering. The wall pumping by the inner wall is still effective in decreasing the electron density from the high density region of ne ~ 1 × 1020m−3. An empirical energy confinement scaling of JT-60 has been drawn in terms of an offset linear function as τE = 0.19 I1.9PPabs + 0.062 a1.8p

From June to October 1987, JT-60 achieved fusion product (ne(0). tau E*.Ti(0)) of 6*1019 m-3.keV.s with hydrogen plasma at plasma current of 2.8 to 3.1 MA with neutral beam power of approximately 20 MW. The central electron density of 1.3*1020 m-3 was obtained at plasma current of 3 MA with 13 approximately 20 MW neutral beam power and the confinement time reached 0.14-0.18 s. An offset linear scaling law like the Shimomura-Odajima scaling on confinement time will be able to reproduce experimental data better than that of the Goldston type scaling. With low beam energy injection approximately 40 keV, confinement degradation was found. Many short periods (0.05 approximately 0.1 s) of H-mode phase were found in outside X-point divertor discharges with NB or NB+RF(LH or IC) heating power above 16 MW. However, improvement in energy confinement time was limited to 10 %. The ballooning/interchange stability analyses were also made for the outside X-point divertor equilibrium in connection with H0-phase capability. Heating powers of 9.5 MW and 1.9 MW were obtained by LHRF, ICRF injection, respectively. In combined LHRF and NB heating, the incremental energy confinement time of 0.064 s was obtained, which is the same level of that of NB heating only. In combined NB and on-axis ICRF heating of low ne discharge, an incremental energy confinement time of 0.21 s was obtained, which is three times as long as those of NB or ICRF heating only. High energy beam ions were accelerated by ICRF in the central region of the plasma.

In the last year, ohmic heating experiments were performed in JT-60 for about three months; May–June 1985 and March 1986. The major experimental effort has been concentrated on the divertor configuration in anticipation of H-mode operation of JT-60 in high power supplementary heating experiments starting in August 1986. Stable discharges with a divertor have been obtained at Ip = 1.85 MA and ne = 5.7 × 1019m−3. Radiated power from the main plasma in divertor discharges is reduced to one third of that in material limiter discharges. The electron density has a broad radial profile and the effective safety reaches qeff = 2.5. The neutral gas pressure in the divertor chamber becomes significantly high and the control of particle recycling has been attempted.

Recent plasma confinement experiments in Heliotron-D are described in connection with neoclassical diffusion theory. Weakly and relatively strongly ionized plasmas are produced by coaxialgun and Joule heating, respectively. In both cases, the observed diffusion coefficients are found to be in good agreement with the theory, under the conditions of weak magnetic field. Diffusion processes in two different magnetic configurations, epsilon t approximates epsilon h an d epsilon t << epsilon h are also compared experimentally, where epsilon t is t he ratio r/R of the distance from the magnetic axis to the major radius, and epsilon /sub h/ is the modulation depth due to the helical field.

A major portion of the DIII-D program includes studies of the L-H transition, of the VH-mode, of particle transport and control and of the power-handling capability of a divertor. Significant progress has been made in all of these areas and the purpose of this paper is to summarize the major results obtained during the last two years. An increased understanding of the origin of improved confinement in H-mode and in VH-mode discharges has been obtained, good impurity control has been achieved in several operating scenarios, studies of helium transport provide encouraging results from the point of view of reactor design, an actively pumped divertor chamber has controlled the density in H-mode discharges and a radiative divertor is a promising technique for controlling the heat flux from the main plasma.