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Spatially resolved rotational temperature of ground state hydrogen molecules desorbed from plasma-facing surface was measured in QUEST, LTX-β, and DIII-D tokamaks, and the increases of the rotational temperature with the surface temperature and due to collisional-radiative processes in the plasmas were evaluated. The increase due to collisional-radiative processes was calculated by solving rate equations considering electron and proton collisional excitation and deexcitation and spontaneous emission. The calculation results suggest a high sensitivity for the rotational temperature to electron and proton densities, but a negligible sensitivity to the electron, proton, and surface temperatures. In the three tokamaks with different plasma parameters and plasma facing surface materials, the spatial profile of the rotational temperature was estimated using Fulcher-α emission lines (600-608 nm). In QUEST, the spatial profile of the rotational temperature was estimated from spatially resolved spectra. In the other tokamaks, the rotational temperature was evaluated assuming a single point emission with a location determined from the Fulcher-α emission profile as measured with a filtered camera. In metal-walled devices QUEST and LTX-β, the rotational temperature increased with the surface temperature, and the calculated collisional-radiative increase is consistent with measured increase assuming that the rotational temperature at the surface is approximately 450 K higher than the surface temperature. In DIII-D with carbon walls, a larger collisional-radiative increase than the other tokamaks was observed because of the higher density leading to a large difference from the calculated increase compared to the other smaller tokamaks. Measurement of the Fulcher-α emission profile with higher spatial resolution in DIII-D may reduce the difference and reveal the effect of the surface temperature on the rotational temperature. These results shows the increases in the rotational temperature with the surface temperature and due to the collisional-radiative processes.

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The spatial distribution of the hydrogen atom density was evaluated in a spherical tokamak (ST) plasma sustained only with 28 GHz electron cyclotron heating (ECH). The radially resolved H&delta, emissivity was measured using multiple viewing chord spectroscopy and Abel inversion. A collisional-radiative (CR) model analysis of the emissivity resulted in a ground-state hydrogen atom density of 1015&ndash, 1016 m&minus, 3 and an ionization degree of 1&ndash, 0.85 in the plasma.

Spherical tokamak (ST) research in Japan has produced many innovative results: (i) plasma start-up to I p > 70 kA was achieved by electron cyclotron wave (ECW) with N ∥ = 0.75, while electron heating to T e > 500 eV was achieved with N ∥ = 0.26 on QUEST. (ii) The radiofrequency (RF)-induced transport model was combined with the x-ray emission model, and extended magnetohydrodynamics equilibrium with kinetic electrons was developed to interpret fast-electron-dominated lower hybrid wave sustained plasmas on TST-2. (iii) Density as high as 30 times the cutoff density was achieved by electron Berstein wave current drive combined with electron beam injection on LATE. (iv) Multiple plasmoids formed by tearing instability in the elongated current sheet were observed, and flux closure and ion heating by plasmoid-mediated fast magnetic reconnection were observed on HIST. (v) Optimization of ECW-assisted inductive start-up with a vertical field with positive decay index was performed on TST-2. (vi) Stabilization of the vertical displacement event by a set of upper and lower helical field coils was demonstrated on TOKASTAR-2. (vii) A 6 h discharge was achieved by cool-down of the center stack cover on QUEST, where the plasma duration limit was consistent with the wall saturation time estimated by modeling. (viii) Extension of ion heating by plasma merging was achieved on TS-3U, TS-4U, UTST, MAST, and ST40.

Presence of a large population of fast particles may qualitatively modify the tokamak equilibrium from the one that can be described by the standard magneto-hydrodynamics (MHD). The kinetic modification of the MHD equilibrium was studied for a lower-hybrid wave driven plasma on the TST-2 spherical tokamak. The analysis was performed using an equilibrium reconstruction method based on an extended MHD model that considers a two-component plasma of bulk MHD fluid and kinetic fast electrons. Scrape-off-layer current appeared naturally in the extended MHD analysis because of the finite electron orbit excursion from the flux surfaces. This reduced the bulk pressure contribution to the toroidal current substantially. The resulting poloidal flux profile was more consistent with that inferred from the Thomson scattering measurement.

Fully non-inductive second (2nd) harmonic electron cyclotron (EC) plasma current ramp-up was demonstrated with a newlly developed 28 GHz system in the QUEST spherical tokamak. A high plasma current of 54 kA was non-inductively ramped up and sustained stably for 0.9 s with a 270 kW 28 GHz wave. A higher plasma current of 66 kA was also non-inductively achieved with a slow ramp-up of the vertical field. We have achieved a significantly higher plasma current than those achieved previously with the 2nd harmonic EC waves. This fully non-inductive 2nd harmonic EC plasma ramp-up method might be useful for future burning plasma devices and fusion reactors, in particular for operations at half magnetic field with the same EC heating equipment.

The plasma current is ramped up primarily by a 28 GHz electron cyclotron wave (ECW) in the Q-shu University experiment Steady-State Spherical Tokamak (QUEST), with multiple harmonic resonance layers from the second to the fourth stay in the plasma core. A steering antenna comprising two quasi-optical mirrors enhances the power density of ECWs. The ECW beam is injected obliquely from the low-field side where the parallel refractive index is N ∥ = 0.75 at the second-harmonic resonance layer. Analysis of the resonance condition has found that energetic electrons moving forward along the magnetic field resonate more effectively than those moving backward. Such symmetry breaking is consistent with the results of the current ramp-up experiment. The peak plasma current reaches I p > 70 kA, constantly injecting a beam of radio frequency power of 100 kW. Ray-tracing by the TASK/WR code demonstrates that the power of the 28 GHz extraordinary mode is absorbed by energetic electrons via single-pass cyclotron absorption.

An O-mode microwave imaging reflectometry system has been installed in a spherical tokamak named TST-2. The illumination wave frequency is 23 - 32 GHz, which corresponds to the cutoff electron density of 0.65 - 1.3 × 1019 m−3. The microwave image of the scattered wave is formed on an imaging detector named horn-antenna millimeter-wave imaging detector by imaging optics that consist of an ellipsoidal aluminum mirror, a Teflon lens, and a dielectric plate. The detected channel numbers are 6 (poloidal) × 6 (toroidal) × 2 (radial). The data of power and phase of the scattered wave are sampled every 0.5 μs and are stored by the LABCOM system at the National Institute for Fusion Science via a private network named SNET. The scattered wave quickly fluctuates (∼5 μs) and has a large amplitude (>100 times). From the time evolution of the phase at the internal reconnection event, it is inferred that the reduction in amplitude may be owing to the phase mixing of scattered wave in each detector channel.

A compact hard X-ray camera has been developed to study the behavior of fast electrons accelerated by lower hybrid waves in the TST-2 spherical tokamak. The camera consists of a pinhole made of tungsten alloy, a thin resin window, a thin LYSO scintillator disk, 5 fiber optic light guides and 5 photomultiplier tubes and shields. The camera is installed on a tangential port, and two-dimensional images are obtained by rotating the system around the axis of the port on a shot by shot basis. The typical energy range is 10 - 250 keV, and the energy resolution is about 100% (FWHM) at 100 keV. The measured images of a lower hybrid wave driven plasma show bright region at the inboard side (i.e., high field side).

A 28~GHz system with a high-power gyrotron tube has been used for the QUEST spherical tokamak to form an over-dense plasma for electron Bernstein wave heating (EBWH) and current drive with an 8.2~GHz-wave. Non-inductive high-density plasma ramp-up experiments with dual-frequency (dual$-f$) electron cyclotron (EC) [ 8.2~GHz and 28~GHz ] waves were conducted. A spontaneous density jump (SDJ) to an over-dense state was observed as a bifurcation phenomenon in the dual$-f$ wave experiment. The over-dense plasma on the 8.2~GHz-wave was non-inductively ramped up to 25~kA, and was maintained for 0.4~s under stable plasma equilibrium after two such jumps in one shot. Heating to mildly energetic electrons and bulk electrons was observed even in the over-dense region. The electrostatic EBWH effect on the mildly energetic electrons in the over-dense region is assessed following a dispersion analysis of the 8.2 GHz-wave. The bulk electron heating effect observed is explained as heat exchange from mildly energetic electrons heated by the electrostatic EBWH wave. Remarkably, a high hard-X-ray-radiation temperature ~ 500~keV was also observed in tangential viewing for current-carrying electrons in the over-dense core region. Synergetic heating from the overlap of different 28~GHz EC harmonic resonances as well as higher harmonic heating is discussed for maintaining the highly energetic electrons in the over-dense core region. In addition, the SDJ process and mechanism are considered based on the discussion of the electron heating effects with the 8.2 GHz-wave.

A CW phased-array antenna system for electron cyclotron/Bernstein wave heating and current drive (ECWH/CD, EBWH/CD) experiments was developed in the QUEST. The antenna was designed to excite an elliptically polarized pure O-mode wave in oblique injection for the O-X-B mode conversion scenario, and its good performance was confirmed at a high power level. Long pulse discharges with a plasma current of 10 kA and 15 kA were non-inductively attained for 37 s and 20 s, respectively, with only radio frequency (RF) power. Divertor configurationswere also obtained in the RF-sustained plasmas. A new operational window for sustained plasma current was observed in the high-density plasma with a higher RF incident power. Two new heating and current drive systems with an 8.56GHz klystron and a 28GHz gyrotron are being prepared to conduct CW EBWH/CD experiments in the high-density plasma. © 2012 The Japan Society of Plasma Science and Nuclear Fusion Research.

An indirect effect of the electron cyclotron current drive (ECCD) on the ion temperature in the plasma core region was observed in the Large Helical Device. The reference (no ECCD) discharge with a central ion temperature Ti0 of ∼3.0 keV is operated by a standard high ion temperature discharge procedure. To investigate the ECCD effect, a co- or counter-ECCD was applied to the reference discharge, and was turned off immediately before the Ti0 peaked in the reference discharge. In the co-ECCD and counter-ECCD applications, the Ti0 temporarily increased and decreased by ∼0.5 keV from Ti0 in the reference discharge, respectively. The mechanism of this phenomenon is presently unclear, but may be exploited as a practical knob for controlling the central ion temperature.

The poloidal eigenmode of the geodesic acoustic mode (GAM) is studied in the limit of high safety factor. In this limit, the poloidal gyroradius cannot be treated as a perturbation or as an expansion parameter. Analytical expressions for the poloidal structure of the GAM potential, the radial wavenumber dependence of the frequency, the phase velocity, and the group velocity are obtained. The spatial structure of the poloidal eigenmode including the higher-order gyroradius effect is revealed theoretically.

The radial eigenmode of the geodesic acoustic mode (GAM) is investigated, taking into account the finite ion Larmor radius effect. The radial eigenfunction and the eigenfrequency (real frequency and damping rate) are obtained numerically. The eigenfrequency is found to be not continuous, but discrete. In addition, we obtain an analytic expression for the real eigenfrequency, which shows good agreement with numerical results. (© 2008 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

A method for zonal flow study by using direct density fluctuation measurements is proposed. When ambient drift-wave turbulence is modulated by zonal flows (i.e. in the drift-wave-zonal flow systems), an envelope of the ambient density fluctuations has spectral peaks around zonal flow frequencies. A spectral peak at the geodesic acoustic mode (GAM) frequency is observed in the envelope of the ambient density fluctuations measured in edge plasma of the JFT-2M tokamak. The significant cross-bicoherence is also found between the ambient density fluctuations and its envelope in the GAM frequency. This result demonstrates that we can measure the GAM only by using density fluctuation data. This method provides a possibility of zonal flow research in burning core plasma by density fluctuation diagnostics such as microwave reflectometry.

Single-particle orbits in a low aspect ratio axisymmetric configuration have been studied to clarify their role in current initiation using electron cyclotron waves. In this study, electron guiding centre orbits are studied neglecting collisions and charge exchange. Current is generated mainly by the toroidal drift of mirror-trapped electrons. Analytic scaling laws for the current and the poloidal field generation are derived, and they are found to be functions of temperature, poloidal field strength, number of electrons and parameters defining the magnetic and geometrical configuration. The current increases with temperature and decreases with the poloidal field strength as long as the temperature is much smaller than the characteristic temperature, whose thermal velocity is defined by the product of poloidal cyclotron frequency and the major radius.

Several techniques for initiating the plasma current without the use of the central solenoid are being developed in TST-2. While TST-2 was temporarily located at Kyushu University, two types of start-up scenarios were demonstrated. (1) A plasma current of 4 kA was generated and sustained for 0.28 s by either electron cyclotron wave or electron Bernstein wave, without induction. (2) A plasma current of 10 kA was obtained transiently by induction using only outboard poloidal field coils. In the second scenario, it is important to supply sufficient power for ionization (100 kW of EC power was sufficient in this case), since the vertical field during start-up is not adequate to maintain plasma equilibrium. In addition, electron heating experiments using the X–B mode conversion scenario were performed, and a heating efficiency of 60% was observed at a 100 kW RF power level. TST-2 is now located at the Kashiwa Campus of the University of Tokyo. Significant upgrades were made in both magnetic coil power supplies and RF systems, and plasma experiments have restarted. RF power of up to 400 kW is available in the high-harmonic fast wave frequency range around 20 MHz. Four 200 MHz transmitters are now being prepared for plasma current start-up experiments using RF power in the lower-hybrid frequency range. Preparations are in progress for a new plasma merging experiment (UTST) aimed at the formation and sustainment of ultra-high β ST plasmas.

Plasma start-up and sustainment without an inductive field have been studied in the TST-2@K spherical tokamak using high power RF sources (8.2?GHz/up to 170?kW). Steady state discharges with a plasma current of 4?kA were achieved. The line integrated density was about 3 ? 1017?m?2 and the electron temperature was 160?eV. A truncated equilibrium was introduced to reproduce magnetic measurements. It was found that a positive Pfirsch?Schl?ter current in the open field line region at the outboard boundary makes a significant contribution to the current. Insensitivity of the current to variations in the vertical field and RF power variation was also found.

Experiments aiming at 'day long operation at high performance' have been carried out in TRIAM-1M. The record value of the discharge duration was updated to 5 h and 16 min. Steady-state tokamak operation is studied under the localized plasma wall interaction conditions. The distributions of the heat load, the particle recycling flux and impurity source are investigated to understand the co-deposition and wall pumping. The formation and sustainment of an internal transport barrier (ITB) in enhanced current drive mode has been investigated by controlling the lower hybrid driven current profile by changing the phase spectrum. An ITER relevant remote steering antenna for electron cyclotron wave injection was installed and a relativistic Doppler resonance of the oblique propagating extraordinary wave with energetic electrons driven by lower hybrid waves was studied.

The spatial structure and Z dependence of transport coefficient were studied by injecting impurity pellets (C, Al, and Ti) and measuring high-spatial resolution bremsstrahlung in steady-state neutral beam heated plasmas on the Large Helical Device [A. Iiyoshi et al., Fusion Techol. 17, 169 (1990)]. The D and V were evaluated from a diffusive/convective model with minimizing the residual error between the experimental and computational bremsstrahlung intensities taking the recycling rate into account. As a result, the existence of an inward convection was observed at ρ>0.6, and no convection was required at ρ

The 'fishbone' fast wave travelling wave antenna was developed for LHD to provide a capability for rotational transform profile control by current drive. The fishbone antenna is equivalent to two combline antennae stacked vertically. The antenna operates around 75 MHz and excites a wavenumber of 14 m−1 when the phase difference between adjacent current straps is 90°. A test of a combline antenna with plasma load on the TST-2 spherical tokamak suggested the possibility that this type of antenna does not need to be installed in the immediate vicinity of the last closed flux surface. Optimization of the design was performed based on measurements on mock-up antennas and model calculations. In the fishbone antenna, controlled excitation of the even mode (with currents in the top and bottom halves of a current strap in the same direction) is necessary. A predominant excitation of the even mode was realized in the LHD fishbone antenna with simulated loading by selecting an appropriate operating frequency.

Compatibility between the plasma and low activation ferritic steel, which is a candidate material for fusion demonstration reactors, has been investigated step by step in the JFT-2M tokamak. We have entered the third stage of the Advanced Material Tokamak EXperiment (AMTEX), where the inside of the vacuum vessel wall is completely covered with ferritic steel plates ferritic inside wall (FIW). The effects of a FIW on the plasma production, impurity release, the operation region, and H-mode characteristics have been investigated. No negative effect has been observed up to now. A high normalized beta plasma of βN ~ 3, having both an internal transport barrier and a steady H-mode edge was obtained. A remarkable reduction in ripple trapped loss from 0.26 MW m−2 (without ferritic steel) to less than 0.01 MW m−2 was demonstrated by the optimization of the thickness profile of FIW. A code to calculate fast ion losses, taking into account the full three-dimensional magnetic structure was developed, and values obtained using the code showed good agreement with experimental results. Thus, encouraging results are obtained for the use of this material in the demo-reactor.

Various types of MHD events including internal reconnection events are studied on the TST-2 spherical tokamak. In weak MHD events no positive current spike was observed, but in strong MHD events with positive current spikes, a rapid and significant impurity ion temperature increase was observed. The decrease in the poloidal magnetic energy is the most probable energy source for ion heating. The plasma current shows a stepwise change. The magnitude of this step correlates with the temperature increase and is found to be a good indicator of the strength of each event.

The National Spherical Torus Experiment (NSTX) at Princeton University, Princeton, NJ, is designed to assess the potential of the low-aspect-ratio spherical torus concept for magnetic plasma confinement. The plasma has been heated by up to 7 MW of neutral beam injection (NBI) at an injection energy of 100 keV and up to 6 MW of high harmonic fast wave (HHFW) at 30 MHz. NSTX has achieved /spl beta//sub T/ of 32%. A variety of MHD phenomena have been observed to limit /spl beta/. NSTX has now begun addressing /spl tau//sub E/ scaling, /spl beta/ limits, and current drive issues. During the NBI heating experiments, a broad T/sub i/ profile with T/sub i/ up to 2 keV, T/sub i/>T/sub e/ and a large toroidal rotation were observed. Transport analysis suggests that the impurity ions have diffusivities approaching neoclassical. For L-Mode plasmas, /spl tau//sub E/ is up to two times the ITER97L L-Mode scaling and exceeds the ITER98pby2 H-Mode scaling in some cases. Transitions to H-Mode have been observed which result in an approximate doubling of /spl tau//sub E/ after the transition in some conditions. During HHFW heating, T/sub e/>T/sub i/ and T/sub e/ up to 3.5 keV were observed. Current drive has been studied using both coaxial helicity injection with up to 390 kA of toroidal current and HHFW. HHFW has produced H-modes with significant bootstrap current fraction at low I/sub p/, high q, and high /spl beta//sub p/.

AC Ohmic coil operation experiments with frequencies up to 10 kHz were performed on the TST-2 spherical tokamak device, and the pre-ionization process was studied in detail. The minimum loop voltage for pre-ionization was 0.4 V, which corresponds to 0.5 V ${\rm m}^{-1}$ at the inboard limiter. Dependences of growth rate and saturation level of the process on various parameters were obtained, and they are compared with a time-dependent 0-dimensional model based on Townsend avalanche and loss along field lines. Most of the dependences are reproduced qualitatively by the model, and quantitative differences are within a factor of several. However, the external vertical field dependence of the appearance time, which is defined as the time to observe a plasma, and the isotope effect cannot be reproduced by the model. An ambipolar diffusion state which is predicted theoretically but mitigated experimentally is discussed. It was found that secondary electron emission at the limiter surfaces is a candidate mechanism to mitigate the state.

Non-inductive plasma start-up is a critical issue for spherical tokamaks since there is not enough room to provide neutron shielding for the center solenoid. Start-up using lower hybrid (LH) waves has been studied on the TST-2 spherical tokamak. Because of the low magnetic field of a spherical tokamak, the plasma density needs to be kept at a very low value during the plasma current ramp-up so that the plasma core remains accessible to the LH waves. However, we have found that higher density was required to sustain larger plasma current. The achievable plasma current was limited by the maximum operational toroidal field of TST-2. The existence of an optimum density for LH current drive and its toroidal field dependence is explained through a numerical simulation based on a ray tracing code and a Fokker–Planck solver. In order to access higher density at the same magnetic field, a top-launch antenna was recently installed in addition to the existing outboard-launch antenna. Increase in the density limit was observed when the power was launched from the top antenna, consistently with the numerical predictions.

As the finalization of a hydrogen experiment towards the deuterium phase, the exploration of the best performance of hydrogen plasma was intensively performed in the large helical device. High ion and electron temperatures, Ti and Te, of more than 6 keV were simultaneously achieved by superimposing high-power electron cyclotron resonance heating onneutral beam injection (NBI) heated plasma. Although flattening of the ion temperature profile in the core region was observed during the discharges, one could avoid degradation by increasing the electron density. Another key parameter to present plasma performance is an averaged beta value $\left\langle \beta \right\rangle $ . The high $\left\langle \beta \right\rangle $ regime around 4% was extended to an order of magnitude lower than the earlier collisional regime. Impurity behaviour in hydrogen discharges with NBI heating was also classified with a wide range of edge plasma parameters. The existence of a no impurity accumulation regime, where the high performance plasma is maintained with high power heating >10 MW, was identified. Wide parameter scan experiments suggest that the toroidal rotation and the turbulence are the candidates for expelling impurities from the core region.

Nationally coordinated research on spherical tokamak is being conducted in Japan. Recent achievements include: (i) plasma current start-up and ramp-up without the use of the central solenoid by RF waves (in electron cyclotron and lower hybrid frequency ranges), (ii) plasma current start-up by AC Ohmic operation and by coaxial helicity injection, (iii) development of an advanced fuelling technique by compact toroid injection, (iv) ultra-long-pulse operation and particle control using a high temperature metal wall, (v) access to the ultra-high-β regime by high-power reconnection heating, and (vi) improvement of spherical tokamak plasma stability by externally applied helical field.

Measurements of edge plasma parameters such as current density, electron density, and electron temperature were performed during internal reconnection events in TST-2 Ohmic plasmas. The measured current density consists of two components: a slowly varying component and a spiky bipolar component. The magnitude of the slowly varying component is comparable to the mean current density averaged over the poloidal cross section, and it seems to represent the global transport from the core to the edge. The spiky bipolar component is about an order of magnitude larger than the slowly varying component, but the spatial structure seems to be localized and its effect on plasma confinement is not catastrophic.

The initial experiments on the Large Helical Device (LHD) have extended confinement studies on currentless plasmas to a large scale (R = 3.9 m, a = 0.6 m). Heating by NBI of 3 MW produced plasmas with a fusion triple product of 8 × 1018m-3keVs at a magnetic field strength of 1.5 T. An electron temperature of 1.5 keV and an ion temperature of 1.1 keV were achieved simultaneously at a line averaged electron density of 1.5 × 1019 m-3. The maximum stored energy reached 0.22 MJ with neither unexpected confinement deterioration nor visible MHD instabilities, which corresponds to β = 0.7%. Energy confinement times reached a maximum of 0.17 s. A favourable dependence of energy confinement time on density remains in the present power density (~40 kW/m3) and electron density (3 × 1019 m-3) regimes, unlike the L mode in tokamaks. Although power degradation and significant density dependence are similar to the conditions on existing medium sized helical devices, the absolute value is enhanced by up to about 50% from the International Stellarator Scaling 95. Temperatures of both electrons and ions as high as 200 eV were observed at the outermost flux surface, which indicates a qualitative jump in performance compared with that of helical devices to date. Spontaneously generated toroidal currents indicate agreement with the physical picture of neoclassical bootstrap currents. Change of magnetic configuration due to the finite β effect was well described by 3-D MHD equilibrium analysis. A density pump-out phenomenon was observed in hydrogen discharges, which was mitigated in helium discharges with high recycling.

The Large Helical Device (LHD) has successfully started running plasma confinement experiments after a long construction period of eight years. During the construction and machine commissioning phases, a variety of milestones were attained in fusion engineering which successfully led to the first operation, and the first plasma was ignited on 31 March 1998. Two experimental campaigns were carried out in 1998. In the first campaign, the magnetic flux mapping clearly demonstrated a nested structure of magnetic surfaces. The first plasma experiments were conducted with second harmonic 84 and 82.6 GHz ECH at a heating power input of 0.35 MW. The magnetic field was set at 1.5 T in these campaigns so as to accumulate operational experience with the superconducting coils. In the second campaign, auxiliary heating with NBI at 3 MW has been carried out. Averaged electron densities of up to 6 × 1019m-3, central temperatures ranging from 1.4 to 1.5 keV and stored energies of up to 0.22 MJ have been attained despite the fact that the impurity level has not yet been minimized. The obtained scaling of energy confinement time has been found to be consistent with the ISS95 scaling law with some enhancement.

Variation of the plasma position relative to the centre of the helical coil winding is a very effective means of controlling the MHD stability and the trapped particle confinement in heliotron/torsatron systems, but improving one of these two characteristics with this parameter simultaneously has a detrimental effect on the other. The inward shifted configuration is favourable for drift orbit optimization but is predicted to be unstable according to the Mercier criterion. Various physics problems, such as electric field structure, plasma rotation and MHD phenomena, have been studied in the Compact Helical System (CHS) with a compromise intermediate position. With this standard configuration, CHS has yielded experimental results that contribute to the understanding of general toroidal confinement physics and low aspect ratio helical systems. In the recent experiments, it was found that a wide range of inward shifted configurations give stable plasma discharges without any restriction to the special pressure profile. Such an enhanced range of operation made it possible to study experimentally the drift orbit optimized configuration in heliotron/torsatron systems. The effect of configuration improvement was studied with plasmas in a low collisionality regime.

Effects of magnetic shear on electron cyclotron resonance heating (ECRH) are studied in heliotron/torsatron configurations. In such configurations, the poloidal magnetic field is comparable to the toroidal magnetic field, and varies spatially along the minor radius, making a strong magnetic shear. When high power millimeter waves are launched into a plasma, it is coupled to propagating modes at the plasma peripheral region. The existence of a transition layer between the core plasma region and the vacuum region, where the magnetic field direction is largely changed, requires accurate polarization control for good single pass absorption. The mode conversion between the propagation modes due to the magnetic shear also affects the launching conditions. The polarization control experiment by using second harmonic ECRH in Heliotron E [T. Obiki, Fusion Technol. 17, 101 (1990)] are compared with the numerical calculation in which one dimensional second order coupled equations are solved. The polarization dependence experimentally measured is in good agreement with the numerical results including the magnetic shear terms.