Search

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.

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.

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.

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 '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.

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.

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.

Toroidal current penetration is studied in current ramp experiments on the JIPP T-IIU tokamak. The poloidal magnetic field profile in the peripheral region of a plasma (0.5 ≤ r/a ≤ 1.0) has been measured directly with a newly developed fast response Zeeman polarimeter. The experimental results indicate that a skin effect of the toroidal current density is clearly observed during both the current ramp-up (CRU) and the current ramp-down (CRD) experiments. The experimentally obtained toroidal current density profiles are well described by the profiles calculated on the assumption of neoclassical electrical resistivity, although the difference between the neoclassical electrical resistivity and the Spitzer electrical resistivity is not clearly distinguished because of the relatively collisional peripheral plasma region. Quasi-linear Δ' analysis of tearing modes for the measured current density profile is consistent with the time behaviour of the coherent magnetohydrodynamic (MHD) modes such as m = 4/n = 1 or m = 3/n = 1 (m is the poloidal mode number and n is the toroidal mode number) often observed during the CRU phase. An obvious enhancement of the current penetration by the MHD modes is not detected in this ramp-up experiment, where the relative amplitude of the poloidal field fluctuations is less than 0.8%.

Laser scattering signals that indicate the presence of small dust particles (diameter ≤ 2 μm) have been occasionally observed in the JIPPT-IIU tokamak chamber. This phenomenon was reproduced deliberately by spreading carbon dust from the top of the vacuum chamber. No noticeable effect on the plasma was observed for dust falls of up to at least 106 particles (10 μg) in 20 ms during discharges. Dust falling just before plasma startup seemed to be confined but it was soon ejected (in less than 30 ms)

Multiple and small sample volume measurements of the density turbulence and potential profile measurements in tokamak plasmas were conducted using a heavy ion beam probe. The obtained wavenumber-frequency spectrum S(k, omega ) shows that the cross-sections of neutral beam injection (NBI) heated plasmas are divided into three regions of different turbulence characteristics outside the layer where the direction of the density turbulence propagation reverses, a low frequency and low wavenumber mode travelling in the ion diamagnetic drift direction dominates. The region enclosed by this reversal layer is divided into two parts during nearly perpendicular NBI heating: a region where the propagation velocity is near the Etau /Bt poloidal rotation velocity and a bad curvature region of very small wavenumber and high propagation velocity. The region of high propagation velocity, found in NBI plasmas, disappears in the ohmic plasmas. In addition, a small component that propagates in the ion diamagnetic drift direction is observed in NBI plasmas

An ion cyclotron range of frequencies (ICRF) heating experiment was performed on the Compact Helical System (CHS) device in order to identify the problems in establishing this as a reliable heating method. Radiofrequency heating was applied to plasmas with two ion species produced by ECH. Stored energy increased upto 2.2 kJ with the application of a 590 kW ICRF heating pulse, giving a heating efficiency comparable to that of NBI heating. More importantly, the `flat-top' of the stored energy was maintained during the pulse. Good heating was realized with 30% proton and 70% deuteron plasmas at high density (4*1019 m-3). It was found that the best discharges are dominated by electron ICRF heating, which seems to be the reason for the successful results. On the contrary, a flat-top in stored energy has never been obtained in discharges where ion heating dominates. The loss of high energy ions with large pitch angles produced by ion heating appears to degrade the performance and limit the duration of the heating. A similar degradation was encountered in previous ICRF heating experiments in helical systems. A newly designed detector was used to study the loss of the helically trapped particles

Fast changes of electric potential with different polarities are observed during sawtooth oscil- lations in the core region of a tokamak plasma using a heavy ion beam probe. Near the inversion radius the polarity of the observed change of the potential is found to be dependent on the swift movement of the hot core at the crash and is consistent with the prediction of one- fluid MHD theory. Near the magnetic axis the change of the potentials is positive and outside the inversion radius the change is negative. This is in contradiction with the MHD prediction. A periodic crash of the central electron tempera- ture called a sawtooth oscillation is a common fea- ture of tokamak plasmas (l-61. After the crash, a flattening of the pressure throughout the q = 1 sur- face is observed. Kadomtsev proposed a theoretical model of the resistive reconnection (7). The theory predicted the flattening of the q profile as well as the pressure profile inside the q = 1 surface. The predic- tion was, however, found to be in contradiction with recent measurements of the current density in toka- mak plasmas (8). In addition, experiments in large machines such as JET and TFTR showed much faster crashes than those predicted by resistive reconnection (2-61. More refined theories, which take account of the kinetic effect and ergodicity on the reconnections, have now been proposed (9-121. The present Letter reports the first measurement of potential changes during sawtooth oscillations and discusses the impli- cations of the results. The experiment is performed with the JIPP T-IIU tokamak (13). Its major radius is 93 cm and the maximum toroidal field is 3 T. The experiment is conducted at a relatively low electron density of about 2 x 1013 cm3, so that good beam penetra- tion into the centre of the plasma could be obtained. A 450 keV singly ionized thallium beam is injected

The launching conditions for an electron cyclotron wave in a plasma with a sheared magnetic field are studied. One dimensional second order coupled equations are solved for the ordinary and extraordinary modes. It is shown that, for wave absorption, the rotation angle and ellipticity of the wave should be determined by taking into account the effect of the sheared magnetic field, particularly in heliotron/torsatron configurations. Launching conditions for effective heating are obtained for second harmonic extraordinary mode heating in the Heliotron-E, CHS and LHD helical devices

High beta plasmas with a volume averaged equilibrium beta value of 2.1% were produced in CHS using tangential neutral beam injection. This beta value was achieved with the confinement improvement (reheat mode) observed after turning off strong gas puffing. Wall conditioning with titanium gettering was used to make high density operation (ne ? 8 ? 1019 m-3) possible for low magnetic fields (Bt = 0.6 T). The discharges start with the magnetic hill configuration (in vacuum) and finally achieve Mercier stable equilibrium owing to the self-stabilization effect given by the magnetic well which is produced by the plasma pressure. The Shafranov shift was about 40% of a plasma minor radius. Magnetic fluctuations did not increase with increasing plasma pressure when the beta value exceeded 1%. Dynamic poloidal field control was applied to suppress the outward plasma shift with increasing plasma pressure. Such operation gave an additional increase of beta value compared with the constant poloidal field operation

Successful production of high ?p plasmas (??p???1) fully non-inductively (NI) and their long pulse sustainment with the help of modest power (

Non-inductive plasma current start-up and sustainment by waves in the lower-hybrid frequency range (200 MHz) have been studied on the TST-2 spherical tokamak (R0 ⩽ 0.38 m, a ⩽ 0.25 m, Bt0 ⩽ 0.3 T, Ip ⩽ 0.14 MA) using three types of antenna: the 11-element inductively-coupled combline antenna, the dielectric loaded 4-waveguide array antenna, and the 13-element capacitively-coupled combline (CCC) antenna. The maximum plasma currents of 15 kA, 10 kA and 16 kA were achieved, respectively. The highest current drive figure of merit was achieved by the CCC antenna. The efficiency of current drive should improve by reducing prompt orbit losses of high energy electrons by operating at higher plasma current (to improve orbit confinement) and higher toroidal magnetic field (to improve wave accessibility to the plasma core), while keeping the density high enough (to avoid excessive acceleration of electrons), but under the ‘density limit’.

Plasma current start-up and ramp-up using the lower hybrid wave (LHW) were investigated on the TST-2 spherical tokamak. The LHW was launched by a dielectric-loaded waveguide array (grill) antenna. The antenna–plasma coupling of this antenna deteriorates as the input power exceeds several kW. This deterioration is believed to be caused by the density depletion due to the ponderomotive force. This conjecture was confirmed by the measurement of density reduction and the result of a non-linear full wave numerical calculation based on the finite element method (FEM). The plasma current was started and ramped up to 10 kA using this antenna. The ability of this grill antenna to excite the LHW with different n∥ = ck∥/ω was used to identify the most favourable n∥ spectrum for plasma current ramp-up. It was found that effective current drive can be achieved by the LHW with n∥ less than 6. However, even in this case, the energetic electrons which account for a large fraction of the driven current, are lost rapidly because the poloidal field generated by this level of plasma current is not sufficient to confine high energy electrons.

The progress of physical understanding as well as parameter improvement of net-current-free helical plasma is reported for the Large Helical Device since the last Fusion Energy Conference in Daejeon in 2010. The second low-energy neutral beam line was installed, and the central ion temperature has exceeded 7 keV, which was obtained by carbon pellet injection. Transport analysis of the high-T-i plasmas shows that the ion-thermal conductivity and viscosity decreased after the pellet injection although the improvement does not last long. The effort has been focused on the optimization of plasma edge conditions to extend the operation regime towards higher ion temperature and more stable high density and high beta. For this purpose a portion of the open helical divertors are being modified to the baffle-structured closed ones aimed at active control of the edge plasma. It is compared with the open case that the neutral pressure in the closed helical divertor increased by ten times as predicted by modelling. Studies of physics in a three-dimensional geometry are highlighted in the topics related to the response to a resonant magnetic perturbation at the plasma periphery such as edge-localized-mode mitigation and divertor detachment. Novel approaches of non-local and non-diffusive transport have also been advanced.

Plasma current (Ip) start-up in a spherical tokamak (ST) by waves in the lower-hybrid (LH) frequency range was investigated on TST-2. A low current (~1 kA) ST configuration can be formed by waves over a broad frequency range (21 MHz–8.2 GHz in TST-2), but further Ip ramp-up (to ~10 kA) is most efficient with waves in the LH frequency range. Ip ramp-up to 15 kA was achieved with 60 kW of net RF power PRF in the fast wave (FW) polarization at 200 MHz excited by the inductively coupled combline antenna. X-ray measurements showed that the photon flux and temperature are higher in the direction opposite to Ip, consistent with acceleration of electrons by a uni-directional RF wave. There is evidence that the LH wave is excited nonlinearly by the FW, based on the frequency spectra measured by magnetic probes. Similar efficiencies of Ip ramp-up were obtained with the inductive combline antenna and the dielectric-loaded waveguide array ('grill') antenna, and tendencies for the current drive efficiency to increase with plasma current and toroidal field were observed. During operation of the grill antenna, wavevector components were measured by an array of magnetic probes. Results were qualitatively consistent with expectations based on dispersion relations for the FW and the LH wave. A capacitively coupled combline antenna has been developed to improve coupling to the plasma and the wavenumber spectrum of the excited LH wave, and will be tested in 2013.

Spherical tokamaks (STs) have the advantage of high beta capability, but the realization of a compact reactor requires the elimination of the central solenoid (CS). The possibility of using the lower hybrid wave (LHW) to ramp up the plasma current (I p) from zero to a high enough level required for fusion burn in ST is examined theoretically and experimentally. Excitation of a travelling fast wave (FW) by the combline antenna installed on TST-2 was confirmed by a finite element analysis, but efficient current drive requires excitation of the LHW, either directly by the antenna or by mode conversion from the FW. The analysis using the TORLH full-wave solver indicates that core current drive by LHW is possible in the low-density, low I p plasma formed by electron cyclotron heating (ECH). It is important to keep the density low during I p ramp-up, and the wavenumber must be reduced as I p increases. Initial results from TST-2 demonstrate that RF power in the LH frequency range (200 MHz) can achieve initial I p formation, and is more effective than ECH for further ramp-up of I p. I p ramp-up to over 12 kA was achieved by combining ramp-up of the externally applied vertical magnetic field and ramp-up of the RF power. The significant asymmetry observed between co-current drive and counter-current drive is attributed to the presence of RF driven current. An optimized LHW antenna with appropriate polarization and wavenumber spectrum controllability is being designed. The success of the TST-2 experiment would provide a scientific basis for quantitatively evaluating the required CS capability for a low-aspect-ratio reactor.

Non-inductive plasma current start-up by EC and RF power was carried out on the TST-2 device. Low frequency RF (21 MHz) sustainment was demonstrated, and the obtained high βp spherical tokamak configuration has similar equilibrium values as the EC (2.45 GHz) sustained plasma. Equilibrium analysis revealed detailed information on three discharge phases: (i) in the initial current formation phase, the plasma current increases with the stored energy, and the current is in the same order as that predicted by theory. (ii) In the current jump phase, the current density profile, which is peaked near the outboard boundary, is not deformed but increases slowly and the initial closed flux surface appears when the current reaches a maximum. (iii) In the current sustained phase, equilibrium is characterized by the hollowness of the current density profile, and it determines the fraction of the current inside the last closed flux surface to the total current. Both EC and RF injections show a similar equilibrium. While MHD instabilities often terminate the RF sustained plasma, no such phenomenon was observed in the EC sustained plasma.