Search

Electron Cyclotron Resonance Heating (ECRH) has contributed to the achievement of high performance plasma production, high electron temperature plasmas and sustainment of steady-state plasmas in the Large Helical Device (LHD). Our immediate targets of upgrading the ECRH system are 5 MW several seconds and 1 MW longer than one hour power injection into LHD. The improvement will greatly extend the plasma parameter regime. For that purpose, we have been promoting the development and installation of 77 GHz/1-1.5 MW/several seconds and 0.3 MW/CW gyrotrons in collaboration with University of Tsukuba. The transmission lines are re-examined and improved for high and CW power transmission. In the recent experimental campaign, two 77 GHz gyrotrons were operated. One more gyrotron, which was designed for 1.5 MW/2 s output, was constructed and is tested. We have been promoting to improve total ECRH efficiency for efficient gyrotron-power use and efficient plasma heating, e.g. a new waveguide alignment method and mode-content analysis and the feedback control of the injection polarization. In the last experimental campaign, the 77 GHz gyrotrons were used in combination with the existing 84 GHz range and 168 GHz gyrotrons. Multi-frequency ECRH system is more flexible in plasma heating experiments and diagnostics. A lot of experiments have been performed in relation to high electron temperature plasmas by realization of the core electron-root confinement (CERC), electron cyclotron current drive (ECCD), Electron Bernstein Wave heating, and steady-state plasma sustainment. Some of the experimental results are briefly described. © 2009 American Institute of Physics.

Production and heating of over-dense plasmas by 2.45 GHz microwave system under very low field condition was performed and demonstrated on CHS. In this experiment, microwave systems were arranged to aim at taking place mode conversion of launched electron cyclotron wave effectively into electron Bernstein wave (EBW). The power deposition profiles in produced over-dense plasmas were measured directly by using power modulation technique at various magnetic configurations. The power deposition took place dominantly in overdense region. To clarify wave trajectories, power absorption mechanism and mode conversion, a numerical analysis by a ray-tracing method was performed. Calculated ray-trajectories accessed to the region where the experimentally obtained deposition profile has a peak (© 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

Achieving steady-state plasma operation at high plasma temperatures is one of the important goals of worldwide magnetic fusion research. High temperatures of approximately 1–2 keV, and steady-state plasma sustainment operations have been reported. Recently the steady-state operation regime was greatly extended in the Large Helical Device (LHD). A high-temperature plasma was created and maintained for 54 min with 1.6 GJ in the 2005FY experimental programme. The three-dimensional heat-deposition profile of the LHD helical divertor was modified, and during long-pulse discharges it effectively dispersed the heat load using a magnetic axis swing technique developed at the LHD. A sweep of only 3 cm in the major radius of the magnetic axis position (less than 1% of the major radius of the LHD) was enough to disperse the divertor heat load. The steady-state plasma was heated and sustained mainly by hydrogen minority ion heating using ion cyclotron range of frequencies and partially by electron cyclotron of fundamental resonance frequency. By accumulating the small flux of charge-exchanged neutral particles during the long-pulse operation, a high energy ion tail which extended up to 1.6 MeV was observed. This is the first experimental evidence of high energetic ion confinement of MeV range in helical devices. The long-pulse operations lasted until a sudden increase in radiation loss occurred, presumably because of metal wall flakes dropping into the plasma. The sustained line-averaged electron density and temperature were approximately 0.8 × 1019 m−3and 2 keV, respectively, at a 1.3 GJ discharge (#53776) and 0.4 × 1019 m−3and 1 keV at a 1.6 GJ discharge (#66053). The average input power was 680 kW and 490 kW, and the plasma duration was 32 min and 54 min, respectively. These successful long operations show that the heliotron configuration has a high potential as a steady-state fusion reactor.

Steady state divertor operation with high performance plasmas (ne ~ 0.7 × 1019 cm−3, Ti ~ 2 keV) was demonstrated for half an hour in the Large Helical Device (LHD), the superconducting helical device (R = 3.6–3.9 m, a = 0.6 m, B = 3 T, l/m = 2/10). The high performance plasmas have been sustained with an averaged heating power of 680 kW and achieved an injected energy of 1.3 GJ. This required both advanced technological integration of heating systems and divertor heat flux control. In particular, optimization of divertor heat flux distribution along the divertor leg trace on divertor plates and real-time magnetic axis sweeping (R = 3.67–3.7 m) have allowed LHD to access a steady state regime with a margin of safety for the actively cooled divertor plates. The distribution of divertor heat load along the traces was investigated with calorimetric measurements and it was found that there was a localized heat load connected with the loss of high-energy ions produced by ion cyclotron radio frequency near-fields. Orbit analysis shows that the behaviour of high-energy ions is qualitatively in good agreement with the experimental result. Long-pulse discharges were terminated by radiation collapse due to penetration of metallic flakes into the plasma.

A long-pulse plasma discharge of more than 30 min duration was achieved on the Large Helical Device (LHD). A plasma of ne = 0.8 × 1019 m−3 and Ti0 = 2.0 keV was sustained with PICH = 0.52 MW, PECH = 0.1 MW and averaged PNBI = 0.067 MW. The total injected heating energy was 1.3 GJ. One of the keys to the success of the experiment was a dispersion of the local plasma heat load to divertors, accomplished by sweeping the magnetic axis inward and outward. Causes limiting the long pulse plasma discharge are discussed. An ion impurity penetration limited further long-pulse discharge in the 8th experimental campaign (2004).

The Large Helical Device (LHD) achieves high-performance plasma confinement by the coordination of the magnetic surface region and the chaotic field line layer. It is theoretically and experimentally shown that drift surfaces exist for highly energetic particles being extended over the last closed flux surface (LCFS) in the LHD. These particles are considered lost particles due to the loss-cone in the previous theories, where the analyses are limited inside the LCFS. The present theory predicts that the loss-cone is strongly reduced in the LHD and that highly energetic particles confined over the LCFS exist. These are consistent with the LHD experimental results in both the ICRF heating experiments and the low magnetic field neutral beam injection heating experiments. From particle orbit analyses and studies on the connection length of diverter field lines, it is also shown that plasma can exist in the chaotic field line layer located outside the LCFS in the LHD. The plasma in the chaotic field line layer is clearly detected by CCD-cameras in the LHD experiment. This ambient plasma might be expected to play the role of a kind of impregnable barrier for the core plasma, which suppresses both the MHD instabilities and the cooling of the core plasma due to charge exchange processes. The line-tying effects of diverter field lines that are slipped out from the chaotic field line layer can also stabilize the ballooning mode and the vertical displacement events of the plasma column.

A long-pulse plasma discharge for more than 30 min. was achieved on the Large Helical Device (LHD). A plasma of ne = 0.8× 1019 m−3 and Ti0 = 2.0 keV was sustained with PICH = 0.52 MW, PECH = 0.1 MW and averaged PNBI = 0.067 MW. Total injected heating energy was 1.3 GJ, which was a quarter of the prepared RF heating energy. One of the keys to the success of the experiment was a dispersion of the local plasma heat load to divertors, accomplished by shifting the magnetic axis inward and outward.

Two types of improved core confinement were observed during centrally focused electron cyclotron heating (ECH) into plasmas sustained by counter (CNTR) and Co neutral beam injections (NBI) in the Large Helical Device. The CNTR NBI plasma displayed transition phenomena to the high-electron-temperature state and had a clear electron internal transport barrier, while the Co NBI plasma did not show a clear transition or an ECH power threshold but showed broad high temperature profiles with moderate temperature gradient. This indicated that the Co NBI plasma with additional ECH also had an improved core confinement. The electron heat transport characteristics of these plasmas were directly investigated using heat pulse propagation excited by modulated ECH. These effects appear to be related to the m/n = 2/1 rational surface or the island induced by NBI beam-driven current.

This paper reports on the results of an experimental investigation into electron cyclotron resonance heating (ECRH) using the fundamental O-mode in the large helical device (LHD). The aim of the experiment is to understand the mechanism of the selective excitation of the O-mode EC wave and, thereby, optimize the performance of the ECRH. The polarization angle and the ellipticity of the electric field of the incident high power millimetre waves were varied in the experiment using a pair of grating polarizers. The responses of the plasma stored energy and electron temperature were measured to evaluate the effects. The optimum values of the polarization angle and the ellipticity for the ECRH were identified experimentally. A simple model of power branching between the O- and X-modes is introduced in order to interpret the observed experimental results. Specific to the ECRH in the LHD configuration, the effect of magnetic shear in the plasma peripheral region is considered.

The potential of continuous plasma sustainment of Large Helical Device (LHD) was successfully demonstrated by a one hour discharge with 110 kW electron cyclotron heating (ECH) power. The ECH power of frequency 84 GHz generated by a continuous-work (CW) gyrotron is transmitted through an evacuated waveguide transmission line and injected to the LHD vacuum vessel using waveguide antenna. The plasma density was kept at about 1.5 × 1018/m3 and the electron temperature at the plasma center over 1 keV. Due to the low injection power the density was not so high but the plasma was quite stable. The power injection was terminated manually at 3900 seconds from the limitation of the setting of data acquisition, not for any troubles on devices.

The behaviour of helical ripple-trapped particles in large helical device (LHD) have been investigated by two line-of-sights natural diamond detector measurements in plasma sustained with ICRF for two types of resonance configurations, i.e. on-axis and saddle-shaped resonances. The tail temperature increases linearly as the Stix-temperature increases and then saturation follows, indicating the effect of the finite confinement time of the energetic ions. In the linear region, the tail readily increases when the line-of-sight is focused on the larger resonance area. The confinement time of the energetic ions is obtained from the empirical relation based on the Stix theory. The configuration with a saddle-shaped resonance has a longer confinement time than the on-axis resonance. Analyses of experimental measurements are consistent with the theoretical calculations including collisional ripple-induced transport.

Characteristics of MHD instabilities and their impacts on plasma confinement are studied in current free plasmas of the Large Helical Device. Spontaneous L?H transition is often observed in high beta plasmas close to 2% at low toroidal fields (Bt ? 0.75?T). The stored energy starts to rise rapidly just after the transition accompanying the clear rise in the electron density but quickly saturates due to the growth of the m = 2/n = 3 mode (m and n: poloidal and toroidal mode numbers), the rational surface of which is located in the edge barrier region, and edge localized mode (ELM) like activities having fairly small amplitude but high repetition frequency. Even in low beta plasmas without L?H transitions, ELM-like activities are sometimes induced in high performance plasmas with a steep edge pressure gradient and transiently reduce the stored energy up to 10%. Energetic ion driven MHD modes such as Alfv?n eigenmodes (AEs) are studied in a very wide range of characteristic parameters (the averaged beta of energetic ions, ?b?, and the ratio of energetic ion velocity to the Alfv?n velocity, Vb?/VA), of which range includes all tokamak data. In addition to the observation of toroidicity induced AEs (TAEs), coherent magnetic fluctuations of helicity induced AEs (HAEs) have been detected for the first time in NBI heated plasmas. The transition of a core-localized TAE to a global AE (GAE) is also observed in a discharge with temporal evolution of the rotational transform profile, having a similarity to the phenomenon observed in a reversed shear tokamak. At low magnetic fields, bursting TAEs transiently induce a significant loss of energetic ions, up to 40% of injected beams, but on the other hand play an important role in triggering the formation of transport barriers in the core and edge regions.

Significant progress has been made with ion cyclotron range of frequencies (ICRF) heating in the Large Helical Device. This is mainly due to better confinement of the helically trapped particles and less accumulation of impurities in the region of the plasma core. During the past two years, ICRF heating power has been increased from 1.35 to 2.7 MW. Various wave-mode tests were carried out using minority-ion heating, second-harmonic heating, slow-wave heating and high-density fast-wave heating at the fundamental cyclotron frequency. This fundamental heating mode extended the plasma density range of effective ICRF heating to a value of 1×1020 m−3. This use of the heating mode was its first successful application in large fusion devices. Using the minority-ion mode gave the best performance, and the stored energy reached 240 kJ using ICRF alone. This was obtained for the inward-shifted magnetic axis configuration. The improvement associated with the axis-shift was common for both bulk plasma and highly accelerated particles. For the minority-ion mode, high-energy ions up to 500 keV were observed by concentrating the heating power near the plasma axis. The confinement properties of high-energy particles were studied for different magnetic axis configurations, using the power-modulation technique. It confirmed that with the inward-shifted configuration the confinement of high-energy particles was better than with the normal configuration. By increasing the distance of the plasma to the vessel wall to about 2 cm, the impurity influx was sufficiently reduced to allow sustainment of the plasma with ICRF heating alone for more than 2 min.

Internal transport barriers with respect to electron thermal transport (eITB) were observed in the large helical device, when the electron cyclotron resonance heating (ECH) power was highly localized on the centre of a plasma sustained by neutral beam injection. The eITB is characterized by a high central electron temperature of 6–8 keV with an extremely steep gradient, as high as 55 keV m−1 and a low electron thermal diffusivity within a normalized average radius ρ≈0.3 as well as by the existence of clear thresholds for the ECH power and plasma collisionality.

The behaviour of high-energy ions accelerated by an ion cyclotron range of frequency (ICRF) electric field in the large helical device (LHD) is discussed. A better confinement performance of high-energy ions in the inward-shifted magnetic axis configuration was experimentally verified by measuring their energy spectrum and comparing it with the effective temperature determined by an electron slowing down process. In the standard magnetic axis configuration a saturation of the measured tail temperature was observed as the effective temperature was increased. The ratio between these two quantities is a measure of the quality of transfer efficiency from high-energy ions to a bulk plasma; when this efficiency was compared with Monte Carlo simulations the results agreed fairly well. The ratio of the stored energy of the high-energy ions to that of the bulk plasma was measured using an ICRF heating power modulation method; it was deduced from phase differences between total and bulk plasma stored energies and the modulated ICRF heating power. The measured high energy fraction agreed with that calculated using the injected ICRF heating power, the transfer efficiency determined in the experiment and the confinement scaling of the LHD plasma.

Ion temperature at the plasma centre has been measured from Doppler broadening of Ti XXI (2.61 A) and Ar XVII (3.95 A) x-ray lines using a newly installed crystal spectrometer with CCD detector in ECH, NBI and ICRF plasmas of Large Helical Device (LHD). The ion temperature obtained in a range of 0.6 and 3.5 keV was analysed with electron density and compared with electron temperature. A new parameter range of Ti>Te was found in low-density (ne

Intrinsic impurity behaviour in neutral beam heated LHD long pulse discharges with a discharge duration of 10 s was investigated. Spectroscopic and bolometric measurements showed a remarkable temporal increase of core radiation due to metallic impurities. Central impurity accumulation was found in a narrow plasma density region (e = 1–3×1019 m−3) for both density ramp-up discharges and constant density discharges. In the density ramp-up discharges, high-Z impurities were accumulated in the low-density region and diffused out from the core plasma in the high-density region. The impurity behaviour was compared with neoclassical predictions.

Energetic tail formation by ion-cyclotron-range-of-frequencies (ICRF) heating was studied in the large helical device (LHD), by measuring high-energy charge-exchange atom spectra with a specially developed natural diamond detector, which views the LHD plasma perpendicularly. Comparison of measured effective perpendicular temperature of ICRF-driven H+ minority ions with the classical Stix model indicates that perpendicular ions with energy at least up to 150 keV are well confined in the centre region of the LHD. However, deviation from the classical prediction was observed for high-energy perpendicular protons in the outer region of the plasma, indicating ripple-induced transport and charge-exchange losses. With perpendicular fast atom spectrum and flux measurements, small yet notable differences were detected co-injected and counter-injected ICRF-driven beam ion confinement during combined neutral beam injection (NBI) and ICRF heating. These results can be explained by the differences in the orbit topology and ICRF-induced drift.

The experimentally optimized magnetic field configuration of the Large Helical Device [A. Iiyoshi et al., Nucl. Fusion 39, 1245 (1999)], where the magnetic axis is shifted inward by 15 cm from the early theoretical prediction, reveals 50% better global energy confinement than the prediction of the scaling law. This configuration has been investigated further from the viewpoints of high energy particle confinement and magnetohydrodynamic (MHD) stability. The confinement of high energy ions is improved as expected. The minority heating of ion cyclotron range of frequency was successful and the heating efficiency was improved by the inward shift. The confinement of passing particles by neutral beam injection was also improved under low magnetic field strength, and there could be obtained an almost steady high beta discharge up to 3% in volume average. This was a surprising result because the observed pressure gradient exceeded the Mercier unstable limit. The observed MHD activities became as high as beta but they did not grow enough to deteriorate the confinement of high energy ions or the performance of the bulk plasma, which was still 50% better than the scaling. According to these favorable results, better performance would be expected by increasing the heating power because the neoclassical transport can also be improved there.

Enhancement of the output power per gyrotron has been planned in the Large Helical Device (LHD). Three 77-GHz gyrotrons with an output power of more than 1 MW have been operated. In addition, a high power gyrotron with the frequency of 154 GHz (1 MW/5 s, 0.5 MW/CW) was newly installed in 2012 and the total injection power of ECRH reached 4.6 MW. The operational regime of ECRH plasma on the LHD has been extended due to the upgraded ECRH system such as the central electron temperature T{sub e0} = 13.5 keV with n{sub e} = 1×10{sup 19}m{sup −3}. In the LHD, an electron-internal-transport barrier (e-ITB) related to the production of high T{sub e} plasmas has been realized by strongly centre-focused ECRH. The electron thermal confinement clearly improved inside the e-ITB. The radial electric field was measured using the heavy ion beam probe. The formation of the positive E{sub r} was observed in the core region, which well agreed with the prediction of the neoclassical transport theory. The energy confinement characteristics have been investigated in the ECRH plasmas. It was found that higher plasma stored energy and lower radiation power was realized in the outward configuration. The plasma stored energy of 530more » kJ with n{sub e} = 3.2×10{sup 19}m{sup −3}, which is the 1.7 times larger than the previous record in the ECRH plasma in the LHD, has been successfully achieved.« less

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

The large helical device (LHD) has fourteen experiment cycles so far. The ion cyclotron range of frequencies (ICRF) heating has been carried out successfully on LHD; High RF power heating up to 3MW and the long pulse plasma discharge up to about one hour (54 minutes) have been achieved using the poloidal array (PA) antennas. In the 14th experiment cycle (2010) a new one pair of ICRF heating antennas was installed in LHD. This pair consists of two antennas arrayed in the toroidal direction and has a controllability of the wave number along the magnetic field line k∥ with changing the phase difference between them. The target plasma consisted of the helium plasma as the majority with the hydrogen‐ion minority. Several experiment results were obtained and compared with those obtained using the PA antenna (no capability of changing k∥). The fast wave with a large k∥ is excited with (0, π) phasing between two antennas, and a small wave‐number of k∥ with (0, 0) phasing. The plasma loading resistance in (0, π) w...

In LHD, an ECH system for fusion-relevant plasma heating has been greatly progressed with the development of megawatt gyrotrons. Net power of 3.7 MW could be injected into LHD. For stable operation, a new position and profile monitor of the mmw-beams in evacuated transmission lines is developed. © 2010 IEEE.

Production of over-dense plasmas by launching 2.45 GHz microwaves under very low field conditions has been successfully demonstrated on CHS.1 In the experiment, microwaves systems were arranged to aim at taking place mode conversion of launched electron cyclotron wave effectively into electron Bernstein wave (EBW). One system (ECH#1) is launched nearly perpendicularly to the toroidal field, where FX-B conversion scenario is expected. Another (ECH#2) is launched obliquely for O-X-B conversion scenario. This experiment aims at clarifying the excitation, propagate and absorption of mode converted EBW. Power deposition profiles in produced over-dense plasmas were measured directly by using power modulation technique at various magnetic configurations where electron cyclotron resonance (ECR) and upper hybrid resonance (UHR) layers were scanned in space widely.2 Figure 1 is radial profile of electron temperature T e , electron density n e and electron pressure p e . This figure also shows the response in p e to ECH#1 or ECH#2 source p e , and the coherence γp e 2 and phase difference between δ p e and the modulated ECH power Φ pe at B t /B res = 100%. Electron density exceeds 2–3 times O-mode cutoff density (n co ∼ 7.5×1016m−3). Power deposition profile corresponds to Φ pe -profile and the peak position is in over-dense region. Similarly, it was investigated in the cases of B t /B res = 70%, 50% and 20%.

Parametric scaling studies of radiofrequency and neutral beam heated currentless Heliotron E plasmas have been performed. The parametric local electron transport analyses show that the electron energy transport in electron cyclotron heating (ECH) plasmas is nearly of the same magnitude as the neoclassically predicted transport inside the 2/3 radius, while neutral beam injection (NBI) plasmas and plasmas in the ion cyclotron range of frequencies (ICRF) are dominated by anomalous electron transport in the entire region. Scaling studies on the global energy confinement time reveal that ECH, NBI and ICRF plasmas obey approximately identical scalings that are characterized by continuous power degradation and favourable positive density dependence. The global energy confinement time is thought to be affected by the anomalous transport in the peripheral plasma regions in the same way for ECH, NBI and ICRF plasmas although the core plasma properties – such as local electron transport – seem to be different for the individual plasmas.

A kinetic analysis of the global energy confinement time for neutral beam heated Heliotron E plasmas has been performed with a 1-D, time independent transport analysis code, PROCTR-Mod. From a regression analysis of a representative sample selection of the presented data sets, the global energy confinement time, , is found to scale as , where α = 0.53 ± 0.10, β = −0.71 ± 0.09, and γ = 0.35 ± 0.14; the error bar indicates the 95% confidence region deduced from the limited data points; is the line average electron density; Pheat[MW] is the neutral beam heating power; and B[T] is the vacuum magnetic field at the magnetic axis. The data analysis shows that the favourable density dependence partially offsets the unfavourable power dependence and that anomalous electron transport loss becomes dominant in the overall energy balance as the beam power and the plasma density are increased. An alternative scaling law is also presented, which is to fit by an 'offset linear' law, , where ζ = 1.1 ± 0.33 and η = 15.0 ± 1.9; ne[1014 cm−3] is the volume average electron density. The latter scaling is found to provide a better fit to the presented data sets, in spite of its simple form. The parametric scaling of the local electron thermal diffusivity, χe is also discussed on the basis of the kinetic analysis.

There are several important roles for the electron cyclotron resonance heating (ECRH) methods in helical-type plasma experimental devices. One is to produce high-temperature and high-density current-free plasmas. Another is additional heating of electrons and resulting ion heating through equipartition. Plasma temperature and current profile controls with highly localized heating become advanced techniques to improve plasma confinement and to suppress some MHD instabilities. ECRH is also an ultimate candidate of a steady-state plasma sustainment.

Demonstration of a long pulse or a steady state operation of ECH and sustainment of non‐collapsed plasma only by ECH is important in LHD from both the confinement device and the heating system engineering points of view. A gyrotron with a diamond output window is introduced and operated at the power level of 150 kW for more than 1 hour after modification of the cooling and evacuation system of the ECH transmission line. The power of about 110 kW injected into LHD is used to sustain the plasma with the electron density of 1.5 × 1018 m−3 and central temperature of more than 1.0 keV for 3900 sec. The gas puffing rate is carefully controlled so that the plasma density does not exceed the critical value above which the plasma collapsed for given injection power, magnetic configuration and wall condition of LHD. The results of gyrotron operation, transmission system modification for long pulse and optimizations of the magnetic field configuration of LHD and gas puffing for a given injection condition are discussed.

The confinement of high‐energy particles produced by ICRF heating was studied in LHD experimentally setting magnetic axis at 3.6m and 3.75m in two methods. It was concluded that the confinement of high‐energy particles with Rax=3.6 m was better than that with Rax=3.75m. Based on this result the long pulse experiments were conducted in the magnetic configuration of Rax=3.6m. The plasma was sustained for 150sec. The total input power reached 71MJ though the pulse length was limited by the collapse of plasma due to density increase. The second harmonic experiment was done also with Rax=3.6m. The stored energy increased and the particles with the high‐energy were observed.

This paper gives an overview of new ICRF techniques developed on HT-7 superconducting tokamak during 1998 campaign, using the present two simple loop antennas, for wall conditioning/boron film deposition of the vessel and start-up of OH discharges: (1) The results show that He-plasma is reliably produced at the standard ICRH frequency (f=22.5 MHz), in a wide range of toroidal field, gas pressure and different RF power level. ICRF cleaning efficiency is about three times higher than that of glow discharge cleaning, this new technique has been used daily during experiments. (2) Start-up experimental results show that RF power was very effective to make the breakdown time less than 2 ms and helpful to assist the plasma start-up when the ICRF power reached 50 kW level. About 10%–20% reduction in the volt-second consumption was achieved according to different condition, evaluated at the beginning of the current flattop. (3) Boronization is carried out by using IBW antenna, the film depth is about 350 nm, and i...

The final goal of Ion Cyclotron Range of Frequency (ICRF) heating on the Large Helical Device (LHD) is characterized by its high power (up to 12MW) and by steady state operation (30 minutes). Initial ICRF heating experiments were carried out using a pair of loop antemas in the 2nd experimental campaign in 1998. The ICRF heating power was applied to an ECH-produced plasma at an RF power level of 300 kW for 0.2 seconds. An applied frequency of f=25.6 MHz was selected. A cyclotron resonance layer of hydrogen ions was located at the half minor radius during operation at B=1.5 T. A mode conversion layer of a He plasma with a minority of hydrogen ions was located between the magnetic axis and the last closed magnetic flux. The plasma stored energy was observed to increase to twice that of the ECH plasma (PECH=300 kW). The plasma stored energy of the ECH target plasma was 11-13 kJ at an average electron density of ne=8–9×1018 m−3 and a central electron temperature of Te0=400 eV. The plasma stored energy increase...

The steady-state operation of high-performance plasmas in the Large Helical Device (LHD) has progressed since the 2010 IAEA Conference in Korea by means of a newly installed ion cyclotron heating (ICH) antenna (HAS antenna) and an improved electron cyclotron heating (ECH) system. The HAS antenna can control the launched parallel wave number and heat the core plasma efficiently in the case of dipole mode operation. Understanding of the physics and technology of wave heating, particle and heat flow balances, and plasma?wall interactions in LHD has also improved. The heating power of steady-state ICH and ECH exceeded 1?MW and 500?kW, respectively, and a higher density helium plasma with minority hydrogen ions was maintained using the HAS antenna and new 77?GHz gyrotrons. As a result, plasma performance improved, e.g. electron temperature of more than 2?keV at a density of more than 2???1019?m?3 became possible for more than 1?min. Heat flow balance and particle flux balance of steady-state operation were evaluated. Particle balance analysis indicated that externally fed helium and hydrogen particles were mainly absorbed by the chamber wall and divertor plates, even after the 54?min operation.

We will describe the capabilities of a folded waveguide (FWG) antenna and its application to ion Bernstein wave (IBW) heating in the Large Helical Device (LHD). LHD is scheduled to begin plasma operation in 1998. We are currently developing a FWG antenna to launch IBW to both produce and sustain the LHD plasma. We have fabricated a mockup FWG antenna to obtain the fundamental RF characteristics. The mockup FWG can produce various wave numbers parallel to the magnetic field line with either polarization plates or shorting bars in the antenna aperture. Preliminary results show a sharply defined wave number spectrum along the magnetic field line, which could provide efficient core electron heating. This type of direct electron heating is superior to fast wave minority heating, because it can avoid fast orbit losses of deeply trapped high energy ions.

Ion cyclotron range of frequency (ICRF) heating is currently in use on CHS and W7‐AS and is a major element of the heating planned for steady state helical devices. In helical devices, the lack of a toroidal current eliminates both disruptions and the need for ICRF current drive, simplifying the design of antenna structures as compared to tokamak applications. However the survivability of plasma facing components and steady state cooling issues are directly applicable to tokamak devices. Results from LHD steady state experiments should be available on a time scale to strongly influence the next generation of steady state tokamak experiments. The helical plasma geometry provides challenges not faced with tokamak ICRF heating, including the potential for enhanced fast ion losses, impurity accumulation, limited access for antenna structures, and open magnetic field lines in the plasma edge. The present results and near term plans provide the basis for steady state ICRF heating of larger helical devices. An a...

Since the successful first plasma generation in the middle of 2008, three experimental campaigns were successfully made for the KSTAR device, accompanied with a necessary upgrade in the power supply, heating, wall-conditioning and diagnostic systems. KSTAR was operated with the toroidal magnetic field up to 3.6 T and the circular and shaped plasmas with current up to 700 kA and pulse length of 7 s, have been achieved with limited capacity of PF magnet power supplies. The mission of the KSTAR experimental program is to achieve steady-state operations with high performance plasmas relevant to ITER and future reactors. The first phase (2008–2012) of operation of KSTAR is dedicated to the development of operational capabilities for a super-conducting device with relatively short pulse. Development of start-up scenario for a super-conducting tokamak and the understanding of magnetic field errors on start-up are one of the important issues to be resolved. Some specific operation techniques for a super-conducting device are also developed and tested. The second harmonic pre-ionization with 84 and 110 GHz gyrotrons is an example. Various parameters have been scanned to optimize the pre-ionization. Another example is the ICRF wall conditioning (ICWC), which was routinely applied during the shot to shot interval. The plasma operation window has been extended in terms of plasma beta and stability boundary. The achievement of high confinement mode was made in the last campaign with the first neutral beam injector and good wall conditioning. Plasma control has been applied in shape and position control and now a preliminary kinetic control scheme is being applied including plasma current and density. Advanced control schemes will be developed and tested in future operations including active profiles, heating and current drives and control coil-driven magnetic perturbation.

Ion cyclotron range of frequencies (ICRF) heating antennas in LHD are numerically simulated and analyzed by HFSSTM finite-element electromagnetic wave field calculation code. The model includes an accurate vacuum chamber wall of LHD and ICRF antenna structure and a simple model of plasma in a helical configuration. Antenna coupling with plasma is simulated by an artificial freshwater volume with enhanced high permittivity of e = 500-2000. RF current distribution and electromagnetic field distribution on and near the ICRF antenna are analyzed and well elucidated through a comparison with the experimental results. The frequency dependence of experimental loading resistance can be simulated by the calculation, and the RF dissipation on the antenna structure is studied and compared with experimental results. The local high heat load around gaps between the carbon side protectors is well explained, and the effect of gap distance is studied. Comparison with the experimental results reveals that the ICRF heating in the LHD, including the antenna and helical plasma, is well simulated by commercial HFSSTM code analysis. It will also be useful for future improvements in ICRF antenna design in helical devices.

A transition to a high electron temperature state in the core accompanied by an electron internal transport barrier (eITB) was clearly observed in neutral beam (NB) sustained plasmas with strongly focused on-axis electron cyclotron heating (ECH) in Large Helical Device (LHD). During stepwise injection of ECH power, the core electron temperature started to build up spontaneously and became high temperature in the core (high-Te state), while the temperature outside of ρ 0.3 drastically decreased. The transition dynamics of the core temperature is analyzed by the time evolution of electron temperature profiles and heat flux changes based on the data from multi-channel electron cyclotron emission. Effect of an initial island size (m/n=2/1) on the high-Te transition was experimentally investigated by controlling the width by an external coil field. The results show that the island size possibly affects a threshold value of ECH power and collisionality on the high-Te transition.

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

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

Carbonization was successfully applied to Heliotron-E. Iron-inpurity radiations were strongly reduced with the carbonized wall. Main impacts of the metal reduction on plasma behaviors are sustainment of stored energy during high power, long pulse heating by NBI, achievement of a quasi-steady discharge with a low helical field and high beta, and highest electron density with pellet injection in a quasi-steady state. Hydrogen recycling was very high with the carbonized wall and low density operation was impossible. Helium glow discharge was found to be effective to control the hydrogen recycling with a carbon-tiled wall.

Electron-cyclotron and neutral-beam heating have been used to produce currentless plasma with $〈\ensuremath{\beta}〉\ensuremath{\approx}(1\ensuremath{-}2)%$ in the Heliotron-$E$ device. For moderately peaked profiles, magnetohydrodynamic activity that peaks at the $t=1$ surface has been observed. The characteristics of the fluctuations are consistent with theoretical expectations for an $m=1$ pressure-driven instability. The magnetohydrodynamic activity is associated with particle and energy losses but has not yet been directly linked to global confinement.

Recent results of the heating experiment on the Heliotron E plasma are reported. Two heating methods are applied; electron cyclotron resonance heating (ECRH) and neutral injection (NBI) heating. In the ECRH experiment, a currentless high-temperature plasma is produced with the use of a gyrotron capable of delivering an output power of 200 kW for 40-ms duration at 28 GHz. The electron and the ion temperatures at the center −Te(0) and Ti(0) − 1.1 keV and 120 eV, respectively, and the average density n e of 5 × 1012 cm−3 is obtained. The NBI heating has produced the plasma which has the temperatures of Te(0) ∼ Te(0) ∼ 660 eV and the density n e of 3 × 1013 by the NBI power of about 1.6 MW. The heating efficiency has been found to be about 2 ( eV ·P abs ( kW ) n e (10 13 cm −3 ), where P abs is the absorbed neutral beam power, for bot perpendicular and 28° co- and counter-injection angles from the normal direction of the torus. The central beta value reaches about 1%.

Particle control with pump limiters has been successfully demonstrated in a variety of tokamaks. This experiment has obtained, for the first time, experimental data on pump limiter operation in a heliotron configuration. A movable pump limiter module was installed on a horizontal midplane port of Heliotron-E. The limiter module was installed on a horizontal midplane port of Heliotron-E. The limiter assembly consists of a TiC-coated graphite head with single-sided particle collection and active pumping. The location of the limiter is varied from near the vacuum vessel wall or up to 8 cm inside the last closed magnetic flux surface. This flexibility permits the study of a heliotron plasma that is limited either by the magnetic separatrix or by a material limiter. In the configuration investigated, only very low pressures are observed in the pump limiter when it is located outside the last closed flux surface, indicating that the scrape-off layer density is very low ( 11 cm −3 ). For limiter positions inside the last closed flux surface, pressures of 2–6 mTorr are observed (similar to comparable tokamak operation). Erosion patterns on the vacuum vessel, as well as hot spots on the limiter head, indicate that the particle flow out of the confined plasma exhibits complicated patterns. This suggests that for efficient particle collection in a stellarator, pump limiters must be matched to the particle flow patterns.

Shear Alfven wave heating experiments are reported in which a bulk electron heating of the toroidal plasma for a high initial temperature plasma ( T e >50 eV) and a local electron heating for a lower temperature plasma were observed. The results indicate the excitation of the kinetic shear Alfven wave and are consistent with the theoretical calculation of the damping length of the wave.

Parametric studies of volume averaged electron temperature (Te) and global electron energy confinement time ?Ee of ohmically heated Heliotron E plasmas have been performed using a data acquisition computer system. The scaling laws (Te) ? (IOH · B/ne)0.5 and ?Ee ?ne-0.4 · B/IOH1.5 are obtained directly by a code which fits the exponents of the plasma parameters to the electron temperature and confinement time. The ohmically heated plasma confinement time ?Ee is shown to be related to the drift parameter ? (= VDe/VTe). The dependences of the energy confinement time on other plasma parameters is also presented. An investigation is made of the correlation between MHD activity and the confinement.

Measurements of neutral density in ECRH, Ohmically heated and neutral beam heated plasmas on the Heliotron E device are described. Profiles of neutral density determined by the vertically scanned charge exchange neutral flux are compared with the results of the Monte Carlo simulation. In Ohmic and ECRH plasmas, the obtained profiles agree well with the simulation results using Franck-Condon neutral (∼3 eV) as incident particles. In NBI case, the recycling particle with higher energy (typically ∼30 eV) dominates the particle source. The dependence of the neutral density on the average electron density is represented by the parameter I p . Although the neutral density in Ohmic and ECRH plasmas decreases as \bar N e increases, NBI plasma does not have such dependence. The global particle confinement time is estimated, and τ p decreases as I p increases.

The transport of a currentless ECRH plasma in Heliotron E is studied numerically, using a 1-D transport code with an assumed RF power deposition profile. The investigated parameter range is as follows: 300 Te 1000 eV, Ti 120 eV, 2 < e< 10 X 1012 cm−3 and PRF ≈ 90 kW at B = I T. Four typical models for the transport coefficients are used in the numerical calculations. It is found that a neoclassical model results in a remarkable discrepancy of the time evolution of the electron temperature Te(r) and density ne(r) profiles with regard to the measured values. The neoclassical model which includes the effects of the helical field ripple seems to simulate the saturation of Te(0), in agreement with the experimental results. This saturation is due to the dependence of . However, a consistent description of Te(r) and ne(r) cannot be given by this model. In order to reproduce the measured profiles, it is necessary to use radially increasing transport coefficients. Adding Alcator-like anomalous transport terms (independent of Te) to the neoclassical transport model, it is found that e and PRF dependences of Te(0) can be adequately explained. The fourth model investigated has a temperature-dependent . For ions the e dependence of an anomaly factor, ,is also examined.

Measurements of atomic hydrogen density profiles in Heliotron E plasmas were performed, where attention was focussed on the anisotropy of the profiles in the poloidal cross section. Obtained penetration profiles of atomic hydrogen agreed well with the Monte Carlo code calculations of the penetration of Franck-Condon neutrals when averaged electron densities \bar n e are larger than 10 19 m -3 . For \bar n e

Plasma heating experiment is performed using a low-frequency shear Alfv\'en wave in the Heliotron-$D$ device. The excitation of the wave is confirmed, and an efficient plasma heating is observed; electrons are heated up to about 2 times and ions more than 1.5 times of the initial temperatures by applying an rf pulse of 400-kW power level.