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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.

The operation scenarios for plasma breakdown and current ramp-up phases in JT-60SA tokamak have been developed and verified in simulation using the TOSCA code. The induced current in the conducting elements such as vacuum vessel and stabilizing plate increases to the comparable level of plasma current of ∼600 kA during the breakdown phase and thus enhances the strength of stray field. The optimized scenarios for half and full pre-magnetization cases satisfied the conditions required for the plasma initiation. At the initial plasma, the vertical magnetic field required to sustain the plasma position was controlled by the outer equilibrium field (EF) coil currents which compensate for a vertical field due to a large eddy current. The condition for the formation of divertor configurations given by the combination of the magnetic flux for plasma and the plasma current has been satisfied which enables us to develop the operational scenarios with a smooth transition from a limiter to a divertor configuration.

The Monte-Carlo (MC) approach has a lot of flexibility in impurity transport modeling in the SOL and divertor region. However, in the divertor plasma simulation with the noble impurity seeding, characteristic time of the impurity transport especially in the sub-divertor chamber is long because the MC calculation of the impurity gas transport can be finished only by exhaust. The impurity MC calculation for such long exhaust processes is difficult in a series of the iterative calculation of a suite of integrated divertor codes SONIC. In order to overcome such a problem, a backflow model has been developed. Amount of the backflow flux from the sub-divertor chamber to the divertor region is evaluated in advance, and then simulating impurity flux is injected from the exhaust slot to the divertor region like a backflow. By this model, the MC calculation time is reduced significantly and iterative calculation of SONIC becomes possible within a reasonable calculation time. As a demonstration, the SONIC code with the backflow model has been applied to investigation of power handling in JT-60SA divertor. The SONIC simulation showed that low divertor heat load (< 10 MW/m2) with the low SOL density (< 1.5 × 1019m–3), which is required in the full non-inductive current drive scenario, was achieved by the Ar gas puffing of 0.86 Pa m3/s. (© 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

核融合原型炉において中心ソレノイド磁束消費を最小化しながら多様な安全係数分布のプラズマを得るために、電子温度分布の加熱フィードバックを通じた安全係数分布制御を検討した。その結果、正磁気シア、弱磁気シア、負磁気シアの幅広い範囲の安全係数分布を持つプラズマが経験則に基づく予測の60%程度の抵抗性磁束消費で生成できることが明らかになった。目標電子温度の最適化と加熱フィードバックゲインの最適化に、強化学習の手法を用いた。実効電荷分布がランダムに選ばれかつそれが直接観測できない状況でも、目標電子温度分布を適応的に最適化し、低磁束消費のq分布制御が実現できることを示した。

The integrated modelling code TOPICS has been extended to include the core impurity transport, and applied to predictive modelling of JT-60SA high-beta steady-state plasma with the accumulation of impurity seeded to reduce the divertor heat load. In the modelling, models and conditions are selected for a conservative prediction, which considers a lower bound of plasma performance with the maximum accumulation of impurity. The conservative prediction shows the compatibility of impurity seeding with the core plasma with high-beta (βN > 3.5) and full current drive conditions, i.e., when the Ar seeding reduces the divertor heat load below 10 MW/m2, its accumulation in the core is so moderate that the core plasma performance can be recovered by additional heating within the machine capability to compensate the Ar radiation. Due to the strong dependence of accumulation on the pedestal density gradient, the high separatrix density is important for the low accumulation as well as the low divertor heat load. The conservative prediction also shows that JT-60SA has enough capability to explore the divertor heat load control by impurity seeding in high-beta steady-state plasmas.

The application of the electron cyclotron heating (ECH) and electron cyclotron current drive (ECCD) to the JT-60U tokamak started in 1999. Because the power deposition by the electron cyclotron wave is very localized and controllable, the application ofECH/ECCD has been very attractive for the following recent studies in the JT-60U: (a) the extension of plasma performance toward high normalized beta (β N ), (b) high bootstrap current fraction, and (c) long-pulse operation (65 s). Plasma produced in the studies aiming at advanced steady tokamak is considered to be in a kind of "self-organized state" with external input power by joule heating plus additional heating. The internal transport barrier that develops by the additional heating enhances the local bootstrap current by steep pressure gradient, and then the modified plasma current profile establishes a different confining poloidal magnetic field configuration from the initial configuration. In such experimental research in the JT-60U, the ECH contributes as an active tool for the plasma control to study the physical mechanisms of high-IS magnetohydrodynamic instability, internal transport barrier, current hole, and so on. Results of the ECH/ ECCD applications in the JT-60U are briefly reviewed.

Impact of the electron cyclotron range of frequency wave (ECRF) on the internal transport barriers (ITBs) in a weak shear (WS) plasma has been investigated in JT-60U. The fundamental O-mode ECRF of 110 GHz injected obliquely (co-current drive) from the low field side is used. It is observed that the ion temperature (Ti) ITB in a WS plasma can be degraded by ECRF. It is clarified for the first time that the degradation depends increasingly on the EC power (PEC) but decreasingly on the plasma current (Ip). Moreover it is confirmed that ECRF affects the toroidal rotation (Vt) indirectly and results in the flattening of Vt(ρ) and therefore the radial electric field (Er) profiles regardless of the direction of the target Vt(ρ), peaking co or counter direction (relative to the Ip direction). Furthermore, it is recently found that Ti and Vt in the whole ITB region are affected with almost no delay from the EC onset even with off-axis EC deposition. These results indicate that EC injection unveiled a semi-global structure that characterizes Ti ITB in a WS plasma.

Characteristics of the edge pedestal has been studied for the high βp ELMy H mode and the reversed shear ELMy H mode in JT-60U. The contribution of the pedestal is around 20% for the confinement improvement factor, the normalized beta, and the bootstrap fraction. The upper boundary of the fusion performance measure increases with the pedestal poloidal beta, βp-ped. At high triangularity, δ, βp-ped increases with the total βp (βp − tot) almost linearly for positive shear type I ELMy discharges. This dependence is not due to the profile stiffness, since the dependence is the same for the discharges both with and without an internal transport barrier (ITB). In the reversed shear ELMy H mode, βp-ped increases with βp-tot on the same line as the positive shear cases except at high q95 = 8–9.3. In the high plasma current regime, the final structure with both an ITB and edge transport barrier seems to be determined by the balance between the expanding ITB-foot radius and deepening ELM penetration: the ELM penetration radius deepens with increasing pedestal stored energy and then reaches the ITB-foot radius. The ITB-foot seems to behave as a barrier against the ELM crash penetration, and shrinks after a few ELM attacks, and the ELM penetration follows the shrinking ITB-foot.

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

To contribute to high performance long pulse (∼65 s) experiments in JT-60U, the target of the electron cyclotron (EC) operation in long pulse is 0.6 MW for 30 s with four gyrotrons, though 10 MJ (2.8 MW and 3.6 s) was achieved in high power operation before 2003. One of the critical issues for the long pulse operation is detuning due to decay in beam current of the gyrotron. This decay comes from the cathode cooling by continuous electron emission. As a countermeasure for this issue, active adjustments for the heater current and anode voltage during the pulse have successfully extended the duration of a good oscillation condition for the gyrotron. As a result, 0.4 MW for 16 s with one gyrotron to the dummy load and for 8.7 s to the plasma have been achieved up to now.

Characteristics of post-disruption runaway electrons with impurity pellet injection were investigated for the first time using the JT-60U tokamak device. A clear deposition of impurity neon ice pellets was observed in a post-disruption runaway plasma. The pellet ablation was attributed to the energy deposition of relativistic runaway electrons in the pellet. A high normalized electron density was stably obtained with nebar/nGW ˜ 2.2. Effects of prompt exhaust of runaway electrons and reduction of runaway plasma current without large amplitude MHD activities were found. One possible explanation for the basic behavior of runaway plasma current is that it follows the balance of avalanche generation of runaway electrons and slowing down predicted by the Andersson-Helander model, including the combined effect of collisional pitch angle scattering and synchrotron radiation. Our results suggested that the impurity pellet injection reduced the energy of runaway electrons in a stepwise manner.

Stabilization of an m = 3/n = 2 neoclassical tearing mode (NTM) has been studied experimentally by applying a local electron cyclotron current drive (ECCD) in the JT-60U tokamak. The EC power is injected before the mode onset, and its effect is compared with the ECCD applied at the saturation phase. The experimental results show that the ECCD applied at the growth phase is more effective than that applied at the saturation phase. The necessary EC power for the suppression is reduced and the mode onset is delayed, indicating the hysterisis characteristics of the NTM on the ECCD stabilization. The dependence on the EC power and injection angle is also shown.

The `stiffness' of thermal transport in ELMy H-modes is examined in a series of carefully chosen JT-60U plasmas, and measured temperatures are compared with the predictions of several transport models. A heating power scan with constant Tped, a scan of pedestal temperature, Tped, with constant heating power, and an on-axis/off-axis heating comparison are presented. In the power scan, a 45% increase in heating (and a 12% density rise) produces an approximately fixed core temperature profile in a group of five plasmas with the same pedestal temperature. With fixed heating power, we find that a 30–40% increase in Tped is associated with similar increases in core temperature. Heating in the deep core is varied by employing different groups of neutral beams that deposit their power near the magnetic axis and farther from the axis. In these plasmas on-axis heating produces slightly more peaked temperature profiles although they have 60% more heating power inside r = a/2. Transport models are tested by solving the power balance equations to predict temperatures, which are then compared to the measurements. Predictions of the RLWB and IFS/PPPL models generally agree with the measured temperatures outside r~0.3a, but the Multimode model uniformly predicts temperatures that are too high except in the central region. Tests based on these discharges are not able to discriminate between the transport models of varying stiffness, so we conclude that larger changes are needed in the Pheat and Tped scans.

The significance of real-time feedback control is emphasized in this paper as an indispensable method to improve and sustain the improved plasma characteristics in JT-60U high fusion performance discharges as well as to operate the fusion reactor under the optimal divertor conditions with respect to the heat load and exhaust pumping. In accordance, substantial improvement in the equivalent fusion amplification gain of over unity has been reproducibly achieved at the JT-60U tokamak in the reversed shear mode of operation with the robust feedback controls, where the value of target density was deliberately optimized for the reliable internal transport barrier formation, and the magneto-hydrodynamic stability control was performed with the stored energy feedback. The feedback control techniques also demonstrated the effectiveness to produce quasi-steady-state high-performance plasmas. In addition, three major parameters associated with the fusion reactor instrumentations, namely the neutron production rate, operating density, and divertor radiation power, were simultaneously feedback controlled in the ELMy H-mode plasmas. Here, the matrix response function was evaluated to identify the limitations involved with the linear combination of independent controls. Other advanced feedback schemes, such as the feedback suppression of the neoclassical tearing mode required to sustain high plasma pressure in a steady-state, are alsomore » described. Finally, the controversial issues for the future intelligent plasma control necessary for the advanced steady-stated tokamak reactor are addressed.« less

Fusion plasma performance and confinement studies on JT-60 and JT-60U are reviewed. With the main aim of providing a physics basis for ITER and the steady-state tokamak reactors, JT-60/JT-60U has b...

Non-inductive full current drive (CD) in a reversed magnetic shear (RS) plasma with high energy confinement of 1.4 times the H-mode scaling at 82% of the Greenwald density has been achieved simultaneously for the first time in a tokamak; these results were obtained in the JT-60U tokamak. The full-CD state was realized by a combination of waves in the lower hybrid range of frequency and neutral beam injection together with high bootstrap current fraction (fbs) of 62% to the total plasma current. The combination has demonstrated for the first time active current profile modification in a high fbs RS plasma both in the core and in the peripheral regions.

Although both the transport barriers in the plasma interior and edge, namely internal transport barrier (ITB) and H-mode, are simultaneously required to sustain high-β and improved confinement in steady state, experimental observations suggest that theL-H transition and ITB formation are independent phenomena with different formation conditions, and they often exhibit apparent contradictions to each other. Accordingly, the compatibility conditions of edge and internal barrier formation have been investigated in JT-60U. The L-H threshold power is substantially increased in strong ITB plasmas, and high-performance discharges often persist in L-mode. On the other hand, the ITB formation power increases in H-mode plasmas. It was found, however, that the L-H transition criteria expressed in terms of the edge quantities are similar in plasmas with and without the ITB, and the edge density is extremely low in strong ITB plasmas. It has also been demonstrated in JT-60U that the controlled degradation of the ITB quality, in terms of the changes of toroidal torque input, results in an increase of edge density and induces theL-H transition. Therefore, the quality of ITB modifies the L-H transition threshold power, indicating that the additional `hidden' parameter is present in ITB plasmas. Other approaches of dynamically controlling the L-H transition in ITB plasmas, such as the triangularity ramp developed in JT-60U, are also discussed in this paper.

Toward the study of steady-state operation of fusion plasma, it is important to assess the confinement characteristics under a condition in which fractional power of electron heating is dominant to the total heating power. In a fusion plasma electron heating by α particles is dominant. Especially, the influence on the internal transport barrier (ITB) is an important issue, since formation of ITB is a key issue in advanced steady-state operation from a viewpoint of both attaining highly improved confinement and larger bootstrap current fraction to the total current. In JT-60U, the influence on ITBs when fractional power of electron heating to the total power is increased has been investigated by using negative-ion based neutral beam injection and electron cyclotron range of frequency (ECRF). It is found that in a reversed magnetic shear plasma of high confinement improvement, still higher over all confinement can be achieved at electron heating powers of 50-60% to the total absorbed power. Reduction of the ion temperature at ECRF injection is observed in an ITB plasma of positive shear. And with intense electron central heating on lower hybrid current drive plasma, high central electron temperature with ITB has been formed.

An electron cyclotron range of frequency (ECRF) program has been initiated to study the local heating and current drive in JT-60U. A frequency of 110 GHz was adopted to couple the fundamental O-mode from the low-field side with an oblique toroidal injection angle for the current drive. Experiments were performed at an injection power of ~1.5 MW by using three gyrotrons, each of which has generated the output power up to ~0.8 MW for 3 seconds. A strongly peakedTeprofile was observed and the central electron temperature increased up to ~15 keV when the O-mode was absorbed on the axis. The local electron heating clarified the significant difference in the heat pulse propagation between in the plasmas with internal transport barrier (ITB) and without. The driven current estimated by the Motional Stark Effect (MSE) diagnostic showed that the electron cyclotron (EC) waves drove the plasma current up to ~0.2 MA for an injected power of ~1.3 MW at the local electron temperature and density ofTe~6 keV,ne~0.7×1019 m-3. The measured driven current near the axis was consistent with the theoretical prediction using a Fokker-Planck code. In the case of co-electron cyclotron current drive (ECCD), the sawtooth activity in neutral beam (NB) heated plasma was completely suppressed for 1.5 s with the deposition at the inversion radius, while the sawtooth was enhanced for counter-ECCD at the same deposition condition.

Experimental results towards long sustainment of JT-60U reversed shear (RS) plasmas, where high confinement is achieved owing to strong internal transport barriers, are reported. In a high current plasma with an L mode edge, a deuterium-tritium-equivalent fusion power gain QDTeq = 0.5 was sustained for 0.8 s (≈the energy confinement time) by precise adjustment of the plasma beta using feedback control of stored energy. In a high triangularity plasma with an ELMy H mode edge, a high bootstrap current fraction (≈80%) was obtained in a high q regime (q95≈9), which led to the full non-inductive current drive condition. A normalized beta βN ≈2 and an H factor H89≈3.5 (HH98y2≈2.2) were sustained for 2.7 s (about 6 times the energy confinement time). The values of βN and H89 sustained in RS plasmas were improved remarkably by operating with high bootstrap current fraction and high triangularity.

The dominant causes of the reduction in the enhancement factor of thermal energy confinement with increasing plasma density in ELMy H mode plasmas in JT-60U were analysed by separating the thermal stored energy into core and pedestal components. The enhancement factor of the core confinement evaluated by the offset non-linear scaling which was developed in the ITER H mode confinement database decreased significantly with increasing density, while the pedestal component remained almost constant over a wide range of density. The influence of the edge pedestal structure imposed by the destabilization of type I ELMs on thermal energy confinement of the core plasma was also quantitatively clarified. As the density increased, the saturation of the pressure at the pedestal shoulder, pped, caused by the type I ELM activities forced a reduction in the pedestal temperature. The core temperature for each species, in turn, decreased only by an approximately constant factor with a reduction in the pedestal temperature, resulting in the saturation of the thermal stored energy in the high density regime.

The formation properties of an internal transport barrier (ITB) were investigated in a weak positive magnetic shear plasma by changing the neutral beam heating power. The ion thermal diffusivity in the core region shows L-mode state, weak ITB, and strong ITB, depending upon the heating power. Two features of ITB formation were experimentally confirmed. Weak ITB was formed in spite of the absence of an apparent transition in an ion temperature profile. On the other hand, strong ITB appeared after an apparent transition from the weak ITB. In addition, the ion thermal diffusivity at the ITB is correlated to the radial electric field shear. In the case of the weak ITB, ion thermal diffusivity decreased gradually with increases in the radial electric field shear. There exists a threshold in the radial electric field shear, which allows for a change in state from that of weak to strong ITBs.

Two types of JT-60U discharges are studied with an emphasis on toroidal rotation: in one discharge, which is characterized by the existence of an internal transport barrier (ITB), electron heat transport in the core region is affected by the toroidal rotation direction, while in the other discharge, which is a conventional H-mode plasma without an ITB, the clear correlation between the toroidal rotation direction and electron heat transport is not observed. In both discharges, the impurity density is also found to vary together with the rotation velocity profile. With a flux-tube gyrokinetic code, we have found that the effects of the changes in the rotation velocity profile and the impurity density on electron heat transport are different between these discharges. Including the effects explains the tendency observed in the experiments. First, regarding the rotation velocity profile, which influences heat transport through the inertial force, the dependence of heat transport on the rotation direction changes, according to the gradient of the rotation velocity. Next, an increase in the impurity density stabilizes the ion temperature gradient mode, but can destabilize the trapped electron mode. Therefore, it is found that the difference in the impact of the impurity density on electron heat transport in these discharges can be attributed to the difference in the dominant instability.

The characteristics of internal transport barrier (ITB) structures are studied and active ITB control has been developed in JT-60U reversed shear plasmas. The following results are found. Outward propagation of ITBs with steep Ti gradients is limited to the minimum safety factor location ρqmin. However, ITBs with reduced Ti gradients can move to the outside of ρqmin. The lower boundary of the ITB width is proportional to the ion poloidal gyroradius at the ITB centre. Furthermore, active control of the ITB strength based on modification of the radial electric field shear profile is successfully demonstrated by toroidal momentum injection in different directions or an increase of heating power by neutral beams.

Efficient transmission and polarization in 110 GHz transmission line for JT-60U was studied. A 40-m run transmission test line (inner diameter of 31.75 mm) with eight miter bends, which included a pair of polarizers was designed and constructed. The transmission at 1 MW was performed and the transmission efficiency of 89% (without mirror optical unit (MOU)) was obtained. In basis of the good result, a 1 MW, 110 GHz electron cyclotron range of frequency (ECRF) system was constructed in JT-60U. Transmissions of 1 MW-2.0 s and 0.32 MW-5.0 s were performed in JT-60U ECRF system. The polarization of the transmitted wave in the transmission line was measured and the dependence of plasma electron heating on polarization was investigated. The optimum elliptical polarization of the incidence wave for the ECRF experiments in JT-60U was calculated using the cold dispersion relation with the measured polarization. The optimum polarization was applied to the electron heating experiment of JT-60U.

A tearing mode with m = 3 and n = 2, destabilized in the steady state high-βp H-mode discharges with edge localized mode (ELM), was completely stabilized by local heating and current drive using the 110 GHz first harmonic O-mode electron cyclotron (EC) wave. Here, m and n are poloidal and toroidal mode numbers, respectively. The optimum EC wave injection angle was determined by identifying the mode location from an electron temperature perturbation profile and a safety factor profile. The optimum injection angle was also determined by scanning a steerable mirror during a discharge. In a typical discharge where the tearing mode is completely stabilized, the ratio of the electron cyclotron heating power to the total heating power is 0.17, and the ratio of the EC driven current to the total plasma current is 0.02. Stored energy and neutron emission rate were higher for the case with EC wave injection than that without EC wave injection, which suggests that the reduction of the stored energy and the neutron emission rate was recovered by the tearing mode stabilization.

In JT-60U lower hybrid current drive (LHCD) experiments, a reversed magnetic shear configuration that was accompanied by internal transport barriers was successfully maintained in a quasi-steady state almost entirely non-inductively by the LH driven current and the bootstrap current. The normalized β was kept near 1 and the neutron emission rate was nearly steady as well, indicating no accumulation of impurities in the plasma. Diagnostics showed that the profiles of the electron and ion temperatures, the electron density and the current were almost unchanged during the LHCD phase. The applicability of LHCD to H mode plasmas has also been investigated. It was found that the LH waves can be coupled to an H mode edge plasma with a plasma-wall distance of up to about 14 cm.

The operation of JT-60U reversed shear discharges has been extended to a high plasma current, low q regime keeping a large radius of the internal transport barrier (ITB), and a record value of equivalent fusion multiplication factor in JT-60U, QDTeq = 1.25, has been achieved at 2.6 MA. Operational schemes to reach the low q regime with good reproducibility have been developed. The reduction of Zeff was obtained in the newly installed W shaped pumped divertor. The β limit in the low qmin regime, which limited the performance of L mode edge discharges, has been improved in H mode edge discharges with a broader pressure profile, which was obtained by power flow control with ITB degradation. Sustainment of the ITB and improved confinement for 5.5 s has been demonstrated in an ELMy H mode reversed shear discharge.

The operation of reversed shear plasmas in JT-60U has been extended to the low-q, high-Ip region keeping a large radius transport barrier, and a high fusion performance has been achieved. Record values of deuterium-tritium (DT)-equivalent power gain in JT-60U have been obtained: QDTeq = 1.05, τE = 0.97 s, nD(0) = 4.9 × 1019 m-3 and Ti(0) = 16.5 keV. A large improvement in confinement resulted from the formation of an internal transport barrier (ITB) with a large radius, which was characterized by steep gradients in electron density, electron temperature and ion temperature just inside the position of qmin. Large negative shear regions, up to 80% of the plasma minor radius in the low-qmin regime (qmin~2), were obtained by plasma current ramp-up after the formation of the ITB with the pressure and current profiles being controlled by adjustment of plasma volume and beam power. The ITB was established by on-axis beam heating into a low density target plasma with reversed shear that was formed by current ramp-up without beam heating. The confinement time increased with the radius of the ITB and the decrease of qmin at a fixed toroidal field. High H factors, up to 3.3, were achieved with an L mode edge. The effective one fluid thermal diffusivity χeff had its minimum in the ITB. The values of H/q95 and βt increased with the decrease of q95, and the highest performance was achieved at q95 ~3.1 (2.8 MA). The performance was limited by disruptive beta collapses with βN~2 at qmin~2.

Particle transport is modified in lower hybrid current drive discharges of JT-60U. The density profile becomes broad during the lower hybrid wave injection and the profile change depends on the injected wave spectrum. Particle transport coefficients (diffusion coefficient and profile peaking factor) were evaluated using gas-puff modulation experiments. The diffusion coefficient in the current drive discharges is about three times larger than that in the ohmic discharges. The profile peaking factor decreases in the current drive discharges and the evaluated values are consistent with the measured density profiles.

In the JT-60U tokamak a pre-formed reversed magnetic shear configuration was sustained non-inductively by means of lower hybrid current drive (LHCD). It was formed by neutral beam heating at current build-up and had a reversed shear region extending to 65% of the minor radius. The configuration was sustained for 7.5 s, and the extent of the reversed shear region was kept as large as 55% of the minor radius until the very end. The plasma was stable against MHD activities. It is also demonstrated that LHCD could modify reversed magnetic shear keeping plasma parameters, such as the plasma current and the toroidal magnetic field, the same.

The method for reducing a poloidal magnetic flux consumption of external coils is investigated to reduce the size of the central solenoid (CS) in the DEMO reactor. The reduction of the poloidal magnetic flux consumption during a plasma current ramp-up phase by electron cyclotron (EC) heating is investigated using an integrated modeling code suite, TOPICS. A strongly reversed shear q profile tends to be produced if intense off-axis EC heating is applied to obtain a large reduction of the flux consumption. In order to overcome this tendency, we find a method to obtain the optimum temperature profile which minimizes the poloidal flux consumption for a wide range of the q profile. We try to reproduce the optimum temperature profile for a weakly reversed shear q profile using six EC rays of 20 MW. As a result, the resistive flux consumption during the current ramp-up can be reduced by 63% from the estimation using the Ejima constant of 0.45 and the total flux consumption can be reduced by 20% from the conventional estimation. In addition, we find that the resistive flux consumption is closely related to the volume averaged electron temperature and not to the profile shape. Using this relation, the required heating power is estimated to be 31 MW based on a well established global confinement scaling, ITER L-89P. As a result, it is clarified that the poloidal magnetic flux consumption can be reduced by 20% using 20–31 MW of EC heating for a weakly reversed shear q profile. This reduction of the flux consumption accounts for 10% reduction of the CS radius.

Formation of a high performance plasma with βp = 3.6, βN = 1.6, HH98y2 = 1.6 without the use of the Ohmic heating solenoid was demonstrated for the first time. Plasma start-up and plasma current ramp-up were accomplished by vertical field and shaping coils, combined with heating and current drive by EC, LH and NBI. The q profile is deeply reversed with qmin = 5.6 at r/a = 0.7, and the bootstrap current fraction was at least 90%. This result opens up the possibility of eliminating the Ohmic heating solenoid, which has a great impact on improving tokamak and ST reactor designs.

In the WT-3 tokamak, the toroidal plasma current is started and ramped up to 6.3 kA by the electron cyclotron (EC) wave alone, without ohmic heating (OH) power. After generation of the plasma current by the EC wave, the OH current can be started up with a very low ohmically induced electric field, i.e.

Monodispersed polymer spheres having irregular surface were prepared using monodispersed polystyrene spheres of 10μm diameter as nucleus. They were then patched with n-butyl methacrylate polymer particles by dispersion polymerization of n-butyl methacrylate.The surface condition of the monodispersed polymer spheres was controlled by controlling the mixing rate during the dispersion polymerization.More uniformly patched and more minute polymer particles were recognized on the surface of each sphere at higher mixing rates.

Confinement properties of lower hybrid current driven (LHCD) plasmas and electron cyclotron heated (ECH) plasmas in the WT-3 tokamak are studied. The outward particle and energy fluxes are estimated from parameters of the scrape-off layer (SOL) plasma. During the LHCD period the density and temperature of the SOL plasma decrease drastically, while the density of the core plasma increases, suggesting a decrease of the outward flux. These results show that particle confinement is improved by LHCD. On the contrary, during ECH the density of the SOL plasma increases and the density of the core plasma decreases, suggesting a deterioration of particle confinement. Improved confinement is not obtained near the density limit of LHCD; the improvement is not related with MHD activities during LHCD.

A low power polychromatic beam of microwaves is used to diagnose the behavior of turbulent fluctuations in the core of the JT-60U tokamak during the evolution of the internal transport barrier. A continuous reduction in the size of turbulent structures is observed concomitant with the reduction of the density scale length during the evolution of the internal transport barrier. The density correlation length decreases to the order of the ion gyroradius, in contrast with the much longer scale lengths observed earlier in the discharge, while the density fluctuation level remain similar to the level before transport barrier formation.

The recent success in coupling lower hybrid (LH) waves in high performance plasmas at JET together with the first demonstration on FTU of the coupling capability of the new passive active multijunction launcher removed major concerns on the possibility of using LH on ITER. LH exhibits the highest experimental current drive (CD) efficiency at low plasma temperature thus making it the natural candidate for off-axis CD on ITER where current profile control will help in maintaining burning performance on a long-time scale. We review recent LH results: long internal transport barrier obtained in JET with current profile sustained and controlled by LH acting under real time feedback together with first LH control of flat q-profile in a hybrid regime with Te ~ Ti. Minutes long fully non-inductive LH driven discharges on Tore Supra (TS). High CD efficiency with electron cyclotron in synergy with LH obtained in FTU and TS opening the possibility of interesting scenarii on ITER for MHD stabilization. Preliminary results of LH modelling for ITER are also reported. A brief overview of ITER LH system is reported together with some indication of new coming LH experiments, in particular KSTAR where CW klystrons at the foreseen ITER frequency of 5 GHz are being developed.

Evolution of the lower-hybrid(LH)-driven current profile was measured during the formation of an internal transport barrier (ITB) in a reversed magnetic shear discharge. As the ITB developed, the initially centrally peaked LH-driven current profile gradually turned hollow and was sometimes accompanied by an off-axis peak in the electron temperature profile. These observations indicate the concentration of LH power deposition to the ITB for this case as a result of nonlinear coupling between the LH waves and the target plasma.

Electron density behavior during the fast termination phase of the post-disruption runaway plasma was investigated for the fi rst time. Increases in electron density after the fi rst drop of runaway plasma current were observed. At the second current drop, an increase in electron density with density oscillations having a period of ~1 ms was found. After the second current drop, higher electron density was observed in the case of shorter current decay time.

The temporal and spatial responses of electron channels (the electron density, ne, and the electron temperature, Te) and ion channels (the ion temperature, Ti, and the toroidal rotation velocity, V) to central electron cyclotron heating (ECH) have been investigated in positive shear H-mode plasmas with a relatively peaked Ti profile and internal transport barrier (ITB) plasmas on JT-60U. Ion temperature decreases with ECH after the increase in the electron temperature in the core region. The time scale of the change in Ti is ≈30–60 ms in H-mode plasmas and almost constant in radius. In ITB plasmas, the time scale is shorter around the ITB foot and becomes longer inside the ITB foot. The experimentally measured causality indicates that the decrease in Ti is consistent with the ion temperature gradient critical gradient reduction. This is also verified through a comparison with linear gyrokinetic stability analyses. The electron heat diffusivity increases with ECH in both H-mode and ITB plasmas, correlating to the increase in the ion heat diffusivity. Electron density with a relatively flat ne profile does not decrease with ECH. On the other hand, the electron density with a peaked ne profile decreases with ECH. The flattening of the ne profile is observed after the increase in the electron temperature in the core region. The time scale of the change in ne is about 200–350 ms. Linear gyrokinetic stability analyses imply that the growth rate of the trapped electron modes, which increase outward particle flux, becomes more pronounced during ECH. The counter intrinsic rotation with ECH is identified on H-mode plasmas with a small torque input (BAL-NBI). The counter intrinsic rotation is generated after the increase in the electron temperature and correlates to the change in the electron temperature with ECH around the EC deposition. The radial region where the counter intrinsic rotation is observed is wider than the radial region where the electron temperature varies with ECH. The time scale of the change in the toroidal rotation velocity is about 90–200 ms around the ECH deposition and longer than the time scale of the change in Te and Ti.

Plasma start-up assisted by electron cyclotron resonance heating (ECRH) pre-ionization is theoretically studied for the superconducting tokamak JT-60SA by using a zero-dimensional (0-D) model. Because the toroidal electric field is limited to 0.5 Vm −1 , we clarify the conditions for a robust plasma initiation. Five temporal equations are solved for a spatially-uniform plasma: the electron and neutral density equations, the electron and ion energy density equations, and the electric circuit equation. The numerical calculation results show that an absorbed ECRH power of 150 kW is required to start up the plasma, assuming an initial neutral density of 3.0 × 10 18 m −3 , and that reducing the oxygen and carbon impurity densities facilitates successful start-up. We discuss the dependence of the absorbed ECRH power on the neutral density and impurities. The simulation qualitatively reproduces experimental results from JT-60U.

Dynamics of ion internal transport barrier (ITB) formation and impurity transport both in the Large Helical Device (LHD) heliotron and in the JT-60U tokamak are described. Significant differences between heliotron and tokamak plasmas are observed. The location of the ITB moves outwards during the ITB formation regardless of the sign of magnetic shear in JT-60U, and the ITB becomes more localized in plasmas with negative magnetic shear. In LHD, a low T e/T i ratio (

This paper reports the recent development of reversed shear plasmas with a high bootstrap current fraction (fBS) towards the reactor relevant regime. The previous operation regime of high fBS plasmas is limited at q95 > 8 because of the low beta limit, whereas q95 = 5–6 is envisaged in the DEMO reactor. In the 2008 JT-60U experimental campaign, the high fBS plasma was emphasized in the lower q95 regime by developing the large volume configuration close to the conducting wall for wall stabilization. Thanks to the wall stabilization, high fBS plasmas exceeding the no-wall beta limit are obtained at reactor relevant q95 ~ 5.3. Though the high fBS plasmas are terminated by destabilization of the resistive wall mode, a highly integrated performance is obtained. High values of HH98y2 ~ 1.7, βN ~ 2.7, fBS ~ 0.92 and ne/nGW ~ 0.87 are simultaneously achieved under the reactor relevant conditions of low momentum input and electron temperature nearly equal to ion temperature.

H-modes operated at higher l i with the current ramp down have shown higher energy confinement with higher density in JT-60U. The H-factor evaluated for the core plasma depends strongly on l i with the relation of for the case without sawtooth activities. Centre peaked profiles of electron density and electron temperature are obtained in high l i H-modes. While the peripheral current density profiles are largely modified by the current ramp, the pedestal pressure is not significantly changed. The enhanced energy confinement in high l i H-modes is attributed to the core improvement with the peaked profiles of electron density and electron temperature while no explicit difference in pedestal profile is observed. The electron heat diffusivity is reduced at the plasma core in the high l i case, resulting in the centre peaked T e profile while the T i profiles are approximately unchanged.