Two data points in the shaded zones of Figs. 3(a) and 3(b) (one point for each figure), were omitted in the published version of these figures. The correct Fig. 3 is shown below.FIG. 3. Radial profiles of (a) electron density, ( b) electron temperature, and (c) ion temperature at 6 s and (d) safety factor at 5.9 s of the discharge shown in Fig. 2. The volume-averaged plasma minor radius is 1.01 m.

The relationship between particle and heat transport in an internal transport barrier (ITB) has been systematically investigated in reversed shear (RS) and high β p mode plasmas of JT-60U. The electron effective diffusivity is well correlated with the ion thermal diffusivity in the ITB region. The ratio of particle flux to electron heat flux, calculated on the basis of the linear stability analysis, shows a similar tendency to an experiment in the RS plasma with a strong ITB. However, the calculated ratio of ion anomalous heat flux to electron heat flux is smaller than the experiment in the ITB region. Helium and carbon are not accumulated inside the ITB even with ion heat transport close to a neoclassical level, but argon is accumulated. The helium diffusivity (D He ) and the ion thermal diffusivity (χ i ) are 5-15 times higher than the neoclassical level in the high β p mode plasma. In the RS plasma, D He is reduced from 6-7 times to a 1.4-2 times higher level than the neoclassical level when χ i is reduced from 7-18 times to a 1.2-2.6 times higher level than the neoclassical level. The carbon and argon diffusivities estimated assuming the neoclassical inward convection velocity are 4-5 times larger than the neoclassical value, even when χ i is close to the neoclassical level. Argon exhaust from the inside of the ITB is demonstrated by applying electron cyclotron heating (ECH) in the high β p mode plasma, where both electron and argon density profiles become flatter. The flattening of the argon density profile is consistent with the reduction of the neoclassical inward convection velocity due to the reduction of the bulk plasma density gradient. In the RS plasma, the density gradient is not decreased by ECH and argon is not exhausted. These results suggest the importance of density gradient control in suppressing impurity accumulation.

An integrated plasma profile control strategy, ARTAEMIS, is being developed for extrapolating present-day advanced tokamak (AT) scenarios to steady-state operation. The approach is based on semi-empirical modelling and was initially explored on JET (Moreau et al 2008 Nucl. Fusion 48 106001). This paper deals with the general applicability of this strategy for simultaneous magnetic and kinetic control on various tokamaks. The determination of the device-specific, control-oriented models that are needed to compute optimal controller matrices for a given operation scenario is discussed. The methodology is generic and can be applied to any device, with different sets of heating and current drive actuators, controlled variables and profiles. The system identification algorithms take advantage of the large ratio between the magnetic and thermal diffusion time scales and have been recently applied to both JT-60U and DIII-D data. On JT-60U, an existing series of high bootstrap current (∼70%), 0.9 MA non-inductive AT discharges was used. The actuators consisted of four groups of neutral beam injectors aimed at perpendicular injection (on-axis and off-axis), and co-current tangential injection (also on-axis and off-axis). On DIII-D, dedicated system identification experiments were carried out in the loop voltage (V ext) control mode (as opposed to current control) to avoid feedback in the response data from the primary circuit. The reference plasma state was that of a 0.9 MA AT scenario which had been optimized to combine non-inductive current fractions near unity with 3.5 < βN < 3.9, bootstrap current fractions larger than 65% and H 98(y,2) = 1.5. Actuators other than V ext were co-current, counter-current and balanced neutral beam injection, and electron cyclotron current drive. Power and loop voltage modulations resulted in dynamic variations of the plasma current between 0.7 and 1.2 MA. It is concluded that the response of essential plasma parameter profiles to specific actuators of a given device can be satisfactorily identified from a small set of experiments. This provides, for control purposes, a readily available alternative to first-principles plasma modelling.

Aiming at optimization of current profile in high-β plasmas for higher confinement and stability, a real-time control system of the minimum of the safety factor (q min) using the off-axis current drive has been developed. The off-axis current drive can raise the safety factor in the centre and help to avoid instability that limits the performance of the plasma. The system controls the injection power of lower-hybrid waves, and hence its off-axis driven current in order to control q min. The real-time control of q min is demonstrated in a high-β plasma, where q min follows the temporally changing reference q min,ref from 1.3 to 1.7. Applying the control to another high-β discharge (βN = 1.7, βp = 1.5) with m/n = 2/1 neo-classical tearing mode (NTM), q min was raised above 2 and the NTM was suppressed. The stored energy increased by 16% with the NTM suppressed, since the resonant rational surface was eliminated. For the future use for current profile control, current density profile for off-axis neutral beam current drive (NBCD) is for the first time measured, using the motional Stark effect diagnostic. Spatially localized NBCD profile was clearly observed at the normalized minor radius ρ of about 0.6–0.8. The location was also confirmed by multi-chordal neutron emission profile measurement. The total amount of the measured beam driven current was consistent with the theoretical calculation using the ACCOME code. The CD location in the calculation was inward shifted than the measurement.

In JT-60U lower hybrid current drive (LHCD) experiments, a reversed magnetic shear configuration that was accompanied by the internal transport barriers was successfully maintained by means of LHCD almost in the full current drive quasi-steady state for 4.7 s. The normalized beta was kept near 1 and the neutron emission rate was almost steady as well indicating no accumulation of impurities into the plasma. Diagnostics data showed that all the profiles of the electron and ion temperatures, the electron density and the current profile were almost unchanged during the LHCD phase. Moreover, capability of LHCD in H-mode plasmas has been also investigated. It was found that the lower hybrid waves can be coupled to an H-mode edge plasma even with the plasma wall distance of about 14 cm. The maximum coupling distance was found to depend on the edge recycling.

The concept for a compact DEMO reactor named 'SlimCS' is presented. Distinctive features of the concept are low aspect ratio (A = 2.6) and use of a reduced-size centre solenoid (CS) which has the function of plasma shaping rather than poloidal flux supply. The reduced-size CS enables us to introduce a thin toroidal field coil system which contributes to reducing the weight and perhaps lessening the construction cost. Low-A has merits of vertical stability for high elongation (κ) and high normalized beta (β N ), which leads to a high power density with reasonable physics requirements. This is because high κ facilitates high n GW (because of an increase in I p ), which allows efficient use of the capacity of high β N . From an engineering aspect, low-A may ensure ease in designing blanket modules robust to electromagnetic forces acting on disruptions. Thus, a superconducting low-A tokamak reactor such as SlimCS can be a promising DEMO concept with physics and engineering advantages.

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.

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 to the much longer scale lengths observed earlier in the discharge, while the density fluctuation level remain similar to the level before transport barrier formation. 2A milestone in the understanding and control of turbulent transport was the discovery of the High confinement regime (H-mode) in the ASDEX device [1]. The H-mode is characterized by a sudden reduction of thermal diffusion in a thin layer known as a transport barrier near the periphery of the toroidal plasma column. The formation of the transport barrier is correlated with a reduction in the turbulent fluctuations and turbulence induced transport. More recently, further improvement in the stored energy of magnetic confinement systems has been achieved through the spontaneous development of internal transport barriers [2]. Unlike the plasma edge, much less is known about core transport barriers, principally due to the difficulty of turbulence measurements in the central region of the discharge. The specific mechanism for the activation and evolution of core transport barriers is still poorly understood, although indirect evidence suggests a key role played by ExB velocity shear, Shafranov shift and reversed magnetic shear [3][4][5][6][7] in suppressing turbulence induced transport.One method with the potential to probe turbulent fluctuations in the plasma core is correlation reflectometry, where multiple microwave beams are launched into the plasma and reflected from density irregularities at different plasma radii [8. 9]. The diagnostic method (analogous to Ionosonde in ionospheric research [10]) uses the pattern of waves reflected from turbulence near a cutoff layer to extract information on the structure of the turbulent eddies. In this paper we report on the direct measurement of the radial correlation length of core turbulence during the evolution of the internal transport barrier in the JT-60U tokamak [11,12] using the method of reflectometry. The measurements 3 employ the simultaneous reflection of multiple microwave beams from the plasma core, together with full wave simulations [13][14][15] in the real plasma geometry in order to infer the scale length and magnitude of density fluctuations during transport barrier evolution.Microwaves in the frequency range 115-140 GHz are launched into the JT-60U device along the toroidal plasma midplane. The microwaves are launched in the X-mode (microwave electric field polarized perpendicular to the plasma magnetic field) so that the reflecting layer is determined by a combination of the magnetic field strength and the plasma density ...