Y. Sakamoto

Inter-machine comparison of intrinsic toroidal rotation in tokamaks

Parametric scalings of the intrinsic (spontaneous, with no external momentum input) toroidal rotation observed on a large number of tokamaks have been combined with an eye towards revealing the underlying mechanism(s) and extrapolation to future devices. The intrinsic rotation velocity has been found to increase with plasma stored energy or pressure in JET, Alcator C-Mod, Tore Supra, DIII-D, JT-60U and TCV, and to decrease with increasing plasma current in some of these cases. Use of dimensionless parameters has led to a roughly unified scaling with M-A alpha beta(N), although a variety of Mach numbers works fairly well; scalings of the intrinsic rotation velocity with normalized gyro-radius or collisionality show no correlation. Whether this suggests the predominant role of MHD phenomena such as ballooning transport over turbulent processes in driving the rotation remains an open question. For an ITER discharge with beta(N) = 2.6, an intrinsic rotation Alfven Mach number of M-A similar or equal to 0.02 may be expected from the above deduced scaling, possibly high enough to stabilize resistive wall modes without external momentum input.

Impact of toroidal rotation on ELM behaviour in the H-mode on JT-60U

The mitigation of the large pulsed heat loads induced by edge-localized modes (ELMs) on the divertor plates is one of the most important issues for a tokamak fusion reactor. However, ELMs have been completely suppressed in the quiescent H-mode (QH-mode) plasmas produced in the DIII-D tokamak (see Burrell K H et al 2002 Plasma Phys. Control. Fusion 44 A253). One of the key conditions for producing QH-mode plasmas is that the direction of neutral beam injection (NBI) should be opposite to that of the plasma current (i.e. ctr-NBI), which then leads to the toroidal rotation velocity being in the counter direction to the plasma current. By using various combinations of NBI lines in JT-60U, it has been possible to investigate the impact of the toroidal rotation velocity on ELM behaviour by changing the toroidal momentum input in a detailed manner for similar absorbed NB heating power. It has been determined that the ELM frequency decreases with increased counter toroidal rotation velocity at the plasma edge even to the point of the ELMs disappearing. In addition, the magnitude of the pulsed D α signal at the divertor (D div α) decreases with decreasing ELM frequency. These results indicate that it is possible to control the ELM behaviour through the toroidal momentum input.

Stationary high confinement plasmas with large bootstrap current fraction in JT-60U

This paper reports the results of the progress in stationary discharges with a large bootstrap current fraction in JT-60U towards steady-state tokamak operation. In the weak shear plasma regime, high-βp ELMy H-mode discharges have been optimized under nearly full non-inductive current drive conditions by the large bootstrap current fraction (fBS ~ 45%) and the beam driven current fraction (fBD ~ 50%), which was sustained for 5.8 s in the stationary condition. This duration corresponds to ~26τE and ~2.8τR, which was limited by the pulse length of negative-ion-based neutral beams. The high confinement enhancement factor H89 ~ 2.2 (HH98y2 ~ 1.0) was obtained and the profiles of current and pressure reached the stationary condition. In the reversed shear plasma regime, a large bootstrap current fraction (fBS ~ 75%) has been sustained for 7.4 s under nearly full non-inductive current drive conditions. This duration corresponds to ~16τE and ~2.7τR. The high confinement enhancement factor H89 ~ 3.0 (HH98y2 ~ 1.7) was also sustained, and the profiles of current and pressure reached the stationary condition. The large bootstrap current and the off-axis beam driven current sustained this reversed q profile. This duration was limited only by the duration of the neutral beam injection.

Properties of internal transport barrier formation in JT-60U

The dependence of the thermal diffusivity on the heat flux has been investigated in JT-60U plasmas by varying the neutral beam power. In positive magnetic shear (PS) plasmas, the ion thermal diffusivity (χi) in the core region generally increases with the heating power, as with the L mode at low heating powers. However, as a result of the intensive central heating, a weak internal transport barrier (ITB) is formed, and the χi value in the core region starts to decrease. Corresponding to a further increase in the heating power, a strong ITB is formed and the χi value is reduced substantially. In the case of reversed magnetic shear (RS) plasmas, on the other hand, no power degradation of the χi value is observed in any of the heating regimes. The electron thermal diffusivity (χe) is correlated with the χi value in PS and RS plasmas. Furthermore, the dependence of the ion thermal diffusivity on the radial electric field (Er) shear has also been investigated. By considering the non-locality in the relation between the Er shear and the χi value, a threshold in the effective Er shear to change the state from a weak to a strong ITB is identified.

Response of toroidal rotation velocity to electron cyclotron wave injection in JT-60U

The spontaneous toroidal rotation velocity under the no/low direct toroidal momentum input, particularly from the electron cyclotron (EC) wave injection has been investigated in JT-60U plasmas. It is found that the response of toroidal rotation velocity to the central EC injection is towards the co-current direction in L-mode plasma. The region of the change in the toroidal rotation velocity is wider than the EC deposition profile and similar to that in electron temperature. The observed co-rotation velocity in the combined heating of EC and lower hybrid wave increases with the increase in the stored energy and a large positive radial electric field is formed in the strong co-rotating plasma. Furthermore the short pulse off-axis EC injection experiment shows that the perturbation of the toroidal rotation velocity towards the co-direction propagates to the centre with increase in its amplitude, suggesting an inward pinch in momentum transport.

Dynamic transport study of the plasmas with transport improvement in LHD and JT-60U

Transport analysis during the transient phase of heating (a dynamic transport study) applied to the plasma with internal transport barriers (ITBs) in the Large Helical Device (LHD) heliotron and the JT-60U tokamak is described. In the dynamic transport study the time of transition from the L-mode plasma to the ITB plasma is clearly determined by the onset of flattening of the temperature profile in the core region and a spontaneous phase transition from a zero curvature ITB (hyperbolic tangent shaped ITB) or a positive curvature ITB (concaved shaped ITB) to a negative curvature ITB (convex shaped ITB) and its back-transition are observed. The flattening of the core region of the ITB transition and the back-transition between a zero curvature ITB and a convex ITB suggest the strong interaction of turbulent transport in space.

Formation conditions for electron internal transport barriers in JT-60U plasmas

The formation of electron internal transport barriers (ITBs) was studied using electron cyclotron (EC) heating in JT-60U positive shear (PS) and reversed shear (RS) plasmas with scan of neutral beam (NB) power. With no or low values of NB power and with a small radial electric field (Er) gradient, a strong, box-type electron ITB was formed in RS plasmas while a peaked profile with no strong electron ITBs was observed in PS plasmas within the available EC power. When the NB power and the Er gradient were increased, the electron transport in strong electron ITBs with EC heating in RS plasmas was not affected, while electron thermal diffusivity was reduced in conjunction with the reduction of ion thermal diffusivity, and strong electron and ion ITBs were formed in PS plasmas.

Comparison of electron internal transport barriers in the large helical device and JT-60U plasmas

Plasmas with an electron internal transport barrier (ITB), which is characterized by peaked electron temperature profiles, are obtained in the JT-60U tokamak and in the large helical device (LHD) when the electron cyclotron heating (ECH) is focused on the magnetic axis. The maximum values of R/L Te , where R is the major radius and L Te is the scale length of the electron temperature gradient, are similar for the LHD and JT-60U ITB plasmas. However, there is a clear jump of R/L Te observed in LHD but not in JT-60U in the ECH power scan. This result is consistent with the fact that the trigger mechanism of the electron ITB is the fast transition of the radial electric field from a small negative E r to a large positive E r in LHD and a change of the magnetic shear from positive to negative is required for the formation of the electron ITB in JT-60U. There are also differences in the electron temperature profiles inside the ITB. The flattening of the electron temperature profile inside the strong ITB could be explained by the sharp increase of q values observed in JT-60U, while no flattening of the electron temperature profile is observed in LHD, where the central q values stay low.

Measurement of derivative of ion temperature using high spatial resolution charge exchange spectroscopy with space modulation optics

A new technique to measure the first and second derivatives of the ion temperature profile has been developed by using a charge exchange spectroscopy system with space modulation optics. The space observed is scanned up to +/-3 cm with a cosine wave modulation frequency up to 30 Hz by shifting the object lens in front of the optical fiber bundle by 0.5 mm with a piezoelement. The first and second derivatives of ion temperature are derived from the modulation component of the ion temperature measured by using Fourier series expansion.

Enhanced performance and control issues in JT-60U long pulse discharges

Recent experimental results are reported on control issues involved in long timescales and enhanced performance in JT-60U. The control issues in neoclassical tearing mode (NTM) suppression in the weak shear plasma regime include background optimization through decreasing βp(Lq/Lp) at the rational surface and active stabilization of NTMs using ECCD. By optimizing βp(Lq/Lp), a condition of βN ~ 2.5 was sustained for 10 times the current profile relaxation time and one of βN ~ 2.4 with qmin ~ 1.5 was sustained for 2.8 times the current profile relaxation time, with nearly full non-inductive current drive. In addition, a condition of βN ~ 3 was sustained for 5.5 s through stabilization of NTMs using ECCD, and an EC driven current nearly equal to the bootstrap current was required for complete stabilization. In the reversed shear plasma regime, the issue is the existence of the steady state solution with a large fBS value. By controlling the pressure gradient at the internal transport barrier through toroidal rotation to avoid the disruption, a large fBS value of approximately 75% was sustained for 2.7 times the current profile relaxation time, with nearly full non-inductive current drive, and a steady-state solution with a large fBS value is confirmed. The control issues for the edge pedestal and edge localized modes (ELMs) are control of the pedestal pressure and the energy loss through ELMs. The pedestal pressure increases by >40% through the change in toroidal rotation. The type of ELM can be controlled by toroidal rotation from type-I to grassy.