T. Ozeki

H-mode experiments with outer and lower divertors in JT-60

In JT-60, H-mode experiments with outer and lower divertors have been performed. In the outer divertor discharge, an H-mode similar to the modes observed in the lower/upper divertor discharges is obtained. Its threshold absorbed power and electron density are 16 MW and 1.8 × 1019 m−3. In the two schemes of combined heating with NB + ICRF and NB + LHRF, H-mode discharges are also obtained. Moreover, in the new configuration with the JT-60 lower divertor, H-mode phases with and without edge localized modes are obtained. The improvement in the energy confinement time in both divertor configurations is limited to values within 10%. The paper mainly presents the H-mode results of the outer divertor discharges. Also, typical results of the lower divertor discharges are shown for a comparison of the H-mode characteristics of the two configurations.

Conceptual design of the steady state tokamak reactor (SSTR)

Abstract A conceptual design study of the Steady State Tokamak Reactor (SSTR) has been performed in order to provide a realistic goal for the fusion research. The SSTR is designed as a DEMO or power reactor to be built in the near future [Y. Seki et al., Proc. 13th Int. Conf. Plasma Phys. and Contr. Nucl. Fusion Res., Washington, USA, IAEA-CN-53/G-1-2]. The major feature of the SSTR is the maximum utilization of bootstrap current in order to reduce the power required for steady state operation [M. Kikuchi, Nucl. Fusion 30 (1990) 265]. This requirement leads to the choice of moderate plasma current (12 MA) and high poloidal beta (βp = 2) for the device, which are achieved by selecting moderate aspect ratio (A = 4) and high toroidal magnetic field (Bt = 16.5 T). A negative-ion-based NBI system is used for central current drive to realize steady state operation. It is shown that a tokamak system based on the small extension of the present physics and technologies can produce net electricity of ∼ 1 GW if the proper physics and engineering R&D are conducted.

The JT-60 tokamak machine

The paper gives an historical overview of the design and construction of the JT-60 tokamak machine starting with its conceptual design in 1973 through to its completion in March 1985. Further the different components of the JT-60 tokamak are described (vacuum vessel, field coils, support structures, etc.).

Integrated view of disruption dynamics on internal electromagnetic and plasma structures in the small tokamak HYBTOK-II

An integrated view of disruption dynamics on internal plasma and associated electromagnetic structures with a high time resolution is presented in a systematic way. The main results are as follows: (i) observation of electron temperature oscillation due to rotation of magnetic islands of m(poloidal)/n(toroidal) = 3/2 mode numbers before the current quench, (ii) finding of rapid pump-out (~20 µs) of plasma current and particles from the central core region to the edge just at the current quench start time and then returning to a peaked profile during the following slow decay phase and (iii) a new proposal for the evaluation of current quench time and the electron temperature dependence of the current quench time, the so-called τ/S scaling.

MHD instabilities leading to disruptions in low beta JT-60U reversed shear plasmas

High performance reversed shear (RS) discharges with strong internal transport barrier (ITB) and flat pressure profile in the plasma core region disrupt frequently even with low beta. We analysed MHD instabilities leading to low beta disruption with measuring fluctuations and current profile with MSE measurement. We mainly observed two types of disruptions. One is the disruption without precursor at surface q ~ integer. The other is the disruption with n = 1 precursor of γ > 100 ms. The poloidal mode number of the n = 1 mode is equal to the outermost integer of q. The n = 1 mode exists continuously from the peripheral region to the ITB layer or the n = 1 modes exist separately at the peripheral region and at the ITB layer. The phase has a difference of 180° between at the peripheral region and at the ITB layer. To explain these characteristics of disruption, we introduce a simple model where disruption occurs when both MHD instabilities at the plasma surface and at the q = m/n surface in the RS region, where m and n are the poloidal and the toroidal mode numbers of the surface MHD instability, are simultaneously unstable. This simple model can explain almost all observed disruption by two processes. One is the surface mode triggered disruption, which occurs when surface q changes; the corresponding q = m/n surface at ITB layer changes discretely. The other is the internal mode triggered disruption, which occurs when the corresponding q = m/n surface becomes unstable gradually.

Construction and testing of JT-60

The JT-60 project is reviewed in terms of design, R&D, construction, commissioning and project management. Design features of JT-60 have been refined and renewed through periodic assessments. Engineering targets have been achieved by R&D efforts. Construction and commissioning have progressed on schedule with intensive project management and control. JT-60 obtained high performance and has entered into the experimental phase after completion of machine construction.

A simulation study on inductive ITB control in reversed shear tokamak discharges

A self-consistent simulation, including a model for improved core energy confinement, demonstrates that externally applied, inductive current perturbations can be used to control both the location and strength of internal transport barriers (ITBs) in a fully non-inductive tokamak discharge. We find that ITB structures formed with broad non-inductive current sources such as LHCD are more readily controlled than those formed by localized sources such as ECCD. Through this external control of the magnetic shear profile, we can maintain the ITB strength which is otherwise prone to deteriorate when the bootstrap current increases. The inductive current perturbation, which can be implemented by a weak Ohmic power, offers steady-state, advanced tokamak reactors an external means of efficient ITB control for regulating the fusion-burn net output and spatial profile.

Alpha particle loss and neutral beam current drive in fusion plasma with central current hole

Alpha particle loss due to toroidal magnetic field (TF) ripple in a current hole (CH) was numerically assessed. In a tokamak with a conventional aspect ratio A (=4.1), α-particles are well confined in plasma with a hole radius (ρhole) of ≤0.3. When ρhole increases to 0.5, α-particle loss can exceed 10%, resulting in an intolerable heat load on the first wall. When born in a CH, toroidally trapped α-particles are regarded as being produced near the CH boundary. This causes an increase in α-particle loss with ρ hole. In order to reduce the loss to an acceptable level of about 3%, the TF ripple amplitude at the plasma surface must be as low as 0.3%. In contrast, a low aspect ratio tokamak has an advantage over α-particle confinement in the CH, in that the calculated α-particle loss for A = 2 was lower than that for conventional values of A. The neutral beam-driven current in the CH was also investigated numerically. The calculation indicates that, in the CH plasma, the beam-driven current is induced in the outer region, compared with the equilibrium current that sustains the CH.

Numerical simulation on current spike behaviour of JT-60U disruptive plasmas

Characteristics and underlying mechanisms for plasma current spikes, which have been frequently observed during the thermal quench of JT-60U disruptions, were investigated through tokamak simulation code simulations including the passive shell effects of the vacuum vessel. Positive shear and reversed shear (PS and RS) plasmas were shown to have various current spike features in the experiments, e.g. an impulsive increase in the plasma current (positive spike) in the majority of thermal quenches, and a sudden decrease (negative spike), that has been excluded from past consideration, as an exception. It was first clarified that the shell effects, which become significant especially at a strong pressure drop due to the thermal quench of high βp plasmas, play an important role in the current spike in accordance with the initial relation of the radial location between the plasma equilibria and the vacuum vessel. As a consequence, a negative current spike may appear at thermal quench when the plasma is positioned further out from the geometric centre of the vacuum vessel. It was also pointed out that a further lowering in the internal inductance, in contradiction to previous interpretation in the past, is a plausible candidate for the mechanism for positive current spikes observed even in RS plasmas. The new interpretation enables us to reason out the whole character of current spikes of JT-60U disruptions.