H. Ninomiya

Energy confinement of beam-heated divertor and limiter discharges in Doublet III

Observation of the intensity of the recycling particle flux at the main plasma edge for various limiter and divertor discharges indicates that the gross energy confinement of beam-heated discharges is closely related to the intensity of the edge particle flux. In limiter discharges, the global particle confinement time and the energy confinement time τE show many similarities: 1) linear Ip dependence at Ip < 600 kA, 2) no BT dependence, and 3) deterioration against injection power. Improvement of τE by increasing Ip, for example, is associated with high temperatures at the plasma edge region accompanied by reduced particle recycling. – Divertor discharges with low particle recycling around the main plasma show better energy confinement than limiter discharges at high plasma densities. The improvement of τE is primarily originated in the reduction of heat transport at the main plasma edge region, which is associated with the reduction of recycling particle flux at the main plasma edge. Under certain operation condition, for example, excessive cold-gas puffing, the discharge shows relatively high scrape-off plasma density and strong particle recycling between the main plasma and the limiter. The energy confinement time of these discharges degrades somewhat or reduces completely to that of the limiter discharge. – In low-recycling divertor discharges, the central electron and ion temperature is proportional to the injection power, and the plasma stored energy is proportional to ePabs (scales as INTOR scaling). With ≈ 4 MW beam injection, high-temperature and high-density plasmas were obtained (stored energy up to 280 kJ, Te(0) ≈ Ti(0) ≈ 2.5–3.0 keV at e ≈ (6–7) × 1013 cm−3, τE* ≈ 70 ms).

Objectives and design of the JT-60 superconducting tokamak

A fully superconducting tokamak named JT-60SC is designed for the modification programme of JT-60 to enhance economical and environmental attractiveness in tokamak fusion reactors. JT-60SC aims at realizing high-β steady-state operation in the use of low radio-activation ferritic steel in a low ν* and ρ* regime relevant to the reactor plasmas. Objectives, research issues, plasma control schemes and a conceptual design for JT-60SC are presented.

The softening of current quenches in JT-60U

Current quenches caused by density limit disruptions have been investigated in JT-60U divertor plasmas in order to develop general methods of reducing the current decay rate and of suppressing the generation of runaway electrons. The reduction of the impurity influxes at an energy quench and the direct neutral beam (NB) heating of the plasma core during a current quench were beneficial for reducing the speed of the current quench. The low stored energy just before the energy quench and the high safety factor at the plasma edge had the advantage of decreasing the impurity influxes at the energy quench. The detached plasma state was useful for degrading the energy confinement for both joule and NB heated plasmas within a short time period. Runaway electrons were not generated at plasma densities of more than 5*1019 m-3 measured just before the energy quench. Fast controlled plasma shutdown with energy quenches but without a current quench was demonstrated successfully from 2 MA to zero, with a value of dIp/dt of -6 MA/s. This shutdown was obtained by lowering the stored energy within a short time period of 20 ms, actively using the detached plasma state produced by intense helium gas puffing

Observation of very dense and cold divertor plasma in the beam-heated doublet III Tokamak with single-null poloidal divertor

A Langmuir probe array in the divertor plate has been used to investigate the dense, cold divertor plasma associated with remote radiative cooling in neutral-beam-heated, single-null open-divertor discharges in Doublet-Ill. With injected powers of up to 1.2 MW, the divertor plasma becomes denser and colder as the main plasma line-averaged density e increases, reaching ned= 2.8 X 1014 cm−3. Since the electron temperature drops to Ted = 3.5 eV under these conditions, this cold, dense plasma can provide a solution to the problem of wall erosion.

Non-inductively current driven H mode with high beta N and high beta p values in JT-60U

Attainable βp and βN values were widely improved in JT-60U by peaking of the current profile and broadening of the pressure profile. In a quasi-steady state of ELMyy H mode, βp approximately 2.5-3, βN approximately 2.5-3.1 and H factor approximately 1.8-2.2 were sustained for approximately 1 s simultaneously under the full current drive condition (bootstrap current 60%, beam driven current 48% at Ip = 0.5 MA). In a transient case, the maximum values of βp, βp and βN reached 4.7, 1.2 and 4.2, respectively. The achievable βp value increases systematically with safety factor at the edge. To obtain these improved high β plasmas, it is essential to control MHD activities; suppression of βp collapse and medium m/n (poloidal mode number/toroidal mode number) modes by the modification in current and pressure profiles and control of ELM activity by suitable selection of the heating power

Design and development of JT-60 upgrade — Vacuum vessel and first wall

Abstract The JT-60 tokamak was modified wherein the original outer-single null vacuum vessel and poloidal field coils, were exchanged with those of large dee-shaped cross sections. The basic dimensions of the new vessel allows plasmas of 6 MA, 70 to 100 m 3 volume and vertical elongations of 1.4 to 1.7. The vessel was designed along a new concept of a continuous chamber in a double skin construction. Change of the gas to deuterium from hydrogen is being prepared and will enable neutral beams of 40 MW. The first wall was designed to utilize carbon fiber composite for the divertor and isotropic graphite for the other regions. The divertor tiles are cooled by water, while the other regions are cooled by conduction to the vessel wall.

Achievement of high fusion triple product in the JT-60U high βp H mode

Improvement of an enhanced confinement state in a high poloidal beta (βp) regime without sawtooth activity has been achieved in JT-60U. A confinement mode has been demonstrated where both the edge and the core confinement are improved. The attainable βp was also extended to higher values in this improved mode, because of its broader pressure profile. As a result of the improvement in confinement and in attainable βp, the highest value of the fusion triple product has been extended by a factor of 2.5 over that achieved in the 1992 experiments; it has reached (1.1 ± 0.3) × 1021 m-3.s.keV with a central ion temperature of about 37 keV. The D-D neutron emission rate has also been doubled in these experiments and has reached (5.6 ± 0.6) × 1016 s-1

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.

Application of regression analysis to deriving measurement formulas for feedback control of plasma shape in JT-60

The stepwise regression procedure was applied to obtain measurement formulas for equilibrium parameters used in the feedback control of JT-60, because the measurement formulas for feedback control of a D-shaped divertor plasma position and shape could not be derived from an analytical approach. This procedure automatically selects the predictor variables required to identify the plasma control parameters to leave as few variables as possible. The measurement formulas obtained are simple and compact, making them suitable for rapid calculation in a digitized feedback control system. The resultant control system proved to be stable and reliable, and it was experimentally confirmed that the measurement formulas obtained through this procedure were accurate enough to be applicable to the feedback control of plasma configurations in JT-60.

EFFECTS OF THE SAFETY FACTOR AND THE INTERNAL INDUCTANCE ON MHD ACTIVITIES AND ENERGY CONFINEMENT IN JT-60U

On the basis of ohmic and neutral beam heating experiments in JT-60U, the discharge regions in tokamaks are characterized in terms of the internal inductance li and the effective safety factor at the edge, qeff, in the low beta region. Stable discharges have medium li values, and disruptive discharges occur both in higher and lower li regions for a given qeff. There is also a boundary of li below which sawtooth free discharges occur. Another boundary of li corresponds to the quasi-stationary current distribution profiles, along which the maximum value of rinv/a (sawtooth inversion radius/minor radius) is almost proportional to 1/qeff. At a given qeff value, rinv/a increases with increasing li. Concerning L-mode confinement, the H-factor (H = ?E/?EITER89P) is almost proportional to li0.8 over a wide range of qeff (qeff = 2-14). The degradation of confinement in the low q region (qeff (c-Y)/f, where f = 10.0/qeff2, Y = c-0.96(rinv/a) and c = 0.92