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Recent progress in the physics and engineering design study for the modification programme of JT-60 is presented. In order to achieve a steady state high-β plasma operation, which is the dominant issue of this programme, physics design for the MHD control and the stability analysis is investigated. Engineering design and the R & D for the superconducting coils, irradiation shield are performed well towards the mission of programme.

This paper reports results on the progress in steady-state high-βp ELMy H-mode discharges in JT-60U. A fusion triple product, nD(0)τETi(0), of 3.1 × 1020 m−3 s keV under full non-inductive current drive has been achieved at Ip = 1.8 MA, which extends the record value of the fusion triple product under full non-inductive current drive by 50%. A high-beta plasma with βN ~ 2.7 has been sustained for 7.4 s (~60τE), with the duration determined only by the facility limits, such as the capability of the poloidal field coils and the upper limit on the duration of injection of neutral beams. Destabilization of neoclassical tearing modes (NTMs) has been avoided with good reproducibility by tailoring the current and pressure profiles. On the other hand, a real-time NTM stabilization system has been developed where detection of the centre of the magnetic island and optimization of the injection angle of the electron cyclotron wave are done in real time. By applying this system, a 3/2 NTM has been completely stabilized in a high-beta region (βp ~ 1.2, βN ~ 1.5), and the beta value and confinement enhancement factor have been improved by the stabilization.

Dissipated power in the spectral gap is taken into account in deriving analytically the lower hybrid current drive efficiency. The efficiency is determined by a minimum down-shifted phase velocity and the quasi-linear velocity diffusion coefficient at the spectral gap in addition to the original wave spectrum. To explain present experimental results in both JT-60 and ASDEX, must be close to the Landau damping limit and must be the same order of magnitude as the collisional friction at . With the suggested values of these two parameters from experimental results, the driven current is mainly determined by the launched wave spectrum, while most of the wave power is dissipated at the spectral gap. This characteristic can explain both the temperature and the dependence of the current drive efficiency.

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

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

The JT-60 LHRF launcher was designed with broad frequency capability (1.74 to 2.23 GHz) to inject multi-megawatt, multi-second RF waves into plasmas. The launcher is characterized by 32 long vacuum waveguides of ∼ 8 m without front windows to avoid plasma bombardment of the ceramic windows in the large tokamak. The front of the launcher is of a detachable grill so it can be changed easily to optimize the wave spectrum. Three types of grills have been developed, and one of them employs a multijunction technique to drive the plasma current efficiently. Three launchers have been installed on JT-60, and the total injection power has reached up to ∼ 10 MW. A good coupling of reflection coefficients, ∼20%, has been obtained for yarious plasma configurations. The development and operation of the LHRF launcher on JT-60 are described.

Simultaneous operation of lower hybrid current drive and pellet injection could be successfully achieved. With peripheral ablation of the pellets by suprathermal electrons, the same net inward flux of particles is found as with deep penetration of pellets into Ohmically heated plasmas. The density profile ne(r) peaks with the same increment of the peaking factor Qn=ne0/(ne) in both cases. The global energy confinement time rises with density, tau E approximately ne, in the combined operation.

Recent lower hybrid current drive (LHCD), and heating (LHH) experiments on JT-60 are reported. The current drive product of neRpIRF approximately 12.5*1019 m-2 MA was achieved at the LH power of approximately 4.5 MW, and the CD efficiency, the energy confinement, the global power balance and the heat load on divertor plates were investigated in high power LHCD plasmas. Nearly steady state H-mode discharges were found during LHCD with two different frequency injections. Sawtooth suppression in NB heated plasmas by LHCD have shown an improvement in confinement near the plasma center. Parametric instabilities in LH heating experiments were significantly reduced by increasing the plasma current, and the stored energy increased linearly with heating power of up to approximately 9 MW at ne approximately 7*1019 m-3 and Ip=2.75 MA. Parametric instabilities near the plasma edge in the ion heating regime were also reduced in peaked density plasmas produced by pellet injection and LH waves increased the central plasma pressure at ne(0) > 1.4*1020 m-3.

Emphases in JT-60 experiments are placed on (1) lower-hybrid (LH) current drive characteristics with a multi-junction type launcher, and (2) the confinement study with combination of neutral beam injection LH current drive and pellet injection. The new multi-junction LH launcher provides a sharp N/sub /// spectrum with high directivity for N/sub ///=1-3.4. The current drive efficiency and the radial distribution of high energy electron production show clear correlation with injected N/sub ///: the current drive efficiency has the maximum at low N/sub ///( approximately 1.3) while flattening of plasma current is more effective in large N/sub ///. A broad radial distribution of high energy electron current and approximately 30% reduction in sawtooth inversion radius were obtained by high N/sub /// ( approximately 2.5) LH injection. To fully suppress the sawtooth activity, low N/sub /// ( approximately 1.3) injection was found to be more effective. Improved energy confinement has been obtained with hydrogen pellet injection. Energy confinement time was enhanced up to 40% relative to usual gas fuelled discharges. The discharge has a strongly peaked electron density profile with ne(0)/(ne) approximately 5 and ne(0) approximately 2.0*1020 m-3. The improved discharges are characterized by a strongly peaked pressure profile within the q=1 magnetic surface, and degrades when a large sawtooth recovers or the pressure gradient may reach a critical value. When large (3 mm, 4 mm) and fast (2.2 km/s) pellets were injected, 30% energy confinement improvement was obtained even during the NB heating of 14 MW. Further investigations of IDC characteristics have been made. The oxygen impurity lines from the main plasma and the main radiative loss drop first. Then the plasma stored energy starts to rise. The particle recycling is reduced around the main plasma, and is localized in the neighborhood of the X-point with a time lag of approximately 0.2 sec. Eventually the discharge shows a significant remote radiative cooling power at the divertor region.

Initial ohmic heating experiments in JT-60 were performed for a three month period of April-June 1985. A maximum plasma current of 1.6 MA was obtained for both divertor and limiter discharges. Low-q discharges of qeff=2.5 and high density discharges of 4.8*1019 m-3 in line-averaged density were obtained in the divertor configuration. In divertor discharges radiated loss from the main plasma can be kept at 20-30% of the ohmic input.

From June to October 1987, JT-60 achieved fusion product (ne(0). tau E*.Ti(0)) of 6*1019 m-3.keV.s with hydrogen plasma at plasma current of 2.8 to 3.1 MA with neutral beam power of approximately 20 MW. The central electron density of 1.3*1020 m-3 was obtained at plasma current of 3 MA with 13 approximately 20 MW neutral beam power and the confinement time reached 0.14-0.18 s. An offset linear scaling law like the Shimomura-Odajima scaling on confinement time will be able to reproduce experimental data better than that of the Goldston type scaling. With low beam energy injection approximately 40 keV, confinement degradation was found. Many short periods (0.05 approximately 0.1 s) of H-mode phase were found in outside X-point divertor discharges with NB or NB+RF(LH or IC) heating power above 16 MW. However, improvement in energy confinement time was limited to 10 %. The ballooning/interchange stability analyses were also made for the outside X-point divertor equilibrium in connection with H0-phase capability. Heating powers of 9.5 MW and 1.9 MW were obtained by LHRF, ICRF injection, respectively. In combined LHRF and NB heating, the incremental energy confinement time of 0.064 s was obtained, which is the same level of that of NB heating only. In combined NB and on-axis ICRF heating of low ne discharge, an incremental energy confinement time of 0.21 s was obtained, which is three times as long as those of NB or ICRF heating only. High energy beam ions were accelerated by ICRF in the central region of the plasma.

The effect of non-circularity of the plasma cross-section on the lower hybrid wave heating (LHWH) is studied by assuming reasonable profiles of the plasma parameters which are consistent with the magnetic flux function obtained by the theory of MHD equilibrium. It is found that, in a tokamak with a non-circular cross-section, the lower hybrid wave can penetrate into the plasma core in the same way as in a conventional tokamak. The absorption regions are in the case of an elliptic cross-section, localized near the midplane. A new function of the poloidal divertor, i.e. the reduction of surface waves with unfavourable effects on LHWH is discussed.

Very slow lower-hybrid waves travelling unidirectionally with a phase velocity of a few times the electron thermal velocity have been used to drive surface currents in a tokamak plasma. Most of the wave energy is dissipated at the plasma surface by virtue of a strong interaction with the electrons. This leads to electron acceleration up to velocities an order of magnitude greater than the electron thermal velocity. Electron trapping by the wave is discussed as a possible mechanism for surface current generation.