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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.

Fusion performance of reversed shear discharges with an {ital L}-mode edge has been significantly improved in a thermonuclear dominant regime with up to 2.8 MA of plasma current in the JT-60U tokamak. The core plasma energy is efficiently confined due to the existence of persistent internal transport barriers formed for both ions and electrons at a large minor radius of r/a{approximately}0.7 near the boundary of the reversed shear region. In an assumed deuterium-tritium fuel, the peak fusion amplification factor defined for transient conditions involving the dW/dt term would be in excess of unity. {copyright} {ital 1997} {ital The American Physical Society}

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.

A new operational scenario of advanced tokamak formation was demonstrated in the JT-60U tokamak. This was accomplished by electron cyclotron and lower hybrid waves, neutral beam injection, and the loop voltage supplied by the vertical field and shaping coils. The Ohmic heating (OH) solenoid was not used but a small inboard coil (part of the shaping coil), providing less than $20%$ of total poloidal flux, was used. The plasma thus obtained had both internal and edge transport barriers, with an energy confinement time of 1.6 times H-mode scaling, a poloidal beta of 3.6, and a normalized beta of 1.6, and a large bootstrap current fraction ($g90%$). This result opens up a possibility to reduce, and eventually eliminate, the OH solenoid from a tokamak reactor, which will greatly improve its economic competitiveness.

Basic design parameters of the JT-60SU device are described. The design addresses wide range of shape and current profile control for the tokamak concept improvement with deuterium operation.

Recently, the JT-60U has employed ECRF system in addition to LHCD and ICRF systems. Three types of RF experiments on JT-60U have been performed. A reversed magnetic shear (R/S) configuration with internal transport barrier (ITB) was successfully maintained for 4.7 s in a quasi-steady state by LHCD in addition to the bootstrap current, where all profiles of temperature and plasma current were almost stationary. On the normal operation without heating during plasma current ramp-up, high electron temperature (Te0∼10 keV) with a strongly peaked profile was produced by LHCD alone at the density of ∼0.5×1019 m−3. This peaked temperature profile was featured by a reduction in core density. The central electron temperature and density slowly changed in several seconds. The performances of ICRF (2ωCH) on the R/S configuration were also studied. Degradation of confinement of energetic particles was observed probably due to the large banana orbit loss. An 110 GHz ECRF system has been operated on JT-60U. Its gyrotron has successfully generated power at the level of up to 830 kW for 300 ms and is under conditioning. On the initial operation, the power of 620 kW was injected into plasmas. The central electron temperature increased from 3 keV to 5.5 keV at ne∼0.6×1019 m−3. A controllability of local heating was also demonstrated by the poloidally steerable mirror.

Recent studies in JT‐60U to establish physics base of a steady‐state tokamak reactor are presented. High fusion performance with fusion triple product nD(0)τETi(0)=1.2×1021 keVsm−3 (equivalent QDT=0.6) has been achieved transiently in the high βp H‐mode regime. By optimizing the current and pressure profiles, a quasi steady‐state high βp H‐mode with nD(0)τETi(0)=(0.4−0.5)×1021 keVsm−3 (equivalent QDT=0.25−0.36) was maintained for up to 1.5 s. In order to maintain such high fusion performance discharges for a long time, non‐inductive current drives and their capability of profile control were widely examined in the JT‐60U tokamak. High‐power (∼7 MW) lower hybrid current drive (LHCD) experiments were performed by using a simplified multijunction launcher. A driven current of 3.6 MA has been achieved with a current drive efficiency of (0.3−0.35)×1020 m−2A/W. Neutral beam current drive (NBCD) was also studied by using co/counter and on/off axis tangential beams (∼4 MW). The maximum driven current and the current drive efficiency were 0.55 MA and 0.06×1020 m−2A/W, respectively. Capability to control the current profile by LHCD and NBCD with a wide range of the plasma current was verified. Fully non‐inductively driven discharges (Ip≤1 MA) with good energy confinement (τE/τEITER89P≤2.5) and a high fraction of the bootstrap current (IBS/Ip=0.6−0.7) have been obtained with NBCD and/or LHCD. These results emphasized an importance of the profile control to obtain highly integrated tokamak performance. To establish integrated basis of physics and technology for an advanced steady‐state tokamak reactor and to contribute ITER activity, we continue conceptual design of a steady‐state tokamak, JT‐60 Super Upgrade (Ip=10 MA, Bt=6.25 T, Rp/ap=4.8/1.5 m, PNBCD=65 MW, PLHCD=20 MW with a pulse length of ≳2000 s) as well.

This paper presents overview of a large multijunction launcher for JT‐60U. The launcher is featured by the multijunction module with the oversized taper waveguide, in order to simplify structure of the launcher. This launcher allows high performances of current drive and current profile control by using very sharp and highly directive spectrum. Initial result of coupling property is also described. A good coupling was observed at a power level of ∼0.8 MW with plasma‐launcher distance of

Extension of an operation range for the sawtooth stabilization by a second harmonic minority ion ICRF heating is being pursued in JT‐60U. The sawtooth stabilization is obtained at relatively high density (ne∼3.8×1019 m−3) and low‐q (qeff=4 at Ip=2.4 MA) for OH target plasmas with only 2.8 MW of ICRF power. The sawtooth stabilization is thus obtained at high 〈ne〉/Ptot value of 0.8×1019 m−3 mW−1, which is notably larger than the value achieved on JET, where fundamental resonance minority ion heating is employed. Another peculiar effect of the second harmonic heating is that stabilization occurs only after the inversion radius expands sufficiently, i.e., rinv/a≥0.26. The sawtooth stabilization by ICRF heating becomes easier with NBI‐heated target plasmas (longest stable period up to 1.7 sec), while it tends to be difficult with the LHCD target plasmas.

A plasma current of 2 MA is driven by lower hybrid current drive (LHCD) and the maximum CD efficiency of ηCD=2.5×1019 m−2 A/W is achieved, which agrees with the previously reported scaling: ηCD∼12

We have extended the ACCOME code with LH current drive to model up‐down asymmetric plasma equilibria. We then applied the code to two well‐diagnosed experiments on JT‐60; one with the multijunction (MJ) grill phased to launch waves with n∥=1.45 and another with both the multijunction grill at the same n∥ and a conventional launcher at n∥=1.70. The simulations predict a driven current within 10% of the measured value.

Lower Hybrid Current Drive (LHCD) experiments had been conducted intensively for last two years in JT‐60. The current drive power of ∼5 MW in the frequencies around 2 GHz was injected from conventional 8×4 waveguide array and 24×4 multi‐junction (MJ) waveguide array launchers. The maximum current of 2 MA and the maximum current drive efficiency ηCD of 0.34×1020m−2 A/W were obtained. It is seen that the improvement of current drive efficiency with increasing electron temperatures comes from the spectral gap issue. Experimental data have been further analyzed by using ACCOME‐LH code which are mainly composed of the equilibrium code ‘‘SELENE’’ and Bonoli‐LH code. The experimental results are well explained quantitatively by the calculation, which implies that the N∥ upshift coming from the toroidal and magnetic shear effects through multiple‐path fills the spectrum gap. Sawtooth stabilization with higher harmonic ICRF heating (2ωcH and 3ωcH) and an appreciable direct electron heating by the fast waves at ∼2....

In the last year, ohmic heating experiments were performed in JT-60 for about three months; May–June 1985 and March 1986. The major experimental effort has been concentrated on the divertor configuration in anticipation of H-mode operation of JT-60 in high power supplementary heating experiments starting in August 1986. Stable discharges with a divertor have been obtained at Ip = 1.85 MA and ne = 5.7 × 1019m−3. Radiated power from the main plasma in divertor discharges is reduced to one third of that in material limiter discharges. The electron density has a broad radial profile and the effective safety reaches qeff = 2.5. The neutral gas pressure in the divertor chamber becomes significantly high and the control of particle recycling has been attempted.

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.

Essential divertor functions — density and energy control — have been investigated in the JT-60 divertor with metal walls. Neutral pressure in the divertor chamber increases in proportion to n2e and high recycling state for particle exhaust is realized. At ne ≤ 6 × 1019m−3, wall pumping by the divertor plates is dominant in particle exhaust. However, at ne ≥ 6 × 1019m−3, particle exhaust by active divertor pumping systems becomes essential. Because of the effectiveness of the divertor, radiation loss in the main plasma is reduced to 5–10% of the absorbed power and impurity concentrations are significantly suppressed at very low level (Zeff = 1.5-2.0). In the experiments with graphite walls, short periods of H-mode phases were found in the outside X-point divertor discharges although improvement in energy confinement is limited to within 10%. In limiter discharges, combination of high vessel temperature (210–300 °C) and low current limiter discharges (1.5 MA) without gas puffing is effective as a wall conditioning in obtaining reproducible high density discharges. The maximum parameters, Ip = 3.2 MA, ne = 1.2 × 1020m−3, qeff = 2.2 at Bt = 4.8T, a Murakami parameter of 7.5, and a stored energy of 3.1 MJ have been attained. The maximum density is limited by the occurence of the MARFE. During high power neutral beam heating, enhanced carbon influx was observed possibly due to chemical sputtering. The wall pumping by the inner wall is still effective in decreasing the electron density from the high density region of ne ~ 1 × 1020m−3. An empirical energy confinement scaling of JT-60 has been drawn in terms of an offset linear function as τE = 0.19 I1.9PPabs + 0.062 a1.8p

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.

Travelling lower hybrid waves were observed to be excited with a helical antenna wound around a glass vacuum chamber in a linear test device. This helical antenna was applied on a simple torus over a quarter of the toroidal circumference. The pitch of the helix determines the phase velocity along the magnetic field, which was chosen to be a few times the electron thermal velocity. In high power experiments the toroidal current was generated in its own rf produced plasma. The electron drift has the same direction as the propagation of lower hybrid wave, and is independent of the direction of toroidal magnetic field in the case of short wavelength. The standing wave was not observed to drive an appreciable amount of current.