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The JT-60SA superconducting tokamak is now under construction toward the first plasma in March 2019 as a joint project between the Broader Approach (BA) Satellite Tokamak Programme between Japan and Europe, and the Japanese national programme. The JT-60SA mission is to contribute to early realization of fusion energy by supporting ITER and by complementing ITER in resolving key physics and engineering issues for DEMO reactors. Before procurements of the major components, some R&Ds for key techniques were performed. By November 2014, 25 procurement arrangements (PAs) have been launched (JA: 14 PAs, EU: 11 PAs) covering 88% of the total cost of the BA Satellite Tokamak Programme, and the main components have entered the manufacturing stage. In addition, the JT-60SA tokamak assembly started since January 2013. This paper summarizes the recent progress of the JT-60SA project.

Spatio-temporal measurements of the impurity ion temperature, density, and rotation profiles around the plasma edge region have been made in the JT-60U tokamak, allowing the determination of radial electric field, Er, with the key dimensionless parameter (poloidal Mach number, Upm) at the L-H transition in a number of operational regimes. We found that there is variation in the L-H transition in terms of its time-scale; not only “hard” type transition with a faster time-scale than that seen in the plasma transport (as represented by an energy confinement time, τE) as seen in the many conventional tokamaks, but also “soft” one with a slow time-scale (≈τE) is possible solution, including a complex multi-stage Er transition in the later H-phase. The most important point is that the critical condition for the L-H transition predicted by ion-orbit loss model could be applicable only for “hard” transition (occurred at Upm ≥ 1), and not necessary for “slow” one (occurred even at Upm < 1). Characteristics of the turbulent density fluctuation with the frequency range of 100 kHz at the plasma edge region, in addition to a uniform toroidal MHD oscillation (i.e., n = 0), during ELM-free H-phase are also reported. (© 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

Present status of the JT-60SA (JT-60 Super Advanced) project, implemented jointly by Europe and Japan since 2007, is described. The design of the main tokamak components was completed in late 2008, and all the scientific missions are preserved to contribute to ITER and DEMO reactors. The construction of the JT-60SA has begun with procurement activities for the superconducting magnet systems, vacuum vessel, in-vessel components and other components under the relevant procurement arrangements between the implementing agencies of JAEA (Japan Atomic Energy Agency) in Japan and Fusion for Energy in Europe. Designs and developments of the auxiliary heating systems for JT-60SA have been progressing at JAEA so as to provide the total injection power of 41 MW for 100 s.

The JT-60SA (Super Advanced) project is a joint project of the ITER Satellite Tokamak program and the National Centralized Tokamak program in Japan with missions of supporting ITER, complementing ITER and exploring advanced issues toward DEMO. JT-60SA is a tokamak with superconducting coils, equipped with a poloidal field coil system with wide plasma shape controllability, upper and lower divertors with different shapes, NBI and ECRF with heating power 41 MW and various heating methods, in-vessel coils for suppressing MHD instabilities. With these functions, possibilities of producing ELMy H-mode with improved confinement, full non-inductive current drive of high beta plasmas (β N =3. 7 at Ip=3.5 MA, β N =4.4 at Ip=2.4 MA) and break-even class plasmas necessary for accomplishing the mission have been confirmed. The engineering design of JT-60SA is being done taking large annual neutron production into account. Double skin walls filled with borated water or boron doped concrete are employed for the vacuum vessel and cryostat, respectively, for neutron shield. Divertors structures and first walls are being designed so as to be changed with remote handling systems in the high radiation circumference.

We investigated the relative change in ion temperature (ΔT i /T i ), electron temperature (ΔT e /T e ), pedestal stored energy (ΔW ELM /W ped ), and toroidal rotation velocity (V t ) due to type-I ELMs using diagnostics with fast temporal resolution in JT-60U. The increase in V t was found in the direction of the plasma current right after the ELM. The recovery of V t was found to be faster than that of W dia , T e and T i . It is also found that ΔT i /T ι can be larger than ΔT e /T e .

Electron cyclotron heating (ECH)-assisted startup experiments have been performed in JT-60U. The breakdown loop voltage was successfully reduced from 25 to 4 V (=0.26 V m−1) by 200 kW ECH. This is lower than the 0.3 V m−1, which corresponds to the maximum electric field in ITER. Parameter scans of ECH power, prefill pressure, resonance position and polarization were carried out. The sensitivity of the breakdown to polarization and resonance position was observed. A prefilling gas pressure scan showed that the initial breakdown density increases with prefill pressure when it is is lower than 8 × 10−5 Torr. Higher harmonic ECH was also attempted. The second harmonic ECH-assisted startup was possible with higher ECH power injection. However, the third harmonic ECH-assisted startup was not successful.

In the Japan Atomic Energy Research Institute Tokamak-60 Upgrade (JT-60U), a high- β p ELMy H-mode (high-poloidal-beta high-confinement-mode with edge localized mode) plasma was sustained with β N ∼2.7 for 7.4 s. Real-time neoclassical tearing mode (NTM) stabilization system was established and effective NTM suppression by early electron cyclotron (EC) wave injection was demonstrated. High fusion triple product of n i (0)τ E T i (0)=3.1×10 20 keV⋅s⋅m −3 was achieved using the negative-ion based neutral beam current drive with β N ∼2.5 and the bootstrap current fraction f BS ∼50%. In a hot electron regime, a high electron cyclotron current drive efficiency of 4.2×10 18 A/W/m 2 was achieved at T e ∼21 keV . Innovative current start-up scenario produced a current hole plasma with a very high f BS ∼90%. No accumulation of helium and carbon impurities was observed for internal transport barrier (ITB) plasmas. While argon impurity was accumulated, EC injection effectively exhausted it across ITB. In a regi...

Stabilization of an m = 3/n = 2 neoclassical tearing mode (NTM) has been studied experimentally by applying a local electron cyclotron current drive (ECCD) in the JT-60U tokamak. The EC power is injected before the mode onset, and its effect is compared with the ECCD applied at the saturation phase. The experimental results show that the ECCD applied at the growth phase is more effective than that applied at the saturation phase. The necessary EC power for the suppression is reduced and the mode onset is delayed, indicating the hysterisis characteristics of the NTM on the ECCD stabilization. The dependence on the EC power and injection angle is also shown.

Two types of advanced tokamak scenarios, namely, high βp H-mode and reversed shear (RS) mode, are developed and studied extensively in Japan Atomic Energy Research Institute Tokamak-60 Upgrade (JT-60U) [S. Ide and the JT-60 Team, Phys. Plasmas 7, 1927 (2000)]. These plasmas have both an internal transport barrier (ITB) and an edge transport barrier (ETB). The goal of the recent studies was therefore to get a deeper insight into the physics of ITB and ETB and to increase the discharge performance under a controlled manner based on the physics understanding. In the studies of ETB in an extended high triangularity operation with pellet injection from the high-field side, we have clarified the linkage between main parameters determining the pedestal structure. In the studies of type I edge localized modes (ELMs), simultaneous fast ELM measurements revealed four different phases in an ELM event. It was also established that the density collapse is poloidally asymmetric and is localized at the low-field side. In the studies of ITB, the response of the ion thermal diffusivity to the radial electric field (Er) shear and the effect of electron heating on ITB have been investigated. The importance of nonlocality and the effective Er shear are discussed. In the case of the RS plasma with a strongly reversed current profile, it has been found that there is no plasma current in the central region (called the “current hole”), and it is impossible to drive current in either the positive or negative direction by an electron cyclotron wave inside the current hole. Based on the new understanding of ETB and ITB, the control of the transport barriers is also discussed briefly.

Several important topics relating to MHD behaviour in JT-60U are presented. An extremely hollow current profile was found for the first time in JT-60U with a very small plasma current in the central region, which is called the current hole. Analyses of equilibrium and stability of the current hole discharge show that, as βp increases, the Pfirsh–Schlu;ter current increases, which reduces the poloidal field on the inboard side. It may cause the null point of the magnetic field. β limits due to ideal MHD instability in the current hole do not change very much compared to that of the weakly reversed shear plasma. In high β long pulse plasmas, it is important to suppress the resistive MHD instability, neo-classical tearing mode (NTM), induced by the lack of bootstrap current inside the magnetic island. Analyses of two typical JT-60U discharges with and without NTM show that the polarization current due to the plasma rotation is a plausible candidate to suppress the NTM though it is very sensitive to the value of mode rotation. To establish an advanced control of NTM, a real-time control system, which can identify the island by ECE measurement and control EC ray to the island by a steerable mirror, was prepared in JT-60U, and it successfully stabilized the NTM. Instability due to fast ions is an important subject for sustaining the fusion reaction. In JT-60U, slow frequency sweeping (slow-FS) modes and bursting MHD modes in the frequency range of the toroidicity-induced Alfven eigenmodes (TAEs) are observed by injections of negatively-charged ion based neutral beam of ~360 keV. Analyses by the HINST code show that the resonant TAE mode is a candidate of the slow-FS mode. A bursting mode was observed to change its frequency by 10–20 kHz in 1–5 ms (fast FS mode) and evolve explosively in ~400 μs (abrupt large amplitude event). Analyses of the bursting mode show that the non-linear deformation of the energetic ion distribution may cause the repetition of bursting phenomena and the recovery of lost energetic ions.

Research activities of the Japanese tokamaks JT-60U, JFT-2M, and TRIAM-1M are described. The recent JT-60 program is focused on the establishment of a scientific basis of advanced steady-state oper...

An electron cyclotron range of frequency (ECRF) program has been initiated to study the local heating and current drive in JT-60U. A frequency of 110 GHz was adopted to couple the fundamental O-mode from the low-field side with an oblique toroidal injection angle for the current drive. Experiments were performed at an injection power of ~1.5 MW by using three gyrotrons, each of which has generated the output power up to ~0.8 MW for 3 seconds. A strongly peakedTeprofile was observed and the central electron temperature increased up to ~15 keV when the O-mode was absorbed on the axis. The local electron heating clarified the significant difference in the heat pulse propagation between in the plasmas with internal transport barrier (ITB) and without. The driven current estimated by the Motional Stark Effect (MSE) diagnostic showed that the electron cyclotron (EC) waves drove the plasma current up to ~0.2 MA for an injected power of ~1.3 MW at the local electron temperature and density ofTe~6 keV,ne~0.7×1019 m-3. The measured driven current near the axis was consistent with the theoretical prediction using a Fokker-Planck code. In the case of co-electron cyclotron current drive (ECCD), the sawtooth activity in neutral beam (NB) heated plasma was completely suppressed for 1.5 s with the deposition at the inversion radius, while the sawtooth was enhanced for counter-ECCD at the same deposition condition.

Termination of the runaway electron current generated during plasma disruptions is found in JT-60U during simulated vertical plasma displacement events where the safety factor at the plasma surface qs decreases. For all discharges with runaway electron generation, the runaway current disappears for qs≥2 with the appearance of spikes in the magnetic fluctuations. The growth rate of the spikes in the magnetic fluctuations decreases by an order of magnitude during the termination of runaway current. Corresponding to the loss of runaway electrons by magnetic fluctuations, heat flux pulses are measured at the inner divertor plates, which indicates interaction of the wall with the runaway electrons. The halo current during runaway termination is small and increases after runaway termination with a dominant toroidal mode of n = 1.

Experimental results towards long sustainment of JT-60U reversed shear (RS) plasmas, where high confinement is achieved owing to strong internal transport barriers, are reported. In a high current plasma with an L mode edge, a deuterium-tritium-equivalent fusion power gain QDTeq = 0.5 was sustained for 0.8 s (≈the energy confinement time) by precise adjustment of the plasma beta using feedback control of stored energy. In a high triangularity plasma with an ELMy H mode edge, a high bootstrap current fraction (≈80%) was obtained in a high q regime (q95≈9), which led to the full non-inductive current drive condition. A normalized beta βN ≈2 and an H factor H89≈3.5 (HH98y2≈2.2) were sustained for 2.7 s (about 6 times the energy confinement time). The values of βN and H89 sustained in RS plasmas were improved remarkably by operating with high bootstrap current fraction and high triangularity.

Resistive instabilities and wall stabilization of ideal kink modes with low toroidal mode number n are investigated in JT-60U reversed shear discharges. Localized burst-like MHD activity with n≤3 is observed in the negative shear region with large pressure gradient near the internal transport barrier; this activity is identified as the resistive interchange mode. The burst-like MHD activity is benign with respect to the internal transport barrier and no clear degradation of plasma stored energy by the burst-like MHD activity is observed. The resistive interchange mode, however, results in a major collapse through non-linear coupling with a tearing mode in the positive shear region. A resistive wall mode (RWM) followed by a major collapse is observed in wall stabilized high β discharges exceeding the ideal stability limit with the wall at infinity (βN>βNno-wall). No clear continuous slowing down of the plasma toroidal rotation is observed before the growth of an RWM.

The current drive capability of negative ion based neutral beam injection (N-NBI) in JT-60U has been extended to the reactor relevant regime. The driven current profile and current drive efficiency have been evaluated in a high electron temperature regime Te(0) ≈ 10 keV, and reasonable agreement with the theoretical prediction has been confirmed in this regime. The N-NB driven current reached 1 MA with an injection power of 3.75 MW at a beam energy of 360 keV. A current drive efficiency of 1.55 × 1019A m-2 W-1, approaching the ITER requirement, was achieved in the high βp H mode plasma with Te(0) ≈ 13 keV. This current drive performance permitted sustainment of a high beta (βN = 2.5) and high confinement (HHy2 = 1.4) plasma in the full current driven condition at a plasma current of 1.5 MA. The influence of instabilities on the N-NBI current drive was studied. When a burst-like instability driven by N-NBI occurred in the central region, reductions in loop voltage near the magnetic axis and in the neutron production rate due to loss of beam ions were observed although the lost driven current was at most ~7% of the total driven current. When a neoclassical tearing instability appeared in high beta plasmas, the loss of beam ions was enhanced with increasing instability activity.

With the main aim of providing the physics basis for ITER and steady state tokamak reactors, JT-60U has been optimizing operational concepts and extending discharge regimes towards simultaneous sustainment of high confinement, high βN, high bootstrap fraction, full non-inductive current drive, and efficient heat and particle exhaust utilizing a variety of heating, current drive, torque input and particle control capabilities. In the two advanced operation regimes (reversed magnetic shear (RS mode) and weak magnetic shear (high βp mode)), ELMy H mode discharges characterized by both internal and edge transport barriers and by high bootstrap current fraction fBS have been sustained near the steady state current profile solutions under full non-inductive current drive with appropriate driven current profiles (high βp: HHy2 ≈ 1.4 and βN ≈ 2.5 with negative ion based neutral beams (N-NBs); RS: HHy2 ≈ 2.2 and βN ≈ 2 with fBS ≈ 80%). Multiple pellet injection has extended the density region with high confinement. These operational modes have been extended to the reactor relevant regime with small values of collisionality and normalized gyroradius, and Te~Ti. In the RS regime, QDTeq = 0.5 has been sustained for 0.8 s. Stability has been improved in the high βp mode regime by suppression of the neoclassical tearing mode with local ECCD and in the RS mode regime by wall stabilization. The internal transport barrier structure has been controlled by toroidal rotation profile modification, and transport studies have revealed a semiglobal dynamics of the internal transport barrier structure. Simultaneous pumping at the inner and the outer divertor leg has enhanced He exhaust by ~40%. Argon puff experiments have improved confinement at high density with detached divertor owing to high pedestal temperature Tiped. In H modes, the core confinement degraded with decreasing Tiped, suggesting stiff core profiles. An operational region of grassy ELMs with small divertor heat load has been established at high triangularity, high q95 and high βp. A record value of the neutral beam current drive efficiency of 1.55 × 1019A m-2 W-1 has been demonstrated using N-NB. Abrupt large amplitude events causing a drop in neutron emission have been discovered with frequency inside the TAE gap. Disruption studies have clarified that runaway current is terminated by MHD fluctuations when the surface q decreases to 2-3.

The characteristics of internal transport barrier (ITB) structures are studied and active ITB control has been developed in JT-60U reversed shear plasmas. The following results are found. Outward propagation of ITBs with steep Ti gradients is limited to the minimum safety factor location ρqmin. However, ITBs with reduced Ti gradients can move to the outside of ρqmin. The lower boundary of the ITB width is proportional to the ion poloidal gyroradius at the ITB centre. Furthermore, active control of the ITB strength based on modification of the radial electric field shear profile is successfully demonstrated by toroidal momentum injection in different directions or an increase of heating power by neutral beams.

Quasi-steady-state high βp H mode discharges performed by suppressing neoclassical tearing modes (NTMs) are described. Two operational scenarios have been developed for long sustainment of the high βp H mode discharge: NTM suppression by profile optimization, and NTM stabilization by local electron cyclotron current drive (ECCD)/electron cyclotron heating (ECH) at the magnetic island. Through optimization of pressure and safety factor profiles, a high βp H mode plasma with H89PL = 2.8, HHy,2 = 1.4, βp ≈ 2.0 and βN ≈ 2.5 has been sustained for 1.3 s at small values of collisionality νe* and ion Larmor radius ρi* without destabilizing the NTMs. Characteristics of the NTMs destabilized in the region with central safety factor above unity are investigated. The relation between the beta value at the mode onset βNon and that at the mode disappearance βNoff can be described as βNoff/βNon = 0.05-0.4, which shows the existence of hysteresis. The value of βN/ρi* at the onset of an m/n = 3/2 NTM has a collisionality dependence, which is empirically given by βN/ρi* ∝ νe*0.36. However, the profile effects such as the relative shapes of pressure and safety factor profiles are equally important. The onset condition seems to be affected by the strength of the pressure gradient at the mode rational surface. Stabilization of the NTM by local ECCD/ECH at the magnetic island has been attempted. A 3/2 NTM has been completely stabilized by EC wave injection of 1.6 MW.

Instabilities with frequency chirping in the frequency range of Alfven eigenmodes have been found in the domain 0.1% < βh < 1% and vb||/vA ~1 with high energy neutral beam injection in JT-60U. One instability with a frequency inside the Alfven continuum spectrum appears and its frequency increases slowly to the toroidicity induced Alfven eigenmode (TAE) gap on the timescale of an equilibrium change ( ≈ 200 ms). Other instabilities appear with a frequency inside the TAE gap and their frequencies change very quickly by 10-20 kHz in 1-5 ms. During the period when these fast frequency sweeping (fast FS) modes occur, abrupt large amplitude events (ALEs) often appear with a drop of neutron emission rate and an increase in fast neutral particle fluxes. The loss of energetic ions increases with a peak fluctuation amplitude of θ/Bθ. An energy dependence of the loss ions is observed and suggests a resonant interaction between energetic ions and the mode.

The several hundred keV neutral beamline which is now operating on JT-60U marks the first usage of negative ion sources to produce high power neutral beams. Because this was such a large advance in the state of the art with respect to all system parameters, many new physical processes appeared during the earlier phases of the beam injection experiments. The physical mechanisms responsible for these processes have been explored, and solutions have been implemented for some of them, in particular excessive beam stripping, the secular dependence of the arc and beam parameters, and the non-uniformity of the plasma illuminating the beam extraction grid. This has reduced the percentage of beam heat loading on the downstream grids by roughly a third and has permitted longer beam pulses at higher powers. Progress is being made in improving the negative ion current density and in coping with the sensitivity of the caesium in the ion sources to oxidation by tiny air or water leaks, and the cathode operation is being altered.

Progress on NNB technology enables to demonstrate the N-NB driven current of around 1 MA with the efficiency of 0.155×1020nT−2A/W. A newly installed 110GHz ECH system has demonstrated efficient cen...

In order to obtain improved confinement plasmas with high radiation at high density, Ar gas was injected into ELMy H mode plasmas in JT-60U. A confinement improvement of HH98(y,2) ≈ 1 was obtained with a high radiation loss power fraction (~80%) at an electron density of ~0.65nGW. The HH factor was about 50% higher than that in plasmas without Ar injection.

Recent experimental results for steady state high performance plasma on JT-60U are described, emphasizing plasma control, especially active control of profiles. In reversed shear (RS) plasma, a DT equivalent fusion gain of 0.5 was sustained for 0.8 s (~ the energy confinement time) by controlling plasma β with a newly developed stored energy feedback control technique. An internal transport barrier (ITB) with an RS configuration was successfully maintained by means of LHCD in full current drive quasi-steady state for 4.7 s. The viability of a steady state RS configuration sustained by a large bootstrap current fraction has been demonstrated in ELMy H mode edge RS plasmas for the first time. In high-βp ELMy H mode plasmas, an integrated core plasma performance with H factor of 2.56, βN = 2.4, and bootstrap current fraction and non-inductively driven current fraction close to the ITER requirements was achieved at Ip = 1.5 MA in a near full current drive state by negative ion based NB (N-NB) injection. A rudimentary simulation of a reactor operation with simultaneous feedback control of both the core and divertor plasmas has been demonstrated in ELMy H mode plasma. Aiming at active control of the pressure gradient at the ITB in RS plasmas, a feedback control system of the electron temperature gradient has been developed and optimization of the discharge scenario using this system has recently started.

The JT-60U [Y. Kusama and the JT-60 team, Phys. Plasmas 6, 1935 (1999)] high performance plasmas, such as high poloidal-beta high-confinement mode (high βp H-mode) and reversed magnetic shear (RS) plasmas, have been optimized towards a regime of steady state operation concerning high confinement, high β and high bootstrap current fraction which are requirements of a steady state fusion plasma. What is essential to keep improved confinement in these plasmas is to maintain the internal and/or the edge transport barriers (ITB and/or ETB). A key issue to achieve non-inductive current drive relevant to a steady state fusion reactor is to increase the fraction of the bootstrap current and match the spatial profile to the optimum. In JT-60U, RS plasmas have been optimized not only for the high performance but also for reactor relevant steady state research. In 1999, as the result of the optimization, the equivalent deuterium–tritium (D–T) fusion gain (QDTeq) of 0.5 was sustained for 0.8 s, which is roughly equal...

The current drive and heating properties of negative ion based NBI have been studied comprehensively in JT-60U. It has been confirmed from shine-through measurements of the injected beam (350 keV) that multistep ionization processes are essential in the ionization processes of high energy particles. The profile of the current density driven by a negative ion based NB (N-NB) has been determined experimentally. This is in good agreement with the theoretical prediction, and N-NB driven current reached 0.6 MA with EB = 360 keV and PINJ = 3.7 MW. The current drive efficiency ηCD is increased by increasing electron temperature and improved by increasing beam energy. The fast ions from N-NBs are well confined in the enhanced confinement core by the weak poloidal magnetic field of reversed shear plasmas. A clear H mode transition was obtained with N-NB dominant heating, where the net absorbed power required for an H mode transition seemed similar to the previous result obtained in JT-60U using a low energy beam (90 keV). With the strong electron heating by N-NBI (80% absorbed by electrons), an H factor ( = τE/τITER-89PLE) of 1.64 with Te(0) = 1.4Ti(0) was obtained in the steady state ELMy phase.

In JT-60U lower hybrid current drive (LHCD) experiments, a reversed magnetic shear configuration that was accompanied by internal transport barriers was successfully maintained in a quasi-steady state almost entirely non-inductively by the LH driven current and the bootstrap current. The normalized β was kept near 1 and the neutron emission rate was nearly steady as well, indicating no accumulation of impurities in the plasma. Diagnostics showed that the profiles of the electron and ion temperatures, the electron density and the current were almost unchanged during the LHCD phase. The applicability of LHCD to H mode plasmas has also been investigated. It was found that the LH waves can be coupled to an H mode edge plasma with a plasma-wall distance of up to about 14 cm.

The operation of JT-60U reversed shear discharges has been extended to a high plasma current, low q regime keeping a large radius of the internal transport barrier (ITB), and a record value of equivalent fusion multiplication factor in JT-60U, QDTeq = 1.25, has been achieved at 2.6 MA. Operational schemes to reach the low q regime with good reproducibility have been developed. The reduction of Zeff was obtained in the newly installed W shaped pumped divertor. The β limit in the low qmin regime, which limited the performance of L mode edge discharges, has been improved in H mode edge discharges with a broader pressure profile, which was obtained by power flow control with ITB degradation. Sustainment of the ITB and improved confinement for 5.5 s has been demonstrated in an ELMy H mode reversed shear discharge.

The article treats the recent development of quasi-steady ELMy high ?p H?mode discharges with enhanced confinement and high ? stability, where long sustainment time, an increase in absolute fusion performance and extension of the discharge regime towards low q95 (~3) are emphasized. After modification to the new W shaped pumped divertor, a long heating time (9?s) with a high total heating energy input of 203?MJ became possible without a harmful increase in impurity and particle recycling. In addition, optimization of the pressure profile characterized by the double transport barriers, optimum electron density and/or high triangularity ? made it possible to extend the performance in long pulses. The DT equivalent fusion gain QeqDT ? 0.1 (? = 0.16) was sustained for ~9?s (~50?E, ~10?*p) and QDTeq ? 0.16 (? = 0.3) for 4.5?s at Ip = 1.5?MA. In the latter case with higher ?, an H?factor ( = ?E/?ITER89PLE) of ~2.2, ?N ? 1.9 and ?p ? 1.6 were sustained with 60-70% of the non-inductively driven current. In the low q95 (~3) region, the ? limit was improved by the high ? (~0.46) shape, where ?N ? 2.5-2.7 was sustained for ~3.5?s with the collisionality close to that of ITER-FDR plasmas. The limit of the edge ? parameter in the ELMy phase increases with ?, which is the main reason behind the improved ? limit in a long pulse at high ?. The sustainable value of ?NH also increases with ?. Sustainable ?N is limited by the onset of low n resistive modes. Direct measurement of island width shows agreement with the neoclassical tearing mode theory.

Particle confinement and transport have been systematically analysed for improved confinement modes in JT-60U. A scaling law for the total number of ions in the main plasma of ELMy H mode plasmas was proposed for the first time, with confinement times of the particles fuelled by NBI (centre fuelling) and by recycling and gas puffing (edge fuelling) separately defined. The confinement time increases and decreases with density for centre and edge fuelling, respectively. A comparison between the experimental data and the scaling was made for high βp ELMy H mode and reversed shear plasmas. In the reversed shear plasma, particle confinement was enhanced by a factor of about 2 compared with the scaling, when the enhancement factor of the energy confinement time over the ELMy H mode scaling was larger than 1.2. Density controllability is discussed on the basis of the scaling in ELMy H mode, and it was found that additional fuelling with a confinement time of 0.1-0.4 s is necessary for density control in JT-60U. The particle diffusivity and convection velocity were evaluated by means of a perturbation technique using modulated helium gas puffing in ELMy H mode and reversed shear plasmas. The particle diffusivity and convection velocity were estimated to be 0.2-2 m2/s and from -4 (inward) to 0 m/s for the reversed shear plasma, and 0.4-2 m2/s and from -5.5 (inward) to +2.5 (outward) m/s for the ELMy H mode plasma, respectively. The particle diffusivity around the internal transport barrier was reduced by a factor of 5-6 compared with that in the inside and outside regions in the reversed shear plasma.

High performance regimes of JT-60U plasmas are presented with an emphasis upon the results from the use of a semiclosed pumped divertor with a W shaped geometry. Plasma performance in transient and quasi-steady states has been significantly improved in reversed shear and high ?p regimes. The reversed shear regime elevated an equivalent QeqDT transiently up to 1.25 (nD(0) ?ETi(0) = 8.6 ? 1020m-3?s?keV) in a reactor relevant thermonuclear dominant regime. Long sustainment of enhanced confinement with internal transport barriers (ITBs) with fully non-inductive current drive in a reversed shear discharge was successfully demonstrated with LH wave injection. Performance sustainment has been extended in the high ?p regime with high triangularity, achieving a long sustainment of plasma conditions equivalent to QeqDT ? 0.16?(nD(0) ?ETi(0) ? 1.4? ? ?1020?m-3?s?keV) for ~4.5?s with a large non-inductive current drive fraction of 60-70% of the plasma current. Thermal and particle transport analyses show significant reduction of thermal and particle diffusivities around the ITB, resulting in a strong Er shear in the ITB region. The W shaped divertor is effective for helium ash exhaust, demonstrating a steady exhaust capability of ?*He/?E ? 3-10, which is favourable for ITER. Suppression of neutral backflow and chemical sputtering effects has been observed, while MARFE onset density is rather decreased. Negative ion based neutral beam injection (N-NBI) experiments have created a clear H?mode transition. An enhanced ionization cross-section due to multistep ionization processes was confirmed as theoretically predicted. A current density profile driven by N-NBI is measured in good agreement with the theoretical prediction. N-NBI induced TAEs characterized as persistent and bursting oscillations have been observed from a low hot ? of ?h ? 0.1-0.2% without a significant loss of fast ions.

This paper describes the latest achievements of the Japan Atomic Energy Research Institute Tokamak-60 Upgrade (JT-60U) [H. Kimura and the JT-60 Team, Phys. Plasmas 3, 1943 (1996)] for the establishment of the physics and technology basis for steady-state, fully noninductive current-drive plasmas with high fusion performance. Recent results are highlighted by (1) high performance reversed-magnetic-shear discharges [an idealized equivalent QDT of 1.05, a confinement-enhancement factor (H factor) of 3.23, and a normalized beta value (βN) of 1.88 at Ip=2.8 MA/Pabs=17 MW]; the first observation of thermal transport barriers both for electrons and ions, (2) the first injection of negative-ion-based neutral deuterium beams in the world for studies of current drive, heating, and high-energy particle behavior (achieved so far: 2.5 MW/400 keV, design values: 10 MW/500 keV); high neutralization efficiency of 60% at 370 keV; high current drive efficiency of ηCD=8×1018 m−2 A/W, (3) improved giant edge-localized-mode (...

Formation of a high performance plasma with βp = 3.6, βN = 1.6, HH98y2 = 1.6 without the use of the Ohmic heating solenoid was demonstrated for the first time. Plasma start-up and plasma current ramp-up were accomplished by vertical field and shaping coils, combined with heating and current drive by EC, LH and NBI. The q profile is deeply reversed with qmin = 5.6 at r/a = 0.7, and the bootstrap current fraction was at least 90%. This result opens up the possibility of eliminating the Ohmic heating solenoid, which has a great impact on improving tokamak and ST reactor designs.

The paper compares the transport properties of a set of dimensionless identity experiments performed between JET and JT-60U in the advanced tokamak regime with internal transport barrier, ITB. These International Tokamak Physics Activity, ITPA, joint experiments were carried out with the same plasma shape, toroidal magnetic field ripple and dimensionless profiles as close as possible during the ITB triggering phase in terms of safety factor, normalized Larmor radius, normalized collision frequency, thermal beta, ratio of ion to electron temperatures. Similarities in the ITB triggering mechanisms and sustainment were observed when a good match was achieved of the most relevant normalized profiles except the toroidal Mach number. Similar thermal ion transport levels in the two devices have been measured in either monotonic or non-monotonic q-profiles. In contrast, differences between JET and JT-60U were observed on the electron thermal and particle confinement in reversed magnetic shear configurations. It was found that the larger shear reversal in the very centre (inside normalized radius of 0.2) of JT-60U plasmas allowed the sustainment of stronger electron density ITBs compared with JET. As a consequence of peaked density profile, the core bootstrap current density is more than five times higher in JT-60U compared with JET. Thanks to the bootstrap effect and the slightly broader neutral beam deposition, reversed magnetic shear configurations are self-sustained in JT-60U scenarios. Analyses of similarities and differences between the two devices address key questions on the validity of the usual assumptions made in ITER steady scenario modelling, e.g. a flat density profile in the core with thermal transport barrier? Such assumptions have consequences on the prediction of fusion performance, bootstrap current and on the sustainment of the scenario.

In order to prepare adequate current ramp-up and ramp-down scenarios for ITER, present experiments from various tokamaks have been analysed by means of integrated modelling in view of determining relevant heat transport models for these operation phases. A set of empirical heat transport models for L-mode (namely, the Bohm-gyroBohm model and scaling based models with a specific fixed radial shape and energy confinement time factors of H96-L = 0.6 or HIPB98 = 0.4) has been validated on a multi-machine experimental dataset for predicting the li dynamics within ±0.15 accuracy during current ramp-up and ramp-down phases. Simulations using the Coppi-Tang or GLF23 models (applied up to the LCFS) overestimate or underestimate the internal inductance beyond this accuracy (more than ±0.2 discrepancy in some cases). The most accurate heat transport models are then applied to projections to ITER current ramp-up, focusing on the baseline inductive scenario (main heating plateau current of Ip = 15 MA). These projections include a sensitivity study to various assumptions of the simulation. While the heat transport model is at the heart of such simulations (because of the intrinsic dependence of the plasma resistivity on electron temperature, among other parameters), more comprehensive simulations are required to test all operational aspects of the current ramp-up and ramp-down phases of ITER scenarios. Recent examples of such simulations, involving coupled core transport codes, free-boundary equilibrium solvers and a poloidal field (PF) systems controller are also described, focusing on ITER current ramp-down. © 2011 IAEA, Vienna.

This paper summarizes major plasma physics obtained during the period of JT‐60 operation for 23 years with special emphasis on research towards steady‐state operation of tokamak. Topics included are observation of large bootstrap current fraction, negative shear scenario, demonstration of efficient beam and EC current drive, discovery of current hole, stabilization of the RWM, discovery of ITB (internal transport barrier).

Results are presented of the second harmonic heating experiments in the ion cyclotron range of frequencies (ICRF) on JT-60. Heating and confinement characteristics are investigated by using the hydrogen second harmonic (H-majority 2ωcH scheme and the hydrogen minority second harmonic (H-minority 2ωcH scheme (nH/nHe ≈0.1) in two toroidal phasing modes of antenna currents, in phase ((0,0)) and out of phase ((π, 0)), at PIC ≤ 3 MW, ne =(1.3−6.6) × 1019 m−3 and Ip = 1–2.4 MA. Efficient plasma heating has been demonstrated in both heating schemes. Intense non-Maxwellian ion tails are observed at low density. Sawtooth oscillations with an enhanced period in the electron temperature are found over a wide range of electron densities. The incremental energy confinement time, τEinc, for (π, 0) phasing is larger than for (0,0) phasing, irrespective of the heating scheme. In particular, H-minority 2ωcH heating with (π, 0) phasing shows excellent plasma heating with τEinc ≈ 100–120 ms, which is larger than that of neutral beam injection (NBI) heating by a factor of about two. Different dependences of τEinc on Ip are obtained in the two heating schemes. The (0,0) phasing data of τEinc in H-majority 2ωcH heating increase with Ip up to 1.9 MA, while the (π, 0) phasing data in H-minority 2ωcH heating are almost constant with Ip up to 2.4 MA. The global energy confinement shows an L-mode behaviour. The energy confinement time for H-minority 2ωcH heating at high density agrees well with the Shimomura-Odajima scaling.

Beam acceleration by heating in the second-harmonic ion cyclotron range of frequency (ICRF) in combination with heating by hydrogen neutral beam injection (NBI) was investigated in the JT-60 tokamak. The energy spectra of the accelerated fast ions were measured by a charge exchange neutral energy analyser whose line of sight was intersected by specific beam lines in the plasma core in order to obtain the required information. The dependences of the tail ion temperature on various parameters (electron density, NBI power and the toroidal phasing of the antenna) were examined. Optimized conditions for beam acceleration were evaluated. For combined ICRF heating and NBI heating, an incremental energy confinement time of 210 ms was achieved, which was three times higher than that obtained with NBI heating alone or with ICRF heating alone. This improvement of the energy confinement during combined NBI and ICRF heating can be explained by a build-up of the fast ions accelerated by the ICRF wave. The scaling of the incremental energy confinement time during combined ICRF+NBI heating was obtained.

The energy confinement in high power lower hybrid current driven (LHCD) plasmas has been studied in the JT-60 tokamak. At a plasma current of 1 MA, the diamagnetically estimated energy confinement time in LHCD plasmas has almost the same value as the confinement time in ohmically heated plasmas at e ~ 1.0 × 1019 m−3. The confinement time of high power LHCD plasmas (PLH ≤ 4.5 MW) is longer than that of neutral beam heated plasmas at the same power level. The energy confinement time depends on the LHCD power and on the electron density, but is independent of the plasma current, with α + β ~ −0.3.

The global energy confinement characteristics of neutral beam heated JT-60 discharges are presented. There is a difference in the dependence of the energy confinement time on the plasma current between limiter and divertor discharges. For limiter discharges, the energy confinement increases with plasma current up to 3.2 MA, whereas for divertor discharges this improvement saturates when the safety factor drops below 3, independent of the location of the X-point. The JT-60 L-mode results indicate that the deterioration in energy confinement for q < 3, which is also found in H-mode regimes of other devices, may be a universal characteristic of divertor discharges. Regarding the scaling with plasma size, it is shown that the global/incremental confinement time increases with plasma minor radius. For sufficiently large plasmas, however, the global/incremental confinement time is no longer a function of minor radius.

A strong ion tail with a temperature of ~ 60 keV was produced in JT-60 by lower hybrid wave acceleration of neutral beam injected fast ions. The incremental energy confinement time due to neutral beam heating of a lower hybrid heated plasma was improved by more than a factor of two and reached approximately 110 ms. The confinement of the tail ions is classical and the improvement of confinement can be fully explained by the contribution of the fast ions. The heating process of the beam ions is consistent with the predictions of the quasi-linear theory.

Plasma current (Ip) initiation and ramp‐up utilizing induction by poloidal field (PF) coils located on the outboard side of the torus were demonstrated in both conventional aspect ratio tokamak (100 kA in JT‐60U) and spherical tokamak (10 kA in TST‐2). In the presence of sufficient source of plasma by RF ionization (1 MW in JT‐60U, 100 kW in TST‐2), Ip formation is possible without the existence of a PF null, with the initial vertical field (Bv) in the direction opposite to that required for equilibrium. Utilization of the extra flux swing provided by the PF coils before Bv switches sign is important in arriving at a high enough Ip that matches Bv at the end of PF coil current ramp. In JT‐60U, recharging of the OH coil was observed with only perpendicular and counter NB injection, suggesting overdrive by bootstrap current. In TST‐2, a quasi‐steady‐state Ip of 4 kA was formed and sustained by RF power (PRF) alone, without the use of induction. The plasma current centroid of this type of plasma is located on the outboard side. The dependences of Ip on PRF and Bv were found to be weak. The results presented here offer encouragement to CS‐less operation of a tokamak fusion reactor, which should have a large impact on the economic competitiveness.

Two topics of applications of RF waves to current profile control in JT‐60U are presented; application of lower‐hybrid (LH) waves to safety factor profile control and electron cyclotron (EC) waves to neo‐classical tearing mode (NTM) control. A real‐time control system of safety factor (q) profile was developed. This system, for the first time, enables 1) real time evaluation of q profile using local magnetic pitch angle measurement by motional Stark effect (MSE) diagnostic and 2) control of current drive (CD) location (ρCD) by controlling the parallel refractive index N∥ of LH waves through control of phase difference (Δφ) of LH waves between multi‐junction launcher modules. The method for real‐time q profile evaluation was newly developed, without time‐consuming reconstruction of equilibrium, so that the method requires less computational time. Safety factor profile by the real‐time calculation agrees well with that by equilibrium reconstruction with MSE. The control system controls ρCD through Δφ in suc...

Real‐time neoclassical tearing mode (NTM) stabilization system has been developed in JT‐60U. In this system, mode location is first identified by calculating the plasma shape using the Cauchy condition surface method and evaluating electron temperature perturbation profile in real time. Subsequently, steerable mirror angle is changed so that electron cyclotron (EC) wave is absorbed at the mode location. By using this system, a 3/2 NTM has been completely stabilized, and an increase in the beta value and H‐factor has been observed after the stabilization. Effects of EC injection before the NTM onset (‘early EC injection’) have been investigated. It is found that the NTM is more strongly suppressed for the early injection compared with the case of EC injection after the saturation of the NTM growth (‘late EC injection’). It is also found that the suppression effect for the early injection strongly depends on the EC injection angle, as in the case of the late EC injection.

A central electron temperature of 23 keV was obtained by a combination of ECH and ECCD. A highly localized current density profile was obtained by central ECCD, indicating a diffusion coefficient for resonant electrons of at most 1 m2/s. A comparison of HFS and LFS off‐axis ECCD confirmed the theoretically expected trapped particle effect on the current drive efficiency. LHCD was used to broaden the qmin radius and the ITB foot radius. N‐NBCD was used to shrink the current hole radius. An integrated scenario consisting of (1) a novel plasma start‐up method using RF plasma production and induction by vertical field and shaping coils, (2) a noninductive ramp‐up stage, and (3) a transition to a high‐density, bootstrap‐dominated, high‐confinement plasma, without the use of the OH solenoid, has been demonstrated. The plasma created by this technique had both internal and edge transport barriers, with βp = 3.6 (eβp=1), βN = 1.6, HH98y2 = 1.6 and fBS 90% at Ip = 0.6 MA and ne = 0.5nGW.

Modulation charge eXchange recombination spectroscopy measurements with high spatial and temporal resolution have made the evaluation of the toroidal plasma flow of fully stripped carbon impurity ions ( VϕC6+) in the JT-60U tokamak peripheral region (including, in particular, the separatrix) possible with a better signal-to-noise ratio. By comparing co- and counter-neutral beam injection discharges experimentally, we have identified the boundary condition of VϕC6+ and radial electric field shear ( ∇Er) imposed at the separatrix in high confinement (H-mode) plasmas with edge localized modes (ELMs). The VϕC6+ value at the separatrix is not fixed at zero but varies with the momentum input direction. On the other hand, the ∇Er value is nearly zero (or very weakly positive) at the separatrix. Furthermore, the edge localized mode perturbation does not appear to affect both VϕC6+ and ∇Er values at the separatrix as strongly as that in the pedestal region. The above experimental findings based on the precise edge...