A preparatory study is underway to investigate the feasibility study of high-energetic, non-thermal electron distribution function measurements using a vertical-viewing electron cyclotron emission (ECE) diagnostic on JT-60SA. The system is designed to detect broad ECE spectra (70–260 GHz) due to the second, third, and fourth harmonics using a focusing optics system with four quasi-optical mirrors in the upper port of JT-60SA. A Gaussian beam optics design is performed in vacuum, and ray tracing calculations are performed in plasma using the TRAVIS code to investigate density characteristics. A ceramic viewing dump is also designed to reduce the effects of multiple reflections from the opposing vacuum vessel surface.

The world's largest superconducting tokamak, JT-60SA, has successfully achieved its first plasma operation under the constraint of a low toroidal inductive electric field of 0.15 V m-1. A plasma start-up scenario, leveraging the effective confinement of electrons accelerated by the electron cyclotron wave, proved to be instrumental in reaching this milestone under the challenging conditions. The demonstration of plasma start-up using second harmonic electron cyclotron heating, with an applied toroidal electric field of 0.15 V m-1, strongly validates the feasibility of achieving first plasma operation in ITER.

Superconducting (SC) tokamak JT-60SA plays an essential role in fusion research and development by supporting and complementing the ITER project, providing directions to the DEMO design activity and fostering next generation scientists and engineers. Since the short circuit incident at the terminal joints of equilibrium field coil #1 during the integrated commissioning (IC) in March 2021, both EU and JA implementing agencies (IAs) have examined how to ensure safe operation of JT-60SA by mitigating the risk of possible discharge occurrence inside the cryostat. Based on the experience of the global Paschen tests, the IAs have established a strategy of risk mitigation measures, which is a combination of (i) reinforcement of insulation, (ii) avoiding unnecessary voltage application to the coil systems and (iii) immediate de-energization of the coils when deteriorated vacuum conditions are detected. Thanks to the considerable efforts of the Integrated Project Team members, the IC restarted in May 2023. After confirmation of the SC state of the coil systems (TF, EF and CS), the coil energization test and the plasma operation phase 1 (OP-1) started. The first plasma was successfully achieved on 23 October 2023 with a limited value of voltage and current applied to the coils. The plasma configuration control was also confirmed with low plasma current and low auxiliary heating power conditions. Based on the IO-F4E-QST collaboration, activities of JT-60SA have been shared with the IO and provided an important lesson for ITER assembly and commissioning, and will provide an outstanding contribution to fusion research at large. After OP-1, maintenance & enhancement phase 1 (M/E-1) starts from January 2024, in which in-vessel components are installed, and heating and diagnostic systems are extensively upgraded to allow a high power heating experiment planned in OP-2. In order to make the best use of JT-60SA, a newly organized JT-60SA experiment team will refine the research plan for the future high heating power operation phase.

The avalanche type of transport can induce a long-radial transport and thus can contribute to the global profile formation. In this study, we observed the heat perturbations exhibiting avalanche-like transport in the stellarator/heliotron device, Heliotron J, and the tokamak device, JT-60U. We found that the electron heat propagation in Heliotron J is mainly generated from the heating source region. The relatively high value of the Hurst exponent, which is a signature of avalanches, depends on the total heating power. On the other hand, the electron and ion heat avalanches measured in JT-60U tend to spread from the local peak of the temperature gradient and are not influenced by the heating source profiles. The contrasting features of avalanches in stellarator/heliotrons and tokamaks potentially imply the difference in the temperature profile formation, such as the presence of stiffness.

Turbulent transport is undoubtedly important in magnetic confinement plasmas. It has been reported that a lot of transport dynamics are not satisfying the local diffusive models. Here, we report the quantitative measurement of electron heat flux associated with ballistic propagating long-range transport events, which is considered to be a component of avalanches. In addition, we show the first observations of the substantial impact of avalanche-driven transport on profile resilience (or profile stiffness) observed in JT-60U. We found that, in the channel of the electron heat flux, the ratio between the increment of the avalanche-driven component to that of the total plasma heating becomes dominant (∼80%) in the case of the high-heating limit. This suggests a possible role for avalanche-driven transport to induce profile resilience, which has been evidenced by flux-driven simulations.

In magnetic fusion plasmas, a transport barrier is essential to improve the plasma confinement. The key physics behind the formation of a transport barrier is the suppression of the micro-scale turbulent transport. On the other hand, long-range transport events, such as avalanches, has been recognized to play significant roles for global profile formations. In this study, we observed the impact of the avalanche-type of transport on the formation of a transport barrier for the first time. The avalanches are found to inhibit the formation of the internal transport barrier (ITB) observed in JT-60U tokamak. We found that (1) ITBs do not form in the presence of avalanches but form under the disappearance of avalanches, (2) the surface integral of avalanche-driven heat fluxe is comparable to the time rate change of stored energy retained at the ITB onset, (3) the mean E × B flow shear is accelerated via the ion temperature gradient that is not sustained under the existence of avalanches, and (4) after the ITB formation, avalanches are damped inside the ITB, while they remain outside the ITB.

Plasma with an internal transport barrier (ITB) will be developed in JT-60SA as an attractive operation appropriate for a steady-state fusion reactor. To achieve the ITB plasma while avoiding magnetohydrodynamic instabilities, it is advantageous to simultaneously control the safety factor ( q ) profile and the normalized beta ( β N ). In this study, a control system for simultaneous control of the q profile and β N is studied in simulations prior to the real experiment in JT-60SA. The bootstrap current dominates the total current in the ITB region, which results in a coupling between the pressure profile and the q profile. Thus, it is crucial to control the q profile and β N according to the strength of the ITB. A two-stage neural network (NN)-based control system was developed to address this problem. The first stage estimates the transport properties (i.e. ITB strength) of the plasma from measurements. The second stage consists of several NNs for control of the q profile and β N . According to the ITB strength estimated by the NN in the first stage, the appropriate NN for control is selected from those in the second stage. Each NN in the second stage is trained to control plasmas with different ITB strengths through reinforcement learning employing RAPTOR, an integrated transport code. To validate this system, it is tested in a simulation employing another integrated transport code, TOPICS, to mimic the plasma control in JT-60SA plasmas with various ITB strengths. Stable control of the q profile and β N is achieved in ITB plasmas simulated by both the RAPTOR and TOPICS codes.

Abstract Heat-resistant in-vessel components, i.e., a heat sink, a front-end optics housing, and a diagnostic window have been designed in terms of heat-handling capability and thermal stress and mechanical stress by using a finite element method code. The heat sink, which is exposed to a plasma heat flux of up to 0.3 MW/m2, consists of carbon tiles, a carbon sheet, and a stainless steel heat sink with a water-cooling channel. Analysis shows that at a water flow rate of 0.9 kg/s with a water pressure of 0.5 MPa, an increase in the carbon tile temperature is mitigated below the limit related with detrimental red-hot (900°C). The front-end optics housing temperature and the diagnostic window of sapphire glass temperature are within the allowable temperature. The thermal stress and mechanical stress are less than the allowable value, respectively.

Measuring the time variation of the wavenumber spectrum of turbulence is important for understanding the characteristics of high-temperature plasmas, and the application of a Doppler reflectometer with simultaneous multi-frequency sources is expected. To implement this diagnostic in future fusion devices, the use of a phased array antenna (PAA) that can scan microwave beams without moving antennas is recommended. Since the frequency-scanning waveguide leaky-wave antenna-type PAA has a complex structure, we have investigated its characteristics by modeling it with 3D metal powder additive manufacturing (AM). First, a single waveguide is fabricated to understand the characteristics of 3D AM techniques, and it is clear that there are differences in performance depending on the direction of manufacture and surface treatment. Then, a PAA is made, and it is confirmed that the beam can be emitted in any direction by frequency scanning. The plasma flow velocity can be measured by applying the 3D manufacturing PAA to plasma measurement.

In the negative-ion-based neutral beam injector (N-NBI) of JT-60SA, a hydrogen negative ion beam with energy of 500 keV and a current density of 154 A m −2 for 118 s has been successfully demonstrated. This achievement exceeds the requirements (500 keV, 130 A m −2 , 100 s) for the first time. To maintain stable negative ion production for a long pulse, the optimal temperatures of the chamber wall and plasma grid were analytically examined and were experimentally demonstrated. It was confirmed that the temperature during the beam pulse should be <50 °C for the chamber wall and >200 °C for the plasma grid. Damage to the filament cathode due to an abnormal discharge, so-called arcing, has been mitigated by developing a fast cut-off system of the arc current for around 100 μ s after the arcing. To maintain sufficient voltage holding capability and to reduce the grid heat load due to the beam in the accelerator, techniques developed for the beam acceleration have been applied to this test. As a result of the integration of these techniques, a stable beam over 100 s has been demonstrated successfully. This is the first achievement over a 100 s stable beam with intensity of >75 MW m −2 , which is the required practical level in the N-NBI. These results contribute to the coming NBI system for ITER and DEMO.

A large superconducting machine, JT-60SA has been constructed to provide major contributions to the ITER program and DEMO design. For the success of the ITER project and fusion reactor, understanding and development of plasma controllability in ITER and DEMO relevant higher beta regimes are essential. JT-60SA has focused the program on the plasma controllability for scenario development and risk mitigation in ITER as well as on investigating DEMO relevant regimes. This paper summarizes the high research priorities and strategy for the JT-60SA project. Recent works on simulation studies to prepare the plasma physics and control experiments are presented, such as plasma breakdown and equilibrium controls, hybrid and steady-state scenario development, and risk mitigation techniques. Contributions of JT-60SA to ITER and DEMO have been clarified through those studies.

Construction of the JT-60SA tokamak was completed on schedule in March 2020. Manufacture and assembly of all the main tokamak components satisfied technical requirements, including dimensional accuracy and functional performances. Development of the plasma heating systems and diagnostics have also progressed, including the demonstration of the favourable electron cyclotron range of frequency (ECRF) transmission at multiple frequencies and the achievement of long sustainment of a high-energy intense negative ion beam. Development of all the tokamak operation control systems has been completed, together with an improved plasma equilibrium control scheme suitable for superconducting tokamaks including ITER. For preparation of the tokamak operation, plasma discharge scenarios have been established using this advanced equilibrium controller. Individual commissioning of the cryogenic system and the power supply system confirmed that these systems satisfy design requirements including operational schemes contributing directly to ITER, such as active control of heat load fluctuation of the cryoplant, which is essential for dynamic operation in superconducting tokamaks. The integrated commissioning (IC) is started by vacuum pumping of the vacuum vessel and cryostat, and then moved to cool-down of the tokamak and coil excitation tests. Transition to the super-conducting state was confirmed for all the TF, EF and CS coils. The TF coil current successfully reached 25.7 kA, which is the nominal operating current of the TF coil. For this nominal toroidal field of 2.25 T, ECRF was applied and an ECRF plasma was created. The IC was, however, suspended by an incident of over current of one of the superconducting equilibrium field coil and He leakage caused by insufficient voltage holding capability at a terminal joint of the coil. The unique importance of JT-60SA for H-mode and high-β steady-state plasma research has been confirmed using advanced integrated modellings. These experiences of assembly, IC and plasma operation of JT-60SA contribute to ITER risk mitigation and efficient implementation of ITER operation.

A novel quasilinear turbulent transport model DeKANIS has been constructed founded on the gyrokinetic analysis of JT-60U plasmas. DeKANIS predicts particle and heat fluxes fast with a neural network (NN) based approach and distinguishes diffusive and non-diffusive transport processes. The original model only considered particle transport, but its capability has been extended to cover multi-channel turbulent transport. To solve a set of particle and heat transport equations stably in integrated codes with DeKANIS, the NN model embedded in DeKANIS has been modified. DeKANIS originally determined turbulent saturation levels semi-empirically based on JT-60U experimental data, but now it can also estimate them using a theory-based saturation rule. The new saturation model is still partly connected to experimental data, but it offers the potential for applying DeKANIS independently of the device.

In the JT-60U tokamak, low frequency modes (LFMs) have been reported recently when the m/n = 2/1 neoclassical tearing mode grows sufficiently. LFMs show the low mode frequencies (<20 Hz) with the rotation direction to the counter-current direction (ctr-direction) toroidally, even if strong injection power to the co-current direction (co-direction) is applied by tangential neutral beam injections. As a candidate to drive the rotation to the ctr-direction, the neoclassical toroidal viscosity (NTV) torque is investigated in this study. Indeed, when a LFM is observed, the estimated 'offset velocity' of the NTV torque is larger than the observed toroidal velocity around the q = 2 surface and results in the torque to the ctr-direction (ctr-torque). The torque balance between the NTV torque and the torque induced by the eddy current on the resistive wall is investigated. It is found that the rotation to the ctr-direction having the low mode frequency is obtained as the result of the ctr-torque to the offset velocity. Our investigation suggests the requirement to include the NTV torque in the modeling on the mode locking of neoclassical tearing modes in tokamaks.

The characteristics of ion heat transport inside the internal transport barrier (ITB), which is sustained by weak magnetic shear, are investigated using a cold pulse induced by supersonic molecular-beam injection (SMBI) in JT-60U. It is known that cold-pulse propagation in ITBs usually significantly reduces the temperature and degrades the temperature gradient. When subsequent SMBIs are launched before the temperature has recovered, it is observed that the temperature gradient of the ITB is not monotonically decreased but alternately decreased and increased. Alternating decreasing and increasing phases of the ITB temperature gradient continue for about 1 s (∼6 τ E ), and the properties of the cold-pulse propagation and the flux-gradient relations differ according to the phase. The usual transient transport analysis is also provided by inducing a cold pulse in stationary ITBs. A rapid reduction and recovery of the temperature is observed inside the ITBs, and it is found that the ion-heat flux changes without a variation in the local ion-temperature gradient or a change to the other observable local parameters. The flux-gradient relations exhibit significant hysteresis in two ITB cases: (i) ITBs with strong electron-density gradients and (ii) ITBs with weak ion-temperature gradients. In addition, the range in which hysteresis appears is most likely to depend on the width of the ITB.

The Remote Experimentation Centre (REC) in Japan has been preparing to replicate the full dataset of ITER over 10 000 km distance. In such a multi-site data repository environment, the data location informing service will be essential to find and retrieve the data efficiently. Considering the long latency time and the self sustainability of remote sites, the data location database should be running at each repository site. Multi-master asynchronous replication between cooperating databases will be essential to realize the remote experimental collaborations in fusion research. This study has investigated the functional differences of some relational databases and found that Postgres BDR has the expected database replication capabilities. Bi-directional replication (BDR) tests by using the LHD database and SNET revealed that the throughputs are sufficient for remote collaborations in fusion experiments.

We have used the local- δf gyrokinetic code GS2 to perform studies of the effect of flux-surface shaping on two highly-shaped, low- and high- β JT-60SA-relevant equilibria, including a successful benchmark with the GKV code. We find that for a high-performance plasma, i.e. one with high plasma beta and steep pressure gradients, the turbulent outwards radial fluxes may be reduced by minimizing the elongation. We explain the results as a competition between the local magnetic shear and finite-Larmor-radius (FLR) stabilization. Electromagnetic studies indicate that kinetic ballooning modes are stabilized by increased shaping due to an increased sensitivity to FLR effects, relative to the ion-temperature-gradient instability. Nevertheless, at high enough β , increased elongation degrades the local magnetic shear stabilization that enables access to the region of ballooning second-stability.

We have investigated the role of the m/n = 2/1 neoclassical tearing mode (NTM) for suppression of the sawtooth collapse in JT-60U from the viewpoint of the anomalous transport of the current diffusion, namely flux pumping. In the stabilization experiments of m/n = 2/1 NTMs by electron cyclotron current drive of JT-60U, it has been clarified that the sawtooth collapses occur during or after the stabilization of m/n = 2/1 NTMs. It is also confirmed that the minimum safety factor, q min, is nearly unity before and after the stabilization of an m/n = 2/1 NTM. While the flux pumping by ELM-NTM coupling was reported in DIII-D, the suppression of the sawtooth collapse is observed without ELMs in this study. On the other hand, it is observed that the sawtooth precursor appears during the stabilization of the m/n = 2/1 NTM accompanying the disappearance of the fluctuation of the n = 1 helical core (HC), which is induced by the m/n = 2/1 NTM. In addition, the modulation of the toroidal rotation velocity having the frequency of the n = 1 HC is observed in the core region. Because the helical flow with HCs is a possible source of the dynamo loop voltage in tokamaks, these observations suggest that the suppression of the sawtooth collapse in JT-60U is realized by the dynamo loop voltage due to the n = 1 HC induced by the m/n = 2/1 NTM. Our result indicates that n = 1 HCs induced by other MHD modes also may induce anomalous current transport in tokamaks.

Spatio-temporal structures of ion heat flux and radial electric field are investigated on JT-60U to understand a mechanism of the internal transport barrier (ITB) formation. Contrary to the conventional argument that the ITB formation is due to the E × B flow shear suppressions on turbulence, we observe that the E × B shearing rate during the ITB formation is not enough for turbulence suppressions with k ρ i ∼ 1. Another finding is that the ITB is formed simultaneously with abrupt decreases of ion heat flux in the global region (like the report in Neudatchine et al 1999 Plasma Phys. Control. Fusion 41 L39), and the gradient–flux relations are non-diffusively developed step by step. These observations suggest the importance of the change in the long-radial fluctuations, which is previously observed in JT-60U (Nazikian et al 2005 Phys. Rev. Lett. 94 135002). Therefore, the model of effects by non-uniformity of the radial electric field, under the condition where the scale separation is violated, is employed to explain such fluctuation quenching. Consistent results are obtained showing that the non-uniformity of the radial electric field effectively works on the fluctuations with long-radial wavelength and increases the temperature gradient. It seems natural that the suppression of the long-radial fluctuations drives the global transport improvement to form the ITB.

The impact of fast ions on a trapped electron mode (TEM) is extensively analysed by linear and nonlinear gyrokinetic simulations for a JT-60U plasma at high and low magnetic shear using the Gene code in local approximation. For the first time, it is shown that TEM-induced turbulent transport may remain unaffected by the steep fast ion pressure profile generated by the neutral beam injection. Unlike recent observations of ion temperature gradient (ITG)-induced turbulent transport reductions due to fast ions, TEM-dominated systems could act differently in the presence of a significant fast ion population. The possible role of zonal flows as a saturation mechanism is analyzed, showing that their weak impact in the reported JT-60U scenario might lead to different behaviour of fast ions with respect to the ITG-dominated discharges. It is also shown that Alfvenic shear modes are destabilized at low wave numbers (). They are identified as drift Alfven waves destabilized by ITG, which is a form of Alfvenic ITG instabilities. Deep numerical analysis provides the physical parameter range in which ITG-driven BAEs are stabilized. These results open the way to new possibilities of tailoring future experimental scenarios in order to benefit from transport reduction by fast ions.