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Emissions of ion-cyclotron-range-of-frequency (ICRF) waves have been observed in a solely electron-cyclotron-resonance-heated plasma, namely no fast-ion plasma, on the JT-60U tokamak for the first time. The mode frequencies are around the ion cyclotron frequency at the low-field-side plasma edge. The waves are considered to be electromagnetic and standing waves from poloidal and toroidal phase differences measured with ICRF antennas. Wave characteristics of the ICRF emissions such as frequency, polarization and propagation are similar to those of edge-ion cyclotron emissions (ICEs) driven by neutral-beam-injected fast ions although driving sources are completely different. From analysis of the resonance conditions, this ICRF wave emission is considered to be driven by the electron-cyclotron-resonance-heated fast electrons via the Cherenkov resonance condition. When ICE is used for fast ion diagnostics, such fast-electron-driven ICRF wave emission may complicate its interpretation during electron cyclotron resonance heating.

A relaxation even known as “sawtooth crash” is simulated in a large tokamak plasma with monotonic safety factor close to unity. The domain and the time scale of the event are set to match observations. The simulation follows passive alpha particles with energies 35 keV-3.5 MeV, whose initial density peak is localized in the relaxing domain. While the 35 keV profile flattens, a synergy of multiple physical factors allows the 3.5 MeV profile to remain peaked, facilitating the use of benign sawtooth activity in a fusion reactor to expel helium ash while preserving good confinement of fast alphas., 17th Technical Meeting on Energetic Particles and Theory of Plasma Instabilities in Magnetic Confinement Fusion (EPPI)

サテライト・トカマク計画に関する第36回技術調整会議で、2020年度に実施した統合コミッショニングの結果について報告した。, 36th Technical Coordination Meeting of the Satellite Tokamak Programme

Construction of 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 plasma heating systems and diagnostics have been also progressed, including demonstration of favourable ECRF transmission at multiple frequencies and achievement of long sustainment of high energy intense negative ion beam. Development of all the tokamak operation control systems has been completed together with improved plasma equilibrium control scheme suitable for superconducting tokamaks including ITER. For preparation of 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 essentialy important for dynamic operation in superconducting tokamaks. The Integrated Commissioning has started by vacuum pumping of the Vaccum 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 25.7kA, which is the nominal operating current of the TF coil. For this nominal toroidal field of 2.25T, ECRF was applied and an ECRF plasma was created. The integrated commissioning 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. Unique importance of JT-60SA for H-mode and high- steady-state plasma research has been confirmed using advanced integrated modellings. These experiences of assembly, integrated commissioning and plasma operation of JT-60SA contribute to ITER risk mitigation and efficient implementation of ITER operation.

The ITER Remote Experimentation Centre (REC) project (one of the three sub-projects of the International Fusion Energy Research Centre (IFERC)) is progressing under the agreement between the Government of Japan and the European Atomic Energy Community for the joint implementation of the Broader Approach (BA) activities in the field of fusion energy research. The objectives of the REC activity are to identify the functions and solve the technical issues for the construction of the REC for ITER at Rokkasho, and to develop the remote experiment system and verify the functions required for remote experimentation by using the Satellite Tokamak (JT-60SA) facilities to facilitate the future exploitation of ITER and JT-60SA. The functions of REC will be tested, and the total system will be demonstrated using JT-60SA and existing facilities in the EU, such as JET and WEST. The hardware of the REC has been prepared in Rokkasho Japan, which has the remote experiment room with a large video wall to show the plasma and operation status, IT equipment and a storage system by the reuse of the Helios supercomputer tape library. A broadband network infrastructure of 10Gbps has been installed connected to SINET5. Using this network system, fast data transfer from ITER to REC was examined in 2016, and the transfer of the data volumes expected for the initial ITER experiments has been demonstrated. A secure remote experimentation system has been developed, using JT-60SA, that has functions for preparing and setting of shot parameters, viewing the status of control data, streaming of the plasma status, data-exchange function of shot events, and monitoring of the facility operation. Remote data analysis techniques, data visualisation software, a documentation management and experiment planning system and numerical simulation codes for the preparation and performance estimation of discharges have also been developed.

Active control and stabilization of locked modes (LM) via rotating external magnetic perturbations are numerically investigated under a realistic low resistivity condition. To explore plasma responses to rotating and/or static external magnetic perturbations, we have developed a resistive magnetohydrodynamic code ‘AEOLUS-IT’. By using AEOLUS-IT, dependencies of mode behavior on frequency and polarity of the rotating magnetic perturbation are successfully clarified. Here, the rotational direction of the rotating magnetic perturbation to the equilibrium plasma rotation in the laboratory frame is referred to as ‘polarity’. The rotating magnetic perturbation acts on the background rotating plasma in the presence of a static field. Under such circumstances, there exist bifurcated states of the background rotating plasma, which should be taken into account when studying the dependence of the mode behavior on the rotating magnetic perturbation. It is found that there exist an optimum frequency and polarity of the rotating magnetic perturbation to control the LM, and that the LM is effectively stabilized by a co- polarity magnetic perturbation in comparison with a counter-polarity one.

Fully non-inductive second (2nd) harmonic electron cyclotron (EC) plasma current ramp-up was demonstrated with a newlly developed 28 GHz system in the QUEST spherical tokamak. A high plasma current of 54 kA was non-inductively ramped up and sustained stably for 0.9 s with a 270 kW 28 GHz wave. A higher plasma current of 66 kA was also non-inductively achieved with a slow ramp-up of the vertical field. We have achieved a significantly higher plasma current than those achieved previously with the 2nd harmonic EC waves. This fully non-inductive 2nd harmonic EC plasma ramp-up method might be useful for future burning plasma devices and fusion reactors, in particular for operations at half magnetic field with the same EC heating equipment.

Prediction and likelihood identification of high-beta disruption in JT-60U has been discussed by means of feature extraction based on sparse modeling. In disruption prediction studies using machine learning, the selection of input parameters is an essential issue. A disruption predictor has been developed by using a linear support vector machine with input parameters selected through an exhaustive search, which is one idea of sparse modeling. The investigated dataset includes not only global plasma parameters but also local parameters such as ion temperature and plasma rotation. As a result of the exhaustive search, five physical parameters, i.e., normalized beta βN, plasma elongation κ, ion temperature Ti and magnetic shear s at the q = 2 rational surface, have been extracted as key parameters of high-beta disruption. The boundary between the disruptive and the non-disruptive zones in multidimensional space has been defined as the power law expression with these key parameters. Consequently, the disruption likelihood has been quantified in terms of probability based on this boundary expression. Careful deliberation of the expression of the disruption likelihood, which is derived with machine learning, could lead to the elucidation of the underlying physics behind disruptions.

In JT-60U, simultaneous excitation of n = 1 helical cores (HCs) and m/n = 2/1 Tearing Modes (TMs) was observed [T. Bando et al., Plasma Phys. Control. Fusion 61 115014 (2019)]. In this paper, we have investigated the excitation mechanism of n = 1 HCs with m/n = 2/1 TMs based on the experimental observations and a simple quasi-linear MHD model. In the previous study, it was reported that a "coupling" on the phase of the MHD mode is observed between n = 1 HCs and m/n = 2/1 TMs. In this study, it is found that the coupling is observed with the mode frequency from several Hz to 6 kHz. This indicates that the resistive wall and the plasma control system do not induce the coupling because the both time scales are different from the mode frequency. In addition, n = 1 HCs appear to be the non-resonant mode from the two observations: n = 1 HCs do not rotate with the plasma around the q = 1 surface in the core and the coupling is also observed even when qmin > 1. It is also observed that the electron fluctuation due to an n = 1 HC in the core region disappears with the stabilization of an m/n = 2/1 neoclassical tearing mode by electron cyclotron current drive, implying that n = 1 HCs are driven by m/n = 2/1 TMs. This perspective, n = 1 HCs are driven by m/n = 2/1 TMs, is supported by the observation that the saturated amplitude of the m/n = 1/1 component of the radial displacement in the core is smaller than that of the m/n = 2/1 component. Finally, we revisit a quasi-linear MHD model where the m/n = 1/1 HC is induced directly by the sideband of the current for the m/n = 2/1 TM, which allows to excite the non-resonant m/n = 1/1 mode. The model also describes the characteristic of the coupling, fm/n=1/1(HC) = 2fm/n=2/1(TM).

A 28~GHz system with a high-power gyrotron tube has been used for the QUEST spherical tokamak to form an over-dense plasma for electron Bernstein wave heating (EBWH) and current drive with an 8.2~GHz-wave. Non-inductive high-density plasma ramp-up experiments with dual-frequency (dual$-f$) electron cyclotron (EC) [ 8.2~GHz and 28~GHz ] waves were conducted. A spontaneous density jump (SDJ) to an over-dense state was observed as a bifurcation phenomenon in the dual$-f$ wave experiment. The over-dense plasma on the 8.2~GHz-wave was non-inductively ramped up to 25~kA, and was maintained for 0.4~s under stable plasma equilibrium after two such jumps in one shot. Heating to mildly energetic electrons and bulk electrons was observed even in the over-dense region. The electrostatic EBWH effect on the mildly energetic electrons in the over-dense region is assessed following a dispersion analysis of the 8.2 GHz-wave. The bulk electron heating effect observed is explained as heat exchange from mildly energetic electrons heated by the electrostatic EBWH wave. Remarkably, a high hard-X-ray-radiation temperature ~ 500~keV was also observed in tangential viewing for current-carrying electrons in the over-dense core region. Synergetic heating from the overlap of different 28~GHz EC harmonic resonances as well as higher harmonic heating is discussed for maintaining the highly energetic electrons in the over-dense core region. In addition, the SDJ process and mechanism are considered based on the discussion of the electron heating effects with the 8.2 GHz-wave.

The ITER Remote Experimentation Centre (REC) project (one of the three sub-projects of the International Fusion Energy Research Centre (IFERC)) is progressing under the agreement between the Government of Japan and the European Atomic Energy Community for the joint implementation of the Broader Approach (BA) activities in the field of fusion energy research. The objectives of the REC activity are to identify the functions and solve the technical issues for the construction of the REC for ITER at Rokkasho, and to develop the remote experiment system and verify the functions required for remote experimentation by using the Satellite Tokamak (JT-60SA) facilities to facilitate the future exploitation of ITER and JT-60SA. The functions of REC will be tested, and the total system will be demonstrated using JT-60SA and existing facilities in the EU, such as JET and WEST. The hardware of the REC has been prepared in Rokkasho Japan, which has the remote experiment room with a large video wall to show the plasma and operation status, IT equipment and a storage system by the reuse of the Helios supercomputer tape library. A broadband network infrastructure of 10Gbps has been installed connected to SINET5. Using this network system, fast data transfer from ITER to REC was examined in 2016, and the transfer of the data volumes expected for the initial ITER experiments has been demonstrated. A secure remote experimentation system has been developed, using JT-60SA, that has functions for preparing and setting of shot parameters, viewing the status of control data, streaming of the plasma status, data-exchange function of shot events, and monitoring of the facility operation. Remote data analysis techniques, data visualisation software, a documentation management and experiment planning system and numerical simulation codes for the preparation and performance estimation of discharges have also been developed.

The International Tokamak Physics Activity topical group on integrated operational scenarios has compiled a database of stationary H-mode discharges at q(95) similar to 3 from AUG, C-Mod, DIII-D, J ...

This paper proposes a simulation system that provides close to the real experience of wearing and operating a virtual hand prosthesis with the help of augmented reality (AR) technology. The system allows user wearing a virtual hand prosthesis in an AR scene and the created scene is visual feedback through a display. Meanwhile, the user can operate the virtual hand with his own electromyography (EMG) signals. The system aims at improving the sense of reality when the user tries to operate a virtual hand prosthesis. Efforts have been made in three aspects. We design a virtual hand 3D model which approximates to a real hand in terms of dimensions and kinematics and present an AR environment realization technique using the Microsoft Kinect. To anticipate the hand movement, an EMG-based motion discrimination method using recurrent neural network is also proposed. User evaluation shows that the system improves the realism in wearing and controlling a virtual hand. The system is developed as a prototype for performing task training or control evaluation for myoelectric hand prosthesis.

The operation scenarios for plasma breakdown and current ramp-up phases in JT-60SA tokamak have been developed and verified in simulation using the TOSCA code. The induced current in the conducting elements such as vacuum vessel and stabilizing plate increases to the comparable level of plasma current of ∼600 kA during the breakdown phase and thus enhances the strength of stray field. The optimized scenarios for half and full pre-magnetization cases satisfied the conditions required for the plasma initiation. At the initial plasma, the vertical magnetic field required to sustain the plasma position was controlled by the outer equilibrium field (EF) coil currents which compensate for a vertical field due to a large eddy current. The condition for the formation of divertor configurations given by the combination of the magnetic flux for plasma and the plasma current has been satisfied which enables us to develop the operational scenarios with a smooth transition from a limiter to a divertor configuration.

We have designed in-vessel coils for a correction of magnetic error fields in JT-60SA. In order to design the in-vessel coils, namely, error field correction coils (EFCCs), error fields from several sources such as manufacturing and assembly errors of superconducting coils are calculated by Monte-Carlo approach. Required EFCC currents to correct error fields are evaluated by a least square method. Additionally, by the field line tracing, it is found that resonant magnetic perturbations (RMPs) by the EFCC enable to produce stochastic magnetic structures at the edge region, that is applicable to an ELM control.

By using a one-dimensional model, we analyze plasma start-up assisted by second-harmonic extraordinary-mode electron cyclotron (EC) resonance heating (ECRH). The model leads to energy transport equ...

The JT-60SA tokamak, being built under the Broader Approach agreement jointly by Europe and Japan, is due to start operation in 2020 and is expected to give substantial contributions to both ITER and DEMO scenario optimisation. A broad set of preparation activities for an efficient start of the experiments on JT-60SA is being carried out, involving elaboration of the Research Plan, advanced modelling in various domains, feasibility and conception studies of diagnostics and other sub-systems in connection with the priorities of the scientific programme, development and validation of operation tools. The logic and coherence of this approach, as well as the most significant results of the main activities undertaken are presented and summarised. EURATOM 633053

Transport modelling of plasmas with internal transport barriers in JT-60U and JET tokamaks has been carried out using integrated modelling codes TOPICS and CRONOS for the prediction of high-beta steady-state scenario in JT-60SA, which shares important characteristics with both tokamaks. Typical models of anomalous heat transport, which is one of major uncertainties in the prediction, have been validated for the experimental data in JT-60U and JET, and TOPICS and CRONOS equipped with the models are used for the model verification. It is found that CDBM model predicts temperatures close to experiments or underestimates them, and thus can be used for the conservative prediction, which considers a lower bound of plasma performance. By using the CDBM model, a JT-60SA high-beta steady-state plasma has been conservatively predicted within the machine capability. The conservative prediction shows that the JT-60SA has enough capability to explore high-beta steady-state plasmas and their controllability. Model modifications related with an E × B shear effect to improve the prediction capability are discussed.

The Monte-Carlo (MC) approach has a lot of flexibility in impurity transport modeling in the SOL and divertor region. However, in the divertor plasma simulation with the noble impurity seeding, characteristic time of the impurity transport especially in the sub-divertor chamber is long because the MC calculation of the impurity gas transport can be finished only by exhaust. The impurity MC calculation for such long exhaust processes is difficult in a series of the iterative calculation of a suite of integrated divertor codes SONIC. In order to overcome such a problem, a backflow model has been developed. Amount of the backflow flux from the sub-divertor chamber to the divertor region is evaluated in advance, and then simulating impurity flux is injected from the exhaust slot to the divertor region like a backflow. By this model, the MC calculation time is reduced significantly and iterative calculation of SONIC becomes possible within a reasonable calculation time. As a demonstration, the SONIC code with the backflow model has been applied to investigation of power handling in JT-60SA divertor. The SONIC simulation showed that low divertor heat load (< 10 MW/m2) with the low SOL density (< 1.5 × 1019m–3), which is required in the full non-inductive current drive scenario, was achieved by the Ar gas puffing of 0.86 Pa m3/s. (© 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

核融合原型炉において中心ソレノイド磁束消費を最小化しながら多様な安全係数分布のプラズマを得るために、電子温度分布の加熱フィードバックを通じた安全係数分布制御を検討した。その結果、正磁気シア、弱磁気シア、負磁気シアの幅広い範囲の安全係数分布を持つプラズマが経験則に基づく予測の60%程度の抵抗性磁束消費で生成できることが明らかになった。目標電子温度の最適化と加熱フィードバックゲインの最適化に、強化学習の手法を用いた。実効電荷分布がランダムに選ばれかつそれが直接観測できない状況でも、目標電子温度分布を適応的に最適化し、低磁束消費のq分布制御が実現できることを示した。

The JT-60SA divertor design has been established on the basis of engineering requirements and physics analysis. Heat and particle fluxes under the full input power of 41 MW can give severe heat loads on the divertor targets, while the allowable heat load is limited below 15 MW/m 2 . Dependence of the heat flux mitigation on a D 2 gas-puff is evaluated by SONIC simulations for high density ( n e_ave ∼ 1 × 10 20 m −3 ) high current plasmas. It is found that the peak heat load 10 MW/m 2 with dense ( n ed > 4 × 10 20 m −3 ) and cold ( T ed , T id ⩽ 1 eV) divertor plasmas are obtained at a moderate gas-puff of Γ puff = 15 × 10 21 s −1 . Divertor plasmas are controlled from attached to detached condition using the divertor pump with pumping-speed below 100 m 3 /s. In full non-inductive current drive plasmas with low density ( n e_ave ∼ 5 × 10 19 m −3 ), the reduction of divertor heat load is achieved with the Ar injection.

The identification of the maximum amplitude of the currents circulating in the circuits is a useful indication for the design both of magnet and power supply components in fusion experiments. This paper evaluates the maximum level of coil overcurrents in the poloidal superconducting magnets of JT-60SA, the satellite tokamak that will be built in Naka, Japan, in the framework of EU-JA “Broader Approach” Agreement and that is expected to perform first plasma on 2016. To derive these information, a complete model capable to take into account all the mutually coupled elements was worked out, including the poloidal superconducting coils, the plasma position control in-vessel coils, the vacuum vessel, the stabilizing plates and the plasma. The model was utilized to analyze plasma disruption and quench protection circuit intervention in a large variety of different conditions to identify the possible overcurrent levels. The paper describes the model and the analyses performed, and presents and discusses the results.

The integrated framework consisting of TOPICS, VMEC and FORTEC-3D, has been applied to provide an estimate of the neoclassical toroidal viscosity (NTV) produced by the applied magnetic perturbations (MPs). Six picture-frame error field correction coils in each row in JT-60SA realize the various patterns of the applied magnetic field spectrum. The NTV is investigated for various toroidal phase differences. As long as the same coil current is applied, the n = 3 MPs generate the same NTV, regardless of even or odd parity, or energizing the middle coils or not. The n = 1 MPs give rise to larger NTV than the n = 3 ones. The NTV invoked by the n = 1 resonant MPs is virtually equivalent to that by the n = 1 nonresonant MPs with five-fold coil current.

The integrated framework consisting of TOPICS, VMEC and FORTEC-3D, has been applied to provide an estimate of the neoclassical toroidal viscosity (NTV) produced by the applied magnetic perturbations (MPs). Six picture-frame error field correction coils in each row in JT-60SA realize the various patterns of the applied magnetic field spectrum. The NTV is investigated for various toroidal phase differences. As long as the same coil current is applied, the n = 3 MPs generate the same NTV, regardless of even or odd parity, or energizing the middle coils or not. The n = 1 MPs give rise to larger NTV than the n = 3 ones. The NTV invoked by the n = 1 resonant MPs is virtually equivalent to that by the n = 1 nonresonant MPs with five-fold coil current.

The integrated modelling code TOPICS has been extended to include the core impurity transport, and applied to predictive modelling of JT-60SA high-beta steady-state plasma with the accumulation of impurity seeded to reduce the divertor heat load. In the modelling, models and conditions are selected for a conservative prediction, which considers a lower bound of plasma performance with the maximum accumulation of impurity. The conservative prediction shows the compatibility of impurity seeding with the core plasma with high-beta (βN > 3.5) and full current drive conditions, i.e., when the Ar seeding reduces the divertor heat load below 10 MW/m2, its accumulation in the core is so moderate that the core plasma performance can be recovered by additional heating within the machine capability to compensate the Ar radiation. Due to the strong dependence of accumulation on the pedestal density gradient, the high separatrix density is important for the low accumulation as well as the low divertor heat load. The conservative prediction also shows that JT-60SA has enough capability to explore the divertor heat load control by impurity seeding in high-beta steady-state plasmas.

The Magneto-acoustic Cyclotron Instability (MCI) is a possible emission mechanism for Ion Cyclotron Emissions (ICEs). A dispersion model of the MCI driven by a drifting-ring-type ion velocity distribution has been proposed. In this study, the model was compared with the experimental observations of 3He ICEs [ICEs(3He)] on JT-60U. For this purpose, at first, velocity distributions of deuterium–deuterium fusion produced fast 3He ions at the time of an appearance of the ICE(3He) were evaluated by using a fast ion orbit following code under a realistic condition. The calculated distribution at the edge of the plasma on the midplane on the low field side is shown to have an inverted population and strong anisotropy. This distribution can be reasonably approximated by the drifting-ring-type distribution. Next, dispersions of the MCIs driven by the drifting-ring-type distribution were compared with those of observed ICEs(3He). The comparison shows that toroidal wavenumbers and frequencies of the calculated MCIs ...

The application of the electron cyclotron heating (ECH) and electron cyclotron current drive (ECCD) to the JT-60U tokamak started in 1999. Because the power deposition by the electron cyclotron wave is very localized and controllable, the application ofECH/ECCD has been very attractive for the following recent studies in the JT-60U: (a) the extension of plasma performance toward high normalized beta (β N ), (b) high bootstrap current fraction, and (c) long-pulse operation (65 s). Plasma produced in the studies aiming at advanced steady tokamak is considered to be in a kind of "self-organized state" with external input power by joule heating plus additional heating. The internal transport barrier that develops by the additional heating enhances the local bootstrap current by steep pressure gradient, and then the modified plasma current profile establishes a different confining poloidal magnetic field configuration from the initial configuration. In such experimental research in the JT-60U, the ECH contributes as an active tool for the plasma control to study the physical mechanisms of high-IS magnetohydrodynamic instability, internal transport barrier, current hole, and so on. Results of the ECH/ ECCD applications in the JT-60U are briefly reviewed.

Impact of the electron cyclotron range of frequency wave (ECRF) on the internal transport barriers (ITBs) in a weak shear (WS) plasma has been investigated in JT-60U. The fundamental O-mode ECRF of 110 GHz injected obliquely (co-current drive) from the low field side is used. It is observed that the ion temperature (Ti) ITB in a WS plasma can be degraded by ECRF. It is clarified for the first time that the degradation depends increasingly on the EC power (PEC) but decreasingly on the plasma current (Ip). Moreover it is confirmed that ECRF affects the toroidal rotation (Vt) indirectly and results in the flattening of Vt(ρ) and therefore the radial electric field (Er) profiles regardless of the direction of the target Vt(ρ), peaking co or counter direction (relative to the Ip direction). Furthermore, it is recently found that Ti and Vt in the whole ITB region are affected with almost no delay from the EC onset even with off-axis EC deposition. These results indicate that EC injection unveiled a semi-global structure that characterizes Ti ITB in a WS plasma.

Controllability of plasmas with a large bootstrap current fraction (fBS) has been investigated in JT-60U. Real time control logic for avoidance of collapses by pressure profile control through the toroidal rotation has been newly installed for the long sustainment of reversed magnetic shear plasmas. The new real time control logic can control the injection timing of the neutral beam based on the real time detection of the minimum value of safety factor (qmin) using the motional Stark effect diagnostic. Using the real time control logic, the weak reversed magnetic shear plasma with fBS ~ 70% is sustained for ~8 s. In such a plasma, dynamic change in the current profile, especially qmin, which was induced by the change in the pressure profile at the internal transport barrier (ITB) through the rotation control, was observed. The response of qmin to the change in the ion temperature gradient at the ITB for large fBS plasmas is stronger than that for smaller fBS plasmas.

For the purpose of stable plasma equilibrium control and detailed analysis, it is essential to reconstruct an accurate plasma boundary on the poloidal cross section in tokamak devices. The Cauchy condition surface (CCS) method is a numerical approach for calculating the spatial distribution of the magnetic flux outside a hypothetical surface and reconstructing the plasma boundary from the magnetic measurements located outside the plasma. The accuracy of the plasma shape reconstruction has been assessed by comparing the CCS method and an equilibrium calculation in JT-60SA with a high elongation and triangularity of plasma shape. The CCS, on which both Dirichlet and Neumann conditions are unknown, is defined as a hypothetical surface located inside the real plasma region. The accuracy of the plasma shape reconstruction is sensitive to the CCS free parameters such as the number of unknown parameters and the shape in JT-60SA. It is found that the optimum number of unknown parameters and the size of the CCS that minimizes errors in the reconstructed plasma shape are in proportion to the plasma size. Furthermore, it is shown that the accuracy of the plasma shape reconstruction is greatly improved using the optimum number of unknown parameters and shape of the CCS, and the reachable reconstruction errors in plasma shape and locations of strike points are within the target ranges in JT-60SA.

Remote operation and data analysis are becoming key requirements of any fusion devices. In this framework a well-conceived data management system integrated with a suite of analysis and support tools are essential components for an efficient remote exploitation of any fusion device. The following components must be considered: data archiving data model architecture; remote data and computers access; pulse schedule, data analysis software and support tools; remote control room specifications and security issues. The definition of a device-generic data model plays also important role in improving the ability to share solution and reducing learning time. As for the remote control room, the implementation of an Operation Request Gateway has been identified as an answer to security issues meanwhile remotely proving all the required features to effectively operate a device. Previous requirements have been analyzed for the new JT-60SA tokamak device. Remote exploitation is paramount in the JT-60SA case which is expected to be jointly operated between Japan and Europe. Due to the geographical distance of the two parties an optimal remote operation and remote data-analysis is considered as a key requirement in the development of JT-60SA. It this case the effects of network speed and delay have been also evaluated and tests have confirmed that the performance can vary significantly depending on the technology used.

The ITER Remote Experimentation Centre (REC) is one of the projects under implementation within the BA agreement. The final objective of the REC is to allow researchers to take part in the experimentation on ITER from a remote location. Before ITER first operations, the REC will be used to evaluate ITER-relevant technologies for remote participation. Among the different software tools needed for remote participation, an important one is the Remote Data Access System (RDA), which provides a single software infrastructure to access data stored at the remotely participating experiment, regardless of the geographical location of the users. This paper introduces the European contribution to the RDA system for the REC.

Characteristics of the edge pedestal has been studied for the high βp ELMy H mode and the reversed shear ELMy H mode in JT-60U. The contribution of the pedestal is around 20% for the confinement improvement factor, the normalized beta, and the bootstrap fraction. The upper boundary of the fusion performance measure increases with the pedestal poloidal beta, βp-ped. At high triangularity, δ, βp-ped increases with the total βp (βp − tot) almost linearly for positive shear type I ELMy discharges. This dependence is not due to the profile stiffness, since the dependence is the same for the discharges both with and without an internal transport barrier (ITB). In the reversed shear ELMy H mode, βp-ped increases with βp-tot on the same line as the positive shear cases except at high q95 = 8–9.3. In the high plasma current regime, the final structure with both an ITB and edge transport barrier seems to be determined by the balance between the expanding ITB-foot radius and deepening ELM penetration: the ELM penetration radius deepens with increasing pedestal stored energy and then reaches the ITB-foot radius. The ITB-foot seems to behave as a barrier against the ELM crash penetration, and shrinks after a few ELM attacks, and the ELM penetration follows the shrinking ITB-foot.

The spontaneous toroidal rotation velocity under the no/low direct toroidal momentum input, particularly from the electron cyclotron (EC) wave injection has been investigated in JT-60U plasmas. It is found that the response of toroidal rotation velocity to the central EC injection is towards the co-current direction in L-mode plasma. The region of the change in the toroidal rotation velocity is wider than the EC deposition profile and similar to that in electron temperature. The observed co-rotation velocity in the combined heating of EC and lower hybrid wave increases with the increase in the stored energy and a large positive radial electric field is formed in the strong co-rotating plasma. Furthermore the short pulse off-axis EC injection experiment shows that the perturbation of the toroidal rotation velocity towards the co-direction propagates to the centre with increase in its amplitude, suggesting an inward pinch in momentum transport.

The physics of strong internal transport barriers (ITBs) in JT-60U reversed-magnetic-shear (RS) plasmas has been studied through the modelling on the 1.5 dimensional transport simulation. The key physics to produce two scalings on the basis of the JT-60U box-type ITB database are identified. As for the scaling for the narrow ITB width proportional to the ion poloidal gyroradius, the following three physics are important: (1) the sharp reduction of the anomalous transport below the neoclassical level in the RS region, (2) the autonomous formation of pressure and current profiles through the neoclassical transport and the bootstrap current and (3) the large difference between the neoclassical transport and the anomalous transport in the normal-shear region. As for the scaling for the energy confinement inside ITB (∈ f β p,core ≈ 0.25, where ∈ f is the inverse aspect ratio at the ITB foot and β p,core is the core poloidal beta value), the value of 0.25 is found to be a saturation value due to the MHD equilibrium. The value of ∈ f β p,core reaches the saturation value, when the box-type ITB is formed in the strong RS plasma with a large asymmetry of the poloidal magnetic field, regardless of the details of the transport and the non-inductively driven current.