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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.

We have developed a denoising autoencoder based neural network (NN) method to determine a spectral line intensity with an uncertainty lower than the uncertainty determined by fitting the spectral line. The NN method processes the measured raw spectral line shape, providing a single Gaussian shape based on the training dataset, which consists of synthetically prepared Doppler shift and broadening free spectral lines in the present work. It is found that the uncertainty reduction level significantly depends on the training dataset. Limitations originating from the training dataset are also discussed.

Explosive dynamical events in controlled-nuclear-fusion devices (known as edge-localized modes) display many similarities to solar-flare events on the sun, revealing a new connection between laboratory plasma physics and astronomy. However, to date there has been no direct evidence for the peeling mode structure, due to the lack of decisive diagnostics. Here we report the first evidence for the structure and dynamical steady state of a peeling mode for low-n edge-harmonic oscillations (EHOs) in the quiescent H-mode. EHOs are dominated by the fundamental mode (1fEHO) at both the low- and high-field sides. 1fEHO edge perturbations are confirmed to have kink parity and exhibit the frozen-in-condition predicted by a linear stability analysis. The envelope signal of the 1fEHO mode exhibits repeated cycles of growth and damping to the order of a few hundred Hz associated with small changes in an edge gradient, and results are quantitatively consistent with a limit-cycle-oscillation model.

JT-60SAでは電子温度と電子密度の分布を計測するためにトムソン散乱計測システムを設置する。本発表では、ポートプラグ内の限られた空間に設置可能な集光光学系、レーザー伝送系、分光器、制御システムの概念設計を示す。集光光学系はポートプラグからは独立した片持ち張り構造とし、ディスラプション等が原因となる変位の影響を抑制する。電子温度の計測誤差は、10%以下であると評価され、プラズマ研究に十分な精度で計測できる見込みを得た。, 第23回高温プラズマ計測に関するトピカル会議(HTPD)

NIFSで開発された日本独自の技術であるトレーサー内蔵固体ペレット(TESPEL)を日欧協力で実施しているJT-60SA装置に入射するためのTESPEL入射装置の開発を進めている。発表では、JT-60SAプラズマに不純物を供給するために必要なTESPEL及びTESPEL入射装置の最適化研究の結果及び最新の仕様について報告した。, 第37回プラズマ・核融合学会年会

Disruption is a critical phenomenon in a tokamak reactor. Although disruption causes serious damage to the reactor, its physical mechanism remains unclear. To realize a tokamak reactor, it is necessary to understand and control the disruption phenomenon. The present research constructs a disruption predictor using experimental high-beta plasma data in the JT-60U tokamak. The predictor was constructed using a support vector machine as a linear discriminant, and we focus on a variable selection problem for the binary classification by sparse modeling, specifically, exhaustively searching the best combinations of variables which maximize the predictor performance. By the sparse modeling, we found that the six input parameters as the best combinations. The selected parameters were the n = 1 mode amplitude | B r n = 1 | and its time derivative d | B r n = 1 | / d t , the plasma density (relative to the Greenwald density limit) and its time derivative, and the time derivatives of the plasma internal inductance and plasma elongation. In particular, it was identified that the parameter d | B r n = 1 | / d t , plays a key role on plasma disruption. We should notice that the combination with other plasma parameters is indispensable and remarkably make it possible to improve the performance of disruption prediction.

Helical cores (HCs) in plasmas are one of the important phenomena because the helical structure affects impurity transport and induces loss of confined energetic particles or loss of toroidal momentum. Substantial studies on experimental observation of HCs were reported, because HCs would be excited in hybrid scenarios in ITER or DEMO. In addition to the effect on transport by HCs, recent studies suggested that MagnetoHydroDynamics (MHD) dynamo accompanied with HCs redistributes the current profile in the core and realizes the experimentally observed sawtooth-free plasmas of hybrid scenarios in DIII-D, in which the minimum value of the safety factor profile, qmin, is kept slightly above unity by "flux pumping". Several theoretical models to explain the excitation mechanism of HCs have been proposed, such as the generation of the magnetic island by the excitation of tearing modes (TMs) or HC equilibria induced by internal kink modes. Though substantial experimental observations were reported in previous studies, the experimental observations focusing on the detailed relationship between the mode structures of the HC and other MHD modes have not been reported well. In this paper, we report experimental finding of n = 1 HCs accompanied with saturated m/n = 2/1 TMs having low rotation frequencies in JT-60U with various diagnostics. Here, m is the poloidal mode number and n is the toroidal mode number. The TMs with HCs are observed after an increase of the mode amplitude and a decrease of the mode frequency of m/n = 2/1 precursors having the tearing parity. The decreased mode frequency is lower than 20 Hz typically. With various diagnostics, the coupling of n = 1 helical cores and m/n = 2/1 TMs have been clearly observed. Because the coherent oscillations in the ion temperature are observed in the core region and in the edge region, the flux surfaces including the m/n = 2/1 magnetic island appear to have m = 1 helical deformation. It has also been suggested that the m/n = 2/1 TM and the HC rotate in the electron diamagnetic direction keeping fm/n=1/1(HC) = 2fm/n=2/1(TM) in several plasmas even with different bulk rotation speed. Here, the fm/n=1/1(HC) is the rotation frequencies of HCs and fm/n=2/1(TM) is the rotation frequency of TMs. In addition, the core seems to be shifted to the high field side when the “O” points of the m/n = 2/1 magnetic island line up in the midplane, which is confirmed by the reconstruction of MHD equilibria with Motional Stark Effect measurement and MEUDAS code. Our observations on m/n = 2/1 TMs having HCs would contribute to the understanding of the excitation mechanism of HCs in tokamak plasmas., The 28th International Toki Conference on Plasma and Fusion Research

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.

An incoherent Thomson scattering diagnostic will be installed in the JT-60SA tokamak to measure electron temperature and electron density profiles. The target radial spatial resolution is 25 mm with 46 spatial channels. The accuracy in electron temperature and density is a few percent at ne = 7.5 × 1019 m-3, which is the expected value in the plasma core. This paper presents the designs of collection optics, fibers with their alignment system, and polychromators. The collection optics overcomes unique issues for superconducting fusion devices, i.e., limited design space, high-temperature measurements, and harsh radiation condition. When in several years the more performing plasma will generate intense nuclear radiation, the lens materials of the optics can be replaced by radiation resistant glasses without major changes in the lens holder. It will prevent transmission degradation and keep stable measurement accuracy.

Plasma with an internal transport barrier (ITB) is desirable for a steady-state tokamak reactor because of its high confinement quality and high bootstrap current fraction. However, the local pressure gradient tends to be steep and the plasma often becomes unstable. In this study, an ion temperature gradient control system based on neutral beam injection (NBI) is developed using the reinforcement learning technique. The response characteristics of an ion temperature gradient to NBI are non-linear and sensitive to experimental conditions, which makes it difficult to develop a robust control system. Our control system is trained for plasmas with a wide range of ITB strengths. Using the reinforcement learning technique, the system acquires a robust control feature through several thousand iterations of trial and error in an integrated transport simulation hosted by TOPICS. The control system is composed of neural networks (NNs) whose input variables are the ion temperature gradient, the current NBI power, and the NBI powers for several previous control time steps. The trained system can determine a control output which is suitable for the response characteristics inferred from the input variables. The trained control system is tested in the TOPICS simulation using plasma models based on two experimental plasmas of JT-60U with different ITB strengths. It is shown that the ion temperature gradient can be appropriately controlled for both plasmas, which supports the expectation that this system is applicable to real experiments.

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).

Materials and structures of a collimator for a new neutron emission profile monitor in JT-60SA are examined through Monte Carlo simulations using the Monte Carlo N-Particle transport code. First, the shielding properties of various material combinations are compared in order to determine a combination with high shielding performances against both neutrons and gamma-rays. It is found that a collimator consisting of borated polyethylene and lead has a high shielding performance against neutrons. Moreover, a high shielding performance against gamma-rays is obtained when a lead pipe with a radial thickness of 0.01 m is inserted into a collimation tube. Second, we demonstrate that it is possible to improve the spatial resolution to a desired level by installing a thin tubular extension structure that fits into the limited space available between the main collimator block and the tokamak device. Finally, the collimator structures that meet both the targeted spatial resolutions (

A field of view (FoV) of the resistive bolometer systems specialized for the total radiated power analysis has been designed with bolometer systems having multiple-apertures in JT-60SA. The multi-aperture bolometer system can reduce the required number of bolometer channels for the determination of the total radiated power. The FoV design has been carried out to minimize unobserved and overlapping regions with the following requirements: (i) independent determination of the divertor and the main plasma radiation, (ii) limitation of the use of only three diagnostic ports, and (iii) a redundant FoV design. With the designed FoV, the whole plasma of JT-60SA is viewed by only five bolometers and five replicated bolometers having the same FoV for redundancy. A signal evaluation with the designed FoV with a predetermined radiation profile indicates that even the lowest performing bolometer exhibits a signal to noise (S/N) ratio of 300 or more. The evaluated S/N ratio is high enough for the total radiated power analysis. Using the designed FoV, the total radiated power over the plasma profile can be determined within 5.3% using only five bolometers.

JT-60SA will complement ITER in resolving key issues to finally decide an acceptable DEMO design. Diagnostics play a key role in this mission. The electron temperature and density profiles are measured by a core and an edge Thomson scattering (TS) diagnostics with high spatial resolution, needed to identify the pedestal parameters and small profile structures. The two systems use a common tangential Nd:YAG laser beam path in the plasma equatorial plane. The collection optics for the edge system (low field side) is hosted in a lower oblique port and that for the core system in a horizontal port. The optics fit in the port plug tube and image the scattering volumes into an array of fiber bundles. They both are exposed to a high neutron dose of 10^16 n/cm^2 over 13 years of operation. The optics are supported by a mechanical structure decoupled from the cryostat. A set of filter polychromators with avalanche photodiode (APD) detectors spectrally analyze the scattered radiation. The development of the TS systems is carried out by a joint Japan-EU team. The conceptual design of the edge TS system is presented here. Simulations of the TS signals show acceptable accuracy down to 1 × 10^19 m^−3 electron density, sufficient to measure the edge gradient and even a small region outside the separatrix.

Simulations with an integrated code TOPICS-IB showed that a small pellet can significantly reduce the ELM energy loss by penetrating deeply into the pedestal and triggering high-n ballooning modes localized near the pedestal top, with conditions; the injection from the low-field-side with a speed fast enough to approach the pedestal top when the pedestal pressure is about 95% of natural ELM onset. The effectiveness of the above suitable conditions of pellet injection for ELM pacing has been confirmed by JT-60U and then ITER simulations. The pellet particle content required for ELM pacing is larger for the pedestal plasma with higher density and farther from the stability boundary of ideal ballooning mode near the pedestal top. For an ITER standard scenario, the required pellet particle content is about a few % of pedestal particle content, which gives the physics background to the present design value. Simulations also showed that fueling pellets can be injected from the high-field-side just after ELM pacing pellets without disturbing the pacing. (© 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

Scaling of the density peaking for positive magnetic shear ELMy H-mode plasmas in JT-60U has been developed. Although the density peaking generally depends on collisionality, a variation of the density peaking factor for the same collisionality exists, and the variation is different between plasmas with co-directed and ctr-directed toroidal rotation(co-V T plasmas and ctr-V T plasmas). The variation of the co-V T plasmas can be explained by particle source profile. As a result of scaling, the peaking factor of the co-V T plasmas depends on the particle source rate profile from neutral beams (NBs) as much as collisionality. The dataset of the ctr-V T plasmas has a larger variation of the peaking factor compared to that of the co-V T plasmas. The larger variation stems from that the peaking factor of the ctr-V T plasmas also depends on the normalized ion temperature gradient at the edge region. As a result of scaling, the normalized ion temperature gradient influences the peaking factor as much as or a little larger than collisionality. The parameter regime of the ctr-V T plasmas ranges from the trapped electron mode (TEM) to the ion temperature gradient mode (ITG mode). The plasmas with the positive correlation between the normalized density gradient and the normalized ion temperature gradient are dominated by the TEM. On the other hand, the plasmas with the negative correlation are dominated by the ITG mode.

We report the experimental finding of n = 1 helical cores (HCs) accompanied by saturated m/n = 2/1 tearing modes (TMs) with low mode frequencies in JT-60U. The HCs accompanied by TMs were observed after an increase in the mode amplitude and a decrease in the mode frequency of m/n = 2/1 precursors with tearing parity. The decreased mode frequency is typically lower than 20 Hz. With various diagnostics, the coupling of n = 1 HCs and m/n = 2/1 TMs has been clearly observed. Because the coherent oscillations in the ion temperature are observed in both the core region and the edge region, the flux surfaces including the m/n = 2/1 magnetic island appear to have m = 1 helical deformation. It has also been suggested that the m/n = 2/1 TM and the HC rotate in the electron diamagnetic direction keeping fm/n=1/1(HC) = 2fm/n=2/1(TM) in several plasmas. Here, fm/n=1/1(HC) is the mode frequency of HCs and fm/n=2/1(TM) is the mode frequency of TMs. In addition, the core seems to be shifted to the high-field side when the O-points of the m/n = 2/1 magnetic island line up in the midplane, which is confirmed by reconstructions of magnetohydrodynamic equilibria with motional Stark effect measurement and the MEUDAS code. Our observation of m/n = 2/1 TMs having HCs contributes to the understanding of the excitation mechanism of HCs in tokamak plasmas.

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。

Edge localized mode (ELM) must be eliminated which enhances the erosion of divertor plates in the Hmode operation of tokamak reactors. Suppression of ELM has been experimentally achieved by the resonant magnetic perturbation (RMP) with multipartite coils. In a DEMO reactor with strong neutron flux, however, it is desired the coils near the first wall not to be put in. We propose an innovative concept of the RMP for tokamak DEMO reactors without installing coils but inserting ferritic steels of the helical configuration. Helically perturbed magnetic field is naturally formed in the axisymmetric toroidal magnetic field through the helical ferritic steel inserts (FSIs). The perturbation amplitude in the plasma pedestal region can easily be set above several 10-4 of the toroidal field strength in the DEMO reactor condition, which is enough for the RMP to mitigate/suppress ELMs.

Radiation distribution and divertor detachment in argon impurity seeding discharges were investigated in JT-60U. A large peak in the radiation power profile was observed at the edge. A radiation power in the core plasma ( ρ / a ⩽ 0.93–0.95) and power flux to the edge pedestal were evaluated, and transition of ELM characteristics from Type-I to Type-III was seen at an exhaust power of 1.5–1.8 times larger than the L–H threshold power scaling. In the Type-III ELM plasma, the outer divertor plasma became partially detached. Intrinsic carbon impurity rather than the seeding argon dominated the radiated power in the divertor, which was decreased in the detached divertor due to a reduction in carbon influx. During re-attachment, ion saturation current and electron temperature at the outer divertor plasma increased with similar time scales (0.3–0.4 s) of decreasing Ar impurity at the edge and the changing ELM activity.

Two design techniques for the field of view (FoV) of a resistive bolometer system for the analysis of the total radiated power have been developed. The first technique is a new concept of the resistive bolometer system having multiple apertures to achieve a large viewing angle with one bolometer channel. The second technique is an optimization technique of the bolometer location in the major radial direction to minimize error of the analysis. Both techniques have been examined with a virtual tokamak device. By a FoV designed with both techniques, the total radiation power of several radiation profiles can be determined within 3% error of the analysis with one bolometer channel.

Microwave to millimeter-wave diagnostic techniques such as interferometry, reflectometry, scattering and radiometry have been powerful tools for diagnosing magnetically confined plasmas. Recent advances in electronic devices and components together with computer technology have enabled the development of advanced diagnostics in microwave to millimeter-wave region. The two-dimensional and three-dimensional profiles of plasma density/temperature and dynamic behaviors of their fluctuations are measured by using the advanced diagnostic systems. The measurements have clarified the physics issues such as stability, wave phenomena and fluctuation-induced transport. We report here on the representative microwave diagnostics and their contribution to plasma confinement physics.

By using a fully three dimensional magnetic field orbit-following Monte-Carlo code, the energetic ion confinement was investigated for the current conceptual design of the ferromagnetic components in ITER which will be employed for reducing the toroidal magnetic field (TF) ripple. The ferromagnetic insert is effective in the reference standard scenario with Q = 10 (Scenario No. 2) and steady state scenario with Q = 5 (Scenario No. 4) to improve the energetic ion confinement. Over-compensation appears at half of the full toroidal magnetic field and its effect becomes stronger when the quantity of the ferromagnetic insert is increased in order to more reduce the TF ripple at the full toroidal magnetic field. Though the current design is acceptable, whether to increase the ferromagnetic insert to achieve lower TF ripple amplitude at the full field operation depends on how prospected are possibilities of lower field operations. Planned test blanket modules do not induce large loss (

The design activity of JT-60SA (JT-60 Super Advanced) addresses the remodeling of JT-60 to a superconducting tokamak device under the JA-EU collaborative ITER–BA project. For the JT-60SA control system, the existing system should be reused as much as possible from the viewpoint of cost-effectiveness. Initial considerations of the configuration of the advanced supervisory control system (SVCS) including the following systems are: (1) an ultimate flexibility real-time control system, (2) a precision timing system able to diagnose a plasma event having a time range of ∼1 μs, and (3) a safety shutdown control system. In this report, we present the design study of the JT-60SA SVCS with the focus on the above systems.

The effect of the pressure profile on the energy loss caused by edge localized modes (ELMs) has been investigated by using an integrated simulation code TOPICS-IB based on a core transport code with a stability code for the peeling-ballooning modes and a transport model for scrape-off-layer and divertor plasmas. The steep pressure gradient inside the pedestal top is found to broaden the region of the ELM enhanced transport through the broadening of eigenfunctions and enhance the ELM energy loss. The ELM energy loss in the simulation becomes larger than 15 % of the pedestal energy, as is shown in the database of multi-machine experiments. (© 2008 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

The dynamics of SOL-divertor plasmas after an ELM crash is studied with a one-dimensional particle simulation code, PARASOL. The ELM crash occurs off-centrally in the SOL region, and the ELM heat flux to the near divertor plate and that to the far divertor plate become asymmetric. The peak heat flux to the near plate is larger as compared to the far plate. The asymmetry in the peak heat flux increases with the connection-length ratio. The imbalance in the heat deposition, however, is small. The radial transport loss of ELM flux creates the asymmetry in the heat deposition, but the imbalance is still not large even for the large radial transport loss rate. The electron heat flux to the far plate brought by the SOL current is one of the causes of a small imbalance in the heat deposition. Another cause is the asymmetric SOL flow and its convective heat flux, whose stagnation point stays for a long period near the ELM crash location. (© 2008 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

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.

This paper summarizes results of dimensionless identity experiments in JT-60U and JET, aimed at the comparison of the H-mode pedestal and ELM behaviour in the two devices. Given their similar size, dimensionless matched plasmas are also similar in their dimensional parameters (in particular, the plasma minor radius a is the same in JET and JT-60U). Power and density scans were carried out at two values of Ip, providing a q scan (q95 = 3.1 and 5.1) with fixed (and matched) toroidal field. Contrary to initial expectations, a dimensionless match between the two devices was quite difficult to achieve. In general, pped in JT-60U is lower than in JET and, at low q, the pedestal pressure of JT-60U with a Type I ELMy edge is matched in JET only in the Type III ELM regime. At q95 = 5.1, a dimensionless match in ?*, ?* and ?p,ped is obtained with Type I ELMs, but only with low power JET H-modes. These results motivated a closer investigation of experimental conditions in the two devices, to identify possible 'hidden' physics that prevents obtaining a good match of pedestal values over a large range of plasmas parameters. Ripple-induced ion losses of the medium bore plasma used in JT-60U for the similarity experiments are identified as the main difference with JET. The magnitude of the JT-60U ripple losses is sufficient to induce counter-toroidal rotation in co-injected plasma. The influence of ripple losses was demonstrated at q95 = 5.1: reducing ripple losses by ?2 (from 4.3 to 1.9?MW) by replacing positive with negative neutral beam injection at approximately constant Pin resulted in an increased pped in JT-60U, providing a good match to full power JET H-modes. At the same time, the counter-toroidal rotation decreased. Physics mechanisms relating ripple losses to pedestal performance are not yet identified, and the possible role of velocity shear in the pedestal stability, as well as the possible influence of ripple on thermal ion transport are briefly discussed. Toroidal rotation of the ITER reference inductive Q = 10 H-mode is predicted to be rather low, of the order of ~1/10 of the frequency of typical JET H-modes. Nonetheless, fast ion ripple losses in that scenario are also predicted to be negligible (1%), and therefore plasma toroidal rotation slow-down or ripple-induced counter-rotation should not affect pedestal parameters and stability in ITER. Finally, the possible effect of ripple on thermal transport may deserve more attention in future experiments and modelling, since the ripple magnitude of ITER is intermediate between that of JET and JT-60U.

An integrated simulation code TOPICS-IB based on a transport code with a stability code for the peeling–ballooning modes and a scrape-off-layer (SOL) model has been developed to clarify self-consistent effects of edge localized modes (ELMs) and the SOL on the plasma performance. Experimentally observed collisionality dependence of the ELM energy loss is found to be caused by both the edge bootstrap current and the SOL transport. The bootstrap current decreases with an increase in collisionality and intensifies the magnetic shear at the pedestal region. The increase in the magnetic shear reduces the width of eigenfunctions of unstable modes, which results in the reduction of both the area of the ELM enhanced transport and the ELM enhanced transport near the separatrix. On the other hand, when an ELM crash occurs, the energy flows into the SOL and the SOL temperature rapidly increases. The increase in the SOL temperature lowers the ELM energy loss due to the flattening of the radial edge gradient. The parallel electron heat conduction determines how the SOL temperature increases. For higher collisionality, the conduction becomes lower and the SOL electron temperature increases more. By the above two mechanisms, the ELM energy loss decreases with increasing collisionality.

Long-spatial-scale density fluctuations with a wave number of the order of 1 cm−1 were suppressed in the internal transport barrier (ITB) region of a JT-60U reversed magnetic shear plasma, after a cold pulse was induced by the shallow pellet injection at the plasma edge. The particle and ion heat diffusivities are reduced with the suppressed fluctuations. The reduction of the electron heat diffusivity is observed only transiently in the inner ITB portion around the layer of the density fluctuation measurement during the cold pulse propagation and it is slower (~ 15 ms) than the reduction of the density fluctuation level (< 5 ms). In the outer ITB portion, the electron heat diffusivity increases transiently. These results indicate decoupling of the electron heat transport and the long-spatial-scale density fluctuations.

A recent series of dimensionless pedestal identity experiments at JET and JT-60U failed to produce a match in the dimensional pedestal parameters and edge-localized mode (ELM) frequency despite a good match in the main dimensionless plasma parameters. This paper describes the progress made in understanding these experimental results. First, it is investigated whether differences in the magnetohydrodynamic stability of the pedestal, including those potentially arising from the 10% difference in the aspect ratio between the two tokamaks, can explain the results. The potential effects of differences in plasma rotation between the two machines are also examined. Given the result that these mechanisms fail to explain the experimental observations and the fact that JT-60U features considerably stronger toroidal magnetic field ripple than JET, the bulk of the paper, however, discusses the effects of ripple losses. The analysis shows that ripple losses of thermal ions can affect H-mode plasma performance very sensitively. Orbit-following simulations indicate that losses due to diffusive transport give rise to a wide radial distribution of enhanced ion thermal transport, whereas non-diffusive losses have a very edge-localized distribution. In predictive transport simulations with an energy sink term in the continuity equation for the ion pressure representing non-diffusive losses, reduced performance as well as an increase in the ELM

A strategy for integrated modeling of burning plasmas at Japan Atomic Energy Agency is described. In order to simulate the burning plasma, which has the complex feature of widely different timescal...

The degradation of energy confinement with increased toroidal beta βT was shown by non-dimensional analysis in JT-60U. The dependence of the energy confinement on βT was examined by both the JT-60U ELMy H-mode confinement database and the dedicated experiment on a single βT scan while and ν* were kept fixed as well as the other magnetic geometrical parameters. In both cases, the degradation of energy confinement with increasing βT was observed, satisfying the relation of . This dependence is a little weaker than that predicted by the IPB98(y, 2) scaling. The fusion power production rate was estimated to increase in proportion to .

Dependence of the energy confinement in ELMy H-mode tokamak on the beta has been investigated for a long time, but a common conclusion has not been obtained so far. Recent non-dimensional transport experiments in JT-60U demonstrated clearly the beta degradation. A database for JT-60U ELMy H-mode confinement was assembled. Analysis of this database is carried out, and the strong beta degradation consistent with the above experiments is confirmed. Two subsets of ASDEX Upgrade and JET data in the ITPA H-mode confinement database are analysed to find the origin of the various beta dependences. The shaping of the plasma cross section, as well as the fuelling condition, affects the confinement performance. The beta dependence is not identical for different devices and conditions. The shaping effect, as well as the fuelling effect, is a possible candidate for causing the variation of beta dependence.

Thermal properties of the redeposition layer on the inner plate of the W-shaped divertor of JT-60U have been measured with laser flash method so as to estimate transient heat loads onto the divertor. Morphology analysis of the redeposition layer was conducted with a scanning electron microscope. Measurement of a redeposition layer sample of more than 200 μm thick, which had been produced near the most frequent striking point, showed following results: (1) the bulk density of the redeposition layer is about half of that of carbon fiber composite material; (2) the specific heat of the layer is roughly equal to that of the isotropic graphite; (3) the thermal conductivity of the redeposition layer is two orders of magnitude smaller than that of the carbon fiber composite. This low thermal conductivity of the redeposition layer is considered to be caused by a low graphitization degree of the redeposition layer. The difference between the divertor heat loads and the loss of the plasma stored energy becomes smaller taking account of thermal properties of the redeposition layer on the inner divertor, whereas estimated heat loads due to the ELMs is still larger than the loss. This is probably caused by the poloidal distribution of the thermal properties.

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.

Further transport reduction is induced by a pellet or electron cyclotron (EC) wave injection after the formation of an internal transport barrier in JT-60U reversed shear plasmas. After pellet injection, the O-mode reflectometer signal drastically decreases, indicating reduction in the electron density fluctuation level. The ion thermal diffusivity decreases by one order of magnitude and reaches the neoclassical level. The effective particle diffusivity also decreases by one order of magnitude, but the electron thermal diffusivity remains constant. After EC wave injection, when no reduction of the O-mode reflectometer signal is observed, the electron thermal diffusivity decreases by a factor of ~2–3, while decreases in the ion thermal diffusivity and the effective particle diffusivity are small. These results indicate that the ion heat transport and particle transport were coupled with the measured density fluctuation, but electron heat transport was decoupled. Ion heat and particle transport and electron heat transport were dominated by different types of fluctuation.