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Partially hydrolyzed guar gum (PHGG), a water-soluble dietary fiber produced by the controlled partial enzymatic hydrolysis of guar gum beans, has various physiological roles. PHGG is expected to influence the immune function and prevent infections. The objective of this study was to examine the effect of continuous ingestion of PHGG for 12 weeks on the development of cold-like symptoms.A placebo-controlled, double blind, randomized, parallel-group comparative study was conducted. 96 healthy Japanese adults received 5.2 g PHGG or placebo daily for 12 weeks. Cold-like symptoms were assessed based on patient diary, and the levels of short-chain fatty acids (SCFAs) in stool and blood immune markers at baseline and at weeks 6 and 12.The cumulative number of "no symptoms" days for all symptoms was significantly larger in the PHGG than in the placebo group. The result of the analysis by severity of cold-like symptoms also showed significant differences, with the PHGG group having a lower severity of cold-like symptoms. Propionic acid at weeks 6 and 12 and n-butyric acid and total SCFAs at week 12 were significantly higher in the PHGG than in the placebo group. The Interferon-γ level was significantly lower at week 6 in the PHGG than in the placebo group.PHGG intake may affect immune function and suppress cold-like symptoms through the production of SCFAs in healthy adults.

Background In the phase III IMpassion130 trial, combining atezolizumab with first-line nanoparticle albumin-bound-paclitaxel for advanced triple-negative breast cancer (aTNBC) showed a statistically significant progression-free survival (PFS) benefit in the intention-to-treat (ITT) and programmed death-ligand 1 (PD-L1)-positive populations, and a clinically meaningful overall survival (OS) effect in PD-L1-positive aTNBC. The phase III KEYNOTE-355 trial adding pembrolizumab to chemotherapy for aTNBC showed similar PFS effects. IMpassion131 evaluated first-line atezolizumab–paclitaxel in aTNBC. Patients and methods Eligible patients [no prior systemic therapy or ≥12 months since (neo)adjuvant chemotherapy] were randomised 2:1 to atezolizumab 840 mg or placebo (days 1, 15), both with paclitaxel 90 mg/m2 (days 1, 8, 15), every 28 days until disease progression or unacceptable toxicity. Stratification factors were tumour PD-L1 status, prior taxane, liver metastases and geographical region. The primary endpoint was investigator-assessed PFS, tested hierarchically first in the PD-L1-positive [immune cell expression ≥1%, VENTANA PD-L1 (SP142) assay] population, and then in the ITT population. OS was a secondary endpoint. Results Of 651 randomised patients, 45% had PD-L1-positive aTNBC. At the primary PFS analysis, adding atezolizumab to paclitaxel did not improve investigator-assessed PFS in the PD-L1-positive population [hazard ratio (HR) 0.82, 95% confidence interval (CI) 0.60-1.12; P = 0.20; median PFS 6.0 months with atezolizumab–paclitaxel versus 5.7 months with placebo–paclitaxel]. In the PD-L1-positive population, atezolizumab–paclitaxel was associated with more favourable unconfirmed best overall response rate (63% versus 55% with placebo–paclitaxel) and median duration of response (7.2 versus 5.5 months, respectively). Final OS results showed no difference between arms (HR 1.11, 95% CI 0.76-1.64; median 22.1 months with atezolizumab–paclitaxel versus 28.3 months with placebo–paclitaxel in the PD-L1-positive population). Results in the ITT population were consistent with the PD-L1-positive population. The safety profile was consistent with known effects of each study drug. Conclusion Combining atezolizumab with paclitaxel did not improve PFS or OS versus paclitaxel alone. ClinicalTrials.gov NCT03125902.

In this study, the clarification of the effect of plasma flow control on the rotor performance was carried out as a wind turbine control technology. The flow fields around rotor and the loads acting on wind turbine were measured by wind tunnel test using a test wind turbine with plasma actuator. As a result, in the low tip speed ratio, the power coefficient of Plasma On is larger than that of Plasma Off and the thrust coefficient of Plasma On slightly increased. In particular, the remarkable increase in the power coefficient was 24% at the most effective tip speed ratio. The wind turbine thrust increased slightly in the low tip speed ratio regime, but was less affected by the plasma flow control. In addition, when the power coefficient and thrust coefficient were calculated using the velocity components of the flow field around the wind turbine, the same tendency as in the wind turbine performance test was shown.

The power output of a wind turbine depends on both wind speed and turbine control. There is a fluctuation in the natural wind speed, which changes the turbine power output. In this paper blade pitch control based on detailed monitoring of the inflow wind, including vertical velocity distributions, has been experimentally evaluated in a field test wind turbine. The inflow wind speeds were observed by 11 sets of ultrasonic anemometers. In this first test, the blade pitch is changed to keep optimum angle of attack based on the local inflow wind speed in the rotor plane. The blade pitch is independently controlled as a reference radial position of 80% of rotor radius. As a result of the pitch controls, the wind turbine power output was increased and the power output fluctuation was decreased for the cyclic pitch control based on the local wind speed. With the cyclic pitch control, the fluctuation in rotor thrust and average rotor thrust were also reduced compere to collective pitch control.

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

We demonstrated an excellent output power (Pout) density performance using a novel InAlGaN/GaN-HEMT with an 80-nm gate for a high-power W-band amplifier. The developed HEMT showed basic reliability for commercial products. A unique double-layer silicon nitride (SiN) passivation film with oxidation resistance was adopted to suppress current collapse. The developed discrete InAlGaN/GaN-HEMT achieved a Pout density of 3.0 W/mm at 96 GHz, and the Pout density of MMIC reached 3.6W/mm at 86 GHz. We proved excellent potential of developed InAlGaN/GaN-HEMT using our unique device technologies. Furthermore, we suggested the physical advantage of the InAlGaN/GaN-HEMT structure using device simulator.

We demonstrated an excellent output power (P out ) density performance using a novel InAlGaN/GaN-HEMT with an 80-nm gate for a W-band amplifier. To eliminate current collapse, a unique double-layer silicon nitride (SiN) passivation film with oxidation resistance was adopted. The developed discrete GaN-HEMT achieved a P out density of 3.0 W/mm at 96 GHz, and we fabricated W-band amplifier MMIC using the air-bridge wiring technology. The P out density of the MMIC reached 3.6 W/mm at 86 GHz. We proved the potential of the developed InAlGaN/GaN-HEMT experimentally using our unique device technology. With the aim of future applications, we developed a novel wiring-inter-layer technology. It consists of a cavity structure and a moisture-resistant dielectric film technology. We demonstrated excellent high-frequency performances and low current collapse originating in humidity-degradation using AlGaN/GaN-HEMT. This is also a valuable technology for InAlGaN/GaN-HEMT.

A 33 to 41-GHz Low Noise Amplifier (LNA) with a 3-dB Noise Figure (NF) using 0.12-µm InAlGaN/GaN HEMT was developed. The LNA consists of a two-stage common-gate amplifier with current reuse topology in order to obtain a high gain with low power consumption. The developed LNA achieved 15-dB gain, and an input return loss of less than −10 dB. The measured NF was 3 dB, and the power consumption was 280 mW. The measured OIP3 and OP1dB were 24 dBm and 13 dBm at 38 GHz under a supply voltage of 20 V. The chip size of the LNA is 1 × 0.7 mm2.

The JT-60SA project is implemented by Japan and Europe, and its construction is progressing toward the first plasma in 2019. The JT-60SA device, a highly shaped large superconducting tokamak with a variety of plasma control actuators, has been designed in order to contribute to ITER and to complement ITER in all the major areas of fusion plasma development necessary to decide DEMO reactor construction: Using ITER- and DEMO-relevant plasma regimes, heating conditions, and its sufficiently long discharge duration, JT-60SA enables studies on MHD stability at high β, heat/particle/momentum transport, high-energy ion physics, pedestal physics including ELM control, and divertor physics.

The JT-60SA tokamak, being built under the Broader Approach agreement jointly by Europe and Japan, is due to start operation in 2019 and is expected to give substantial contributions to both ITER and DEMO scenario optimization. A broad set of preparation activities for an efficient start of the experiments on JT-60SA is being carried out, involving the 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 main activities undertaken are presented and summarized.

A minimum arc current value for each contact material is one of the very important factors in consideration and comparison of arc discharge characteristics for proper selection of contact materials. In this paper, Ag and Ag alloy contacts were operated to break various dc load currents in the range of up to 1.5 A with 10-30 V, and occurrence rates of break arcs (having durations equal to or longer than a few microseconds) were investigated at each condition. As a result, it is confirmed that at the conventional minimum arc current level, break arcs on the order of several tens of microseconds occur at occurrence rate close to 100% when sufficient energy is supplied to contacts upon opening. Thus, careful attention should be paid to when using the term ldquominimum arc currentrdquo in order to avoid misinterpretation.

The concept for a compact DEMO reactor named 'SlimCS' is presented. Distinctive features of the concept are low aspect ratio (A = 2.6) and use of a reduced-size centre solenoid (CS) which has the function of plasma shaping rather than poloidal flux supply. The reduced-size CS enables us to introduce a thin toroidal field coil system which contributes to reducing the weight and perhaps lessening the construction cost. Low-A has merits of vertical stability for high elongation (κ) and high normalized beta (βN), which leads to a high power density with reasonable physics requirements. This is because high κ facilitates high nGW (because of an increase in Ip), which allows efficient use of the capacity of high βN. From an engineering aspect, low-A may ensure ease in designing blanket modules robust to electromagnetic forces acting on disruptions. Thus, a superconducting low-A tokamak reactor such as SlimCS can be a promising DEMO concept with physics and engineering advantages.

The characteristics of momentum transport and plasma rotation in the toroidal direction are studied, using near-perpendicular neutral beam injection (PERP-NBI), co tangential and counter (CTR) tangential NBI in JT-60U. Diffusive and non-diffusive terms of momentum transport are evaluated from the transient analysis by using the momentum source modulation. Fast ion losses due to the toroidal field ripple, which locally induces the edge CTR rotation, are used as a novel momentum source. Parameter dependence of these transport coefficients i.e. the toroidal momentum diffusivity (?) and the convection velocity (Vconv), and the relation between momentum and heat diffusivities (?i) are investigated in L-mode plasmas systematically. The toroidal momentum diffusivity increases with increasing heating power and decreases with increasing plasma current. The relation of ? and ?i to some non-dimensional parameters is investigated. A clear dependence of ?/?i on normalized plasma pressure (?N) is observed. It is also found that toroidal rotation velocity profiles in the case with and without external torque input can be almost reproduced by ? and Vconv estimated from the transient momentum transport analysis at low ? (?N < 0.4).

Results of active control of neoclassical tearing modes (NTMs) by electron cyclotron current drive (ECCD) in JT-60U are described. Growth of an NTM with poloidal mode number m = 3 and toroidal mode number n = 2 has been suppressed by ECCD inside the sawtooth inversion radius in the co-direction, showing the possibility of the coexistence of sawtooth oscillations and a small-amplitude m/n = 3/2 NTM without large confinement degradation. Stabilization of an m/n = 2/1 NTM by ECCD at the mode rational surface has been demonstrated with a small ratio of the current density driven by the electron cyclotron (EC) wave to the local bootstrap current density (∼0.5). In addition, dependence of the stabilization effect on ECCD location has been investigated in detail. It has been found that an m/n = 2/1 NTM can be completely stabilized with the misalignment of the ECCD location less than about half of the full island width, and that the m/n = 2/1 NTM is destabilized with the misalignment comparable to the full island width. Time-dependent, self-consistent simulation of magnetic island evolution using the TOPICS code has shown that the stabilization and destabilization of an m/n = 2/1 NTM are well reproduced with the same set of coefficients of the modified Rutherford equation. The TOPICS simulation has also clarified that EC wave power required for complete stabilization can be significantly reduced by narrowing the ECCD deposition width.

The primary facility for magnetic fusion energy (MFE) research in Japan is JT-60, in which DD plasma is heated up to an electron temperature of T e >10 keV with neutral beam injection (NBI) and radio frequency (RF) powers. It is noted that a normalized β N of 2.3 was maintained for 22 s. The JT-60 team has discovered internal transport barrier (ITB) in high-beta-poloidal discharges as an additional improved confinement mechanism like a previously known edge transport barrier (ETB). The operation forming the ITB has remarkably improved the plasma confinement. The second major MFE facility is a large helical device (LHD) at National Institute for Fusion Science (NIFS). The world's largest superconducting coil system of LHD generates helical magnetic field of 3 T where the stored magnetic energy reaches 1 GJ. A quasi-stationary plasma with an electron density of 4×10 18 m −3 at a temperature of about 1 keV was sustained for 54 min with the help of ion cyclotron range of frequency (ICRF) and electron cyclotron range of frequency (ECRF). The maximum electron density so far achieved is 5×10 20 m −3 with T e ∼1 keV. We report MFE programs related to International Thermonuclear Experimental Reactor (ITER).

The effects of toroidal rotation and toroidal field (TF) ripple on the edge pedestal structure and edge-localized-modes (ELMs) were examined in JT-60U. The amplitude of the TF ripple was changed by the installation of ferritic steel tiles (FSTs). The profile of toroidal rotation VT became less counter with FSTs, particularly in the case where co-NBI was used in a large volume configuration. In this case, the plasma pressure was raised across the whole profile of the plasma. At the plasma edge, higher pedestal temperature was obtained with the growth of pedestal width. However, the effect of FSTs became less significant in a small volume configuration. As the VT became less counter at the pedestal, ELM frequency fELM was reduced and ELM energy loss ΔWELM was increased at fixed power crossing the separatrix Psep. The observed larger ELM energy drop at VT in a relatively co-direction involves the ELM perturbations of the electron temperature profile across an ELM that extends radially more inward. In addition, the inter-ELM transport loss is reduced and the pedestal pressure pped is weakly raised. The effect of FSTs appeared clearly in the large volume configuration where pped is raised even at a given VT at the pedestal while pped is not changed by FSTs at the small volume configuration. This suggests that the reduction of the TF ripple, the magnitude of which can affect the VT profile, plays a role in the increasing of pped. For increased pped due to the use of co-NBI and the reduction of TF ripple, the spatial width of the pedestal ion temperature became greater.

After installation of ferritic steel tiles, fast ion losses due to toroidal field ripple have been reduced by 1/2–1/3. The increase in absorbed power at same injection power can reduce the required number of neutral beam injector (NBI) units to sustain a given normalized beta, βN, resulting in a better flexibility of torque input by increasing the available combination of tangential NBI units. By making use of these advantages to sustain an internal transport barrier (ITB), the performance of long-pulse ELMy H-mode plasmas was improved in terms of sustained duration time for both high βN and high thermal confinement enhancement factor (HH98(y,2)). High βN > 2.3 together with HH98(y,2) ~ 1 was sustained for 23.1 s (~12τR, where τR is the current diffusion time) at q95 ~ 3.3, which also provide high βNHH98(y,2) ≥ 2.2 and a bootstrap current fraction of ≥40%. βNHH98(y,2) of 2.0 was sustained for 28.6 s, which is limited by the maximum injection period of 30s for NBI system. These long-pulse plasmas are possible candidates for ITER hybrid operation scenario. Improved confinement is characterized by the larger thermal components at a given density maintained by lower heating power than in previous experiments. The strength of the ITB depends on the pedestal temperature, which varies with edge density while keeping constant the edge pressure (limited by type I ELMs). The fact that co-toroidal rotation as a result of reduced fast ion losses provides better quality of Te-ITB also contributes the improvement of thermal plasma confinement. These long-pulse plasmas indicate that further investigation to establish high performance plasmas longer than the time scale of wall saturation (τW) with active particle control is essential to establish the operational scenarios for next step devices, where the wall pumping does not work.

Pedestal and global plasma parameters are compared in conventional ELMy H-modes and improved confinement discharges from ASDEX Upgrade (AUG), DIII-D, JET and JT-60U with varying net input power. Both electron and ion pedestal pressures are studied. The pedestal top pressure pPED increases moderately with power in all tokamaks, in broad agreement with the power dependence of the IPB98(y, 2) scaling. Higher pedestal pressures are observed in AUG improved H-modes and in JT-60U high βpol discharges at q95 ~ 6.5 and high triangularity. For all machines and all scenarios a robust correlation between the total and the pedestal thermal stored energy is observed, with the ratio of the two varying between ~0.3 and 0.5. However the relative importance of pedestal and core confinement varies from regime to regime. In AUG the confinement improvement with respect to the IPB98(y, 2) scaling is due to improved pedestal confinement in improved H-modes with early heating and to both improved pedestal and core confinement in improved H-modes with late heating. In DIII-D hybrid discharges the increase in confinement factor compared with conventional H-modes is due to improved confinement in the plasma core. JT-60U reversed shear H-modes have strong internal transport barriers and thus improved core performance. In all four tokamaks improved edge stability is correlated with increasing total βpol and H98(y,2) increases with pedestal βpol. The analysed multimachine data set supports a scaling expression for the pedestal stored energy derived under the assumption that the dominant loss term for the pedestal is by thermal conduction in the edge transport barrier region.

The understanding and predictive capability of transport physics and plasma confinement is reviewed from the perspective of achieving reactor-scale burning plasmas in the ITER tokamak, for both core and edge plasma regions. Very considerable progress has been made in understanding, controlling and predicting tokamak transport across a wide variety of plasma conditions and regimes since the publication of the ITER Physics Basis (IPB) document (1999 Nucl. Fusion 39 2137-2664). Major areas of progress considered here follow. (1) Substantial improvement in the physics content, capability and reliability of transport simulation and modelling codes, leading to much increased theory/experiment interaction as these codes are increasingly used to interpret and predict experiment. (2) Remarkable progress has been made in developing and understanding regimes of improved core confinement. Internal transport barriers and other forms of reduced core transport are now routinely obtained in all the leading tokamak devices worldwide. (3) The importance of controlling the H-mode edge pedestal is now generally recognized. Substantial progress has been made in extending high confinement H-mode operation to the Greenwald density, the demonstration of Type I ELM mitigation and control techniques and systematic explanation of Type I ELM stability. Theory-based predictive capability has also shown progress by integrating the plasma and neutral transport with MHD stability. (4) Transport projections to ITER are now made using three complementary approaches: empirical or global scaling, theory-based transport modelling and dimensionless parameter scaling (previously, empirical scaling was the dominant approach). For the ITER base case or the reference scenario of conventional ELMy H-mode operation, all three techniques predict that ITER will have sufficient confinement to meet its design target of Q = 10 operation, within similar uncertainties.

The 'Progress in the ITER Physics Basis' (PIPB) document is an update of the 'ITER Physics Basis' (IPB), which was published in 1999 [1]. The IPB provided methodologies for projecting the performance of burning plasmas, developed largely through coordinated experimental, modelling and theoretical activities carried out on today's large tokamaks (ITER Physics R&D). In the IPB, projections for ITER (1998 Design) were also presented. The IPB also pointed out some outstanding issues. These issues have been addressed by the Participant Teams of ITER (the European Union, Japan, Russia and the USA), for which International Tokamak Physics Activities (ITPA) provided a forum of scientists, focusing on open issues pointed out in the IPB. The new methodologies of projection and control are applied to ITER, which was redesigned under revised technical objectives. These analyses suggest that the achievement of Q > 10 in the inductive operation is feasible. Further, improved confinement and beta observed with low shear (= high βp = 'hybrid') operation scenarios, if achieved in ITER, could provide attractive scenarios with high Q (> 10), long pulse (>1000 s) operation with beta

A systematic study of the effect of the level of toroidal plasma rotation at the top of the ion temperature pedestal ( ) on the edge localised mode (ELM) characteristics in JT-60U has been performed. The level of toroidal plasma rotation was varied by using different combinations of tangential and perpendicular neutral beam injection (NBI). In the grassy ELM regime at high triangularity (δ) and high safety factor (q), the ELM frequency clearly increased up to ~1400 Hz, when counter (ctr) plasma rotation was increased. The response of the ELM frequency was independent of poloidal beta (βp) in the range 0.84 0.53. Even in non-rotating plasma with balanced-NBIs, a high ELM frequency of ~400 Hz was observed without a large energy loss. When the frequency of the plasma rotation in the co-direction of the plasma current became higher than ~1 kHz, type I ELMs with a frequency of ~20 Hz was observed. The achieved pedestal pressure and plasma confinement were similar both in plasmas with type I ELMs and in plasmas with grassy ELMs. The energy loss due to grassy ELMs was evaluated from the reduction in the electron temperature, and the ratio of the energy loss to the pedestal stored energy was less than 1%.

In this work, we demonstrated an excellent output power (Pout) density of 3.0 W/mm at 96 GHz using a novel collapse-free InAlGaN/GaN-HEMT with an 80-nm gate for a millimeter-wave amplifier. The developed devices showed basic reliability for commercial products. To eliminate the current collapse, a unique double-layer silicon nitride (SiN) passivation film that has oxidation resistance was adopted. We proved the potential of InAlGaN/GaN-HEMT using our unique device technology experimentally and analytically.

To clarify the effect of the plasma-wall separation on the stability of the resistive wall modes (RWM), we have performed current-driven RWM experiments in the JT-60U ohmic plasmas. In the experiment, the plasma was placed near the outer wall, and the plasma current was ramped up so as to destabilize the current-driven RWM. An instability with a clear m/n = 3/1 mode structure was observed just before a thermal collapse. The growth time of this instability is about 10?ms, which is of the order of the skin time of the JT-60U wall. Therefore, the instability has been identified as a current-driven RWM. The plasma-wall separation was systematically changed, and it is found that the growth rates have strong dependence on the wall position. Thus, the growth rates became smaller with decreasing the plasma-wall separation. Additionally, the observed growth rates have been compared with the AEOLUS-FT code, which is based on the resistive MHD equations with a resistive wall, and the model of the RWM dispersion relation. The dependence of the observed growth rate on the wall position is in qualitative agreement with these numerical results.

The driving mechanism of toroidal rotation and the momentum transport are studied on JT-60U in relation to the toroidal rotation in the direction antiparallel to the plasma current, i.e. counter (CTR) direction, with near-perpendicular neutral beam (PERP-NB) injection. Fast ion losses due to the toroidal field ripple induce CTR rotation in the peripheral region, and the magnitude of CTR rotation with PERP-NBs reduces by installing the ferritic steel tiles as a consequence of the reduction in the ripple losses. It is also found that toroidal rotation velocity profiles in the core region can be explained by momentum transport considering diffusivity and convective velocity estimated from transient momentum transport analysis in L-mode plasmas.

The magnetic minor hysteresis loops and parameters of low carbon steel plate samples annealed at different temperatures have been measured and compared with the results of ring samples. The influences of stray flux in the fringe areas of the plate on the magnetic properties were discussed with the help of a finite element approach. It is found that the coercivity H C and the hysteresis loss W H of all of the plate samples sharply deviate from the values of the corresponding ring samples with the same applied field amplitude H a larger than 3000 A/m. It is proposed that the fringe areas of the plate sample are magnetized by the stray flux to an irreversible state at large Ha, and the remanent fields act against the applied field in the inverse direction, resulting in the extreme increase of H C and W H . Although these parameters measured at high-applied field can reflect the microstructure changes, low field results are preferable for the practical use.

This report summarizes Type I edge localized mode (ELM) dynamics measurements from a number of tokamaks, including ASDEX-Upgrade, DIII-D, JET, JT-60U and MAST, with the goal of providing guidance and insight for the development of ELM simulation and modelling. Several transport mechanisms are conjectured to be responsible for ELM transport, including convective transport due to filamentary structures ejected from the pedestal, parallel transport due to edge ergodization or magnetic reconnection and turbulent transport driven by the high edge gradients when the radial electric field shear is suppressed. The experimental observations are assessed for their validation, or conflict, with these ELM transport conjectures.

Several small/no ELM regimes such as EDA, grassy ELM, HRS, QH-mode, type II and V ELMs with good confinement properties have been obtained in Alcator C-Mod, ASDEX-Upgrade, DIII-D, JET, JFT-2M, JT-60U and NSTX. All these regimes show considerable reduction of instantaneous ELM heat load onto divertor target plates in contrast to conventional type I ELM, and ELM energy losses are evaluated as less than 5% of the pedestal stored energy. These small/no ELM regimes are summarized and widely categorized by their pedestal conditions in terms of the operational space in non-dimensional pedestal parameters and requirement of plasma shape/configuration. The characteristics of edge fluctuations and activities of ideal MHD stability leading to small/no ELMs are also summarized.

By conducting power scan experiments in three cases of toroidal field ripple configurations (δ r ≃ 0.4, 1.0 and 2.0%) with variation of the toroidal momentum source (co, balanced, counter) relative to the direction of the plasma current, the roles of the loss of fast ions and toroidal rotation on the H-mode pedestal structure were investigated. It was found that the pedestal pressure was increased with a decrease in fast ion loss power. On the other hand, the variation in toroidal rotation did not affect strongly the pedestal pressure, although the loss of fast ions due to existing large toroidal field ripple forced the toroidal rotation basically to the counter direction in JT-60U. Due to the enhanced counter toroidal rotation induced by the loss of fast ions, similar V tor profiles were obtained for discharges with different toroidal momentum source and loss of fast ions. When the loss of fast ions became smaller, pedestal density was raised remarkably while there was no significant difference in the temperature profiles.

Effects of plasma rotation and losses of fast ions on Type-I ELM characteristics have been systematically studied in the JT-60U tokamak, scanning combinations of NBI (tangential co-, balanced- and counter-NBI plus perpendicular NBI) in the three types of plasma configurations (corresponding toroidal field ripple at the plasma edge, δ r ∼0.4, 1.0 and 2.0%). New findings on the Type-I ELM characteristics are as follows: smaller ELM energy loss normalized by pedestal stored energy, ΔW ELM /W ped , and faster ELM frequency, f ELM , are confirmed in the counter-NBI in comparison with the co-NBI discharges. Nevertheless, the power loss due to ELM, P ELM (= ΔW ELM × f ELM ), normalized by heating power crossing the separatrix, P SEP , is constant regardless of the direction of the momentum injection at each plasma configuration. In contrast, the P ELM /P SEP decreases with increasing δ r . The relationship between ELMs, rotation and the losses of fast ions are discussed.

Properties of magnetohydrodynamic (MHD) plasma based on syngas (CO, H/sub 2/) combustion products were investigated experimentally with shock tube facility. The experiments were carried out under various MHD generator load and shock tube operation conditions. Important characteristics of syngas plasma such as temperature, electric field, conductivity, and total output power were directly measured and evaluated. Special attention was paid to the influence of syngas composition (CO:H/sub 2/:O/sub 2/ ratio). The results show that syngas combustion can provide high plasma ionization and attainable plasma electrical conductivity has an order of 60-80 S/m at gas temperature 3100-3300 K.

Plasma current start-up to 100 kA was achieved successfully in the JT-60U tokamak without the use of the centre solenoid (completely solenoidless tokamak operation). Only poloidal field coils located on the outboard side of the torus were used, in combination with strong ionization by electron cyclotron (EC) power. The presence of a field null was not necessary for plasma current start-up, but the flux conversion efficiency was low in such a case. In a nearly solenoidless start-up, low neutral pressures were favoured, and the optimum location of the EC resonance was slightly to the high field side of the vacuum vessel centre. The required EC power for efficient utilization of flux swing in JT-60U was about 1 MW. A plasma current of 260 kA was maintained for 1 s by NB only, and plasma current ramp-up from 215 to 310 kA was achieved by EC and neutral beam (NB) only (without lower hybrid current drive (LHCD)). However, the ramp-up efficiency was much lower compared with LHCD. Recharging of the centre solenoid was observed with only counter and perpendicular NB injection, indicating bootstrap overdrive. Integration of these elements can lead to the achievement of a completely solenoidless tokamak operation.

Transient transport experiments are performed in plasmas with and without internal transport barriers (ITB) on LHD and JT-60U. The dependence of χe on the electron temperature, Te, and on the electron temperature gradient, ∇Te, is analysed with an empirical non-linear heat transport model. In plasmas without an ITB, two different types of non-linearity of the electron heat transport are observed from cold/heat pulse propagation: the χe depends on Te and ∇Te in JT-60U, while the ∇Te dependence is weak in LHD. Inside the ITB region, there is none or weak ∇Te dependence both in LHD and JT-60U. Growth of the cold pulse driven by the negative Te dependence of χe is observed inside the ITB region (LHD) and near the boundary of the ITB region (JT-60U).

Compatibility of advanced tokamak plasmas with high density and high radiation loss has been investigated in both reversed shear (RS) plasmas and high βp H-mode plasmas with a weak positive shear on JT-60U. In the RS plasmas, the operating regime is extended to high density above the Greenwald density (nGW) with high confinement (HHy2 > 1) and high radiation loss fraction (frad > 0.9) by tailoring the internal transport barriers (ITBs). With a small plasma-wall gap, the radiation loss in the main plasma (inside the magnetic separatrix) reaches 80% of the heating power due to metal impurity accumulation. However, high confinement of HHy2 = 1.2 is sustained even with such a large radiation loss in the main plasma. By neon seeding, the divertor radiation loss is enhanced from 20% to 40% of the total radiation loss. In the high βp H-mode plasmas, high confinement (HHy2 = 0.96) is maintained at high density ( ) with high radiation loss fraction (frad ~ 1) by utilizing high-field-side pellets and argon (Ar) injection. The high is attributed to the formation of strong density ITB. Strong core-edge parameter linkage for confinement improvement is observed, where the pedestal pressure and the core plasma confinement increase together. The measured radiation profile including contributions from all impurities in the main plasma is peaked, and the central radiation is ascribed to the contribution from Ar accumulated inside the ITB. Impurity transport analyses indicate that the Ar density profile, twice as peaked as the electron density profile, which is the same level as that observed in the high βp H-mode plasma, can yield an acceptable radiation profile even with a peaked density profile in a fusion reactor.

The energy loss for grassy edge localized modes (ELMs) has been studied to investigate the applicability of the grassy ELM regime to ITER. The grassy ELM regime is characterized by high frequency periodic collapses of 800–1500 Hz, which is ~15 times faster than that for type I ELMs. The divertor peak heat flux due to grassy ELMs is less than 10% of that for type I ELMs. This smaller heat flux is caused by a narrower radial extent of the collapse of the temperature pedestal. The different radial extent between type I ELMs and grassy ELMs agrees qualitatively with the different radial distribution of the eigenfunctions as determined from ideal MHD stability analysis. The dominant ELM energy loss for grassy ELMs appears to be caused by temperature reduction, and its ratio to the pedestal stored energy was 0.4–1%. This ratio is lower by a factor of about 10 than that for type I ELMs, which typically have between 2–10% fractional loss of the pedestal energy. A systematic study of the effects of counter (CTR) plasma rotation on the ELM characteristics has been performed using a combination of tangential and perpendicular neutral beam injections (NBIs) in JT-60U. In the high plasma triangularity (δ) regime, ELM characteristics (e.g. amplitude, frequency and type) can be changed from type I ELMs to high frequency grassy ELMs as the CTR plasma rotation is increased. On the other hand, in the low δ regime, complete ELM suppression (QH-mode) can be sustained for long periods up to 3.4 s (~18τE or energy confinement times), when the plasma position in terms of the clearance between the first wall and the plasma separatrix is optimized during the application of CTR-NBIs. In JT-60U, a transient QH phase was also observed during the CO-NBI phase with almost no net toroidal rotation at the plasma edge.

This paper reports the results of the progress in stationary discharges with a large bootstrap current fraction in JT-60U towards steady-state tokamak operation. In the weak shear plasma regime, high-βp ELMy H-mode discharges have been optimized under nearly full non-inductive current drive conditions by the large bootstrap current fraction (fBS ~ 45%) and the beam driven current fraction (fBD ~ 50%), which was sustained for 5.8 s in the stationary condition. This duration corresponds to ~26τE and ~2.8τR, which was limited by the pulse length of negative-ion-based neutral beams. The high confinement enhancement factor H89 ~ 2.2 (HH98y2 ~ 1.0) was obtained and the profiles of current and pressure reached the stationary condition. In the reversed shear plasma regime, a large bootstrap current fraction (fBS ~ 75%) has been sustained for 7.4 s under nearly full non-inductive current drive conditions. This duration corresponds to ~16τE and ~2.7τR. The high confinement enhancement factor H89 ~ 3.0 (HH98y2 ~ 1.7) was also sustained, and the profiles of current and pressure reached the stationary condition. The large bootstrap current and the off-axis beam driven current sustained this reversed q profile. This duration was limited only by the duration of the neutral beam injection.

Magnetic properties derived from major hysteresis loops are sensitive to the microstructures of iron-based steels and have therefore been used for nondestructive evaluation of degradation in ferromagnetic materials contained in structural components. A permeability obtained from an ac small field excitation may be a more useful parameter than one obtained from a major hysteresis loop from the viewpoints of electricity consumption and realtime measurement. In this paper, the relation between the ac permeability calculated from impedance measurements of coil elements and the ductile-brittle transition temperature (DBTT) and Vickers hardness of S15C steels is clarified. The possibility of nondestructive evaluation (NDE) using ac permeability measurements is also discussed.

The dependence of the thermal diffusivity on the heat flux has been investigated in JT-60U plasmas by varying the neutral beam power. In positive magnetic shear (PS) plasmas, the ion thermal diffusivity (χi) in the core region generally increases with the heating power, as with the L mode at low heating powers. However, as a result of the intensive central heating, a weak internal transport barrier (ITB) is formed, and the χi value in the core region starts to decrease. Corresponding to a further increase in the heating power, a strong ITB is formed and the χi value is reduced substantially. In the case of reversed magnetic shear (RS) plasmas, on the other hand, no power degradation of the χi value is observed in any of the heating regimes. The electron thermal diffusivity (χe) is correlated with the χi value in PS and RS plasmas. Furthermore, the dependence of the ion thermal diffusivity on the radial electric field (Er) shear has also been investigated. By considering the non-locality in the relation between the Er shear and the χi value, a threshold in the effective Er shear to change the state from a weak to a strong ITB is identified.

Simultaneous fast ELM measurements using a reflectometer, interferometers, Dα intensity and a magnetic probe reveal the detailed characteristics of type I ELMs. From the phase signal of the reflectometer indicating the radial movement of the cut-off layer, four different phases in the ELM event, i.e. a precursor phase, collapse phase, recovery phase and a relaxation phase, were observed. In the collapse phase, the radial extent of the collapse of the density pedestal reached up to twice the pedestal width. A fast drop of the horizontal interferometer signal was observed, while the vertical interferometer on the high-field side edge plasma only exhibited a small and delayed reduction. These features of the plasma response due to ELMs might indicate a non-uniform collapse of the density pedestal in the poloidal direction, localized near the low-field side (LFS) midplane. Expelled particles from the LFS midplane were measured by using scrape-off layer Mach probes, and the heat flux to the divertor target was measured using a fast infrared TV camera. The time constant of the peak heat load was comparable to that of the enhancement of the plasma flow towards the divertor target. The normalized ELM energy loss by pedestal stored energy in JT-60U was found to be less than 10% of the pedestal stored energy. It shows weak dependence on the pedestal parameters such as electron collisionality, parallel ion loss time and the ratio of the pedestal density to the Greenwald density limit.

Recent progress in the physics and engineering design study for the modification programme of JT-60 is presented. In order to achieve a steady state high-β plasma operation, which is the dominant issue of this programme, physics design for the MHD control and the stability analysis is investigated. Engineering design and the R & D for the superconducting coils, irradiation shield are performed well towards the mission of programme.

Advanced tokamak regimes obtained in ASDEX Upgrade, DIII-D, FT-U, JET, JT-60U, TCV and Tore Supra experiments are assessed both in terms of their fusion performance and capability for ultimately reaching steady-state using data from the international internal transport barrier database. These advanced modes of tokamak operation are characterized by an improved core confinement and a modified current profile compared to the relaxed Ohmically driven one. The present results obtained in these experiments are studied in view of their prospect for achieving either long pulses ('hybrid' scenario with inductive and non-inductive current drive) or ultimately steady-state purely non-inductive current drive operation in next step devices such as ITER. A new operational diagram for advanced tokamak operation is proposed where the figure of merit characterizing the fusion performances and confinement, H × βN / q 295, is drawn versus the fraction of the plasma current driven by the bootstrap effect. In this diagram, present day advanced tokamak regimes have now reached an operational domain that is required in the non-inductive ITER current drive operation with typically 50% of the plasma current driven by the bootstrap effect (Green et al 2003 Plasma Phys. Control. Fusion 45 587). In addition, the existence domain of the advanced mode regimes is also mapped in terms of dimensionless plasmas physics quantities such as normalized Larmor radius, normalized collisionality, Mach number and ratio of ion to electron temperature. The gap between present day and future advanced tokamak experiments is quantitatively assessed in terms of these dimensionless parameters.