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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)

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.

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).

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

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.

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.

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.

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.

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.

The influence of dominant electron heating on internal transport barriers (ITBs) was investigated in JT-60U. The electron cyclotron waves were mainly utilized in order to extend the experimental space into a dominant electron heating regime. The experiments were carried out on both reversed magnetic shear (RS) and positive magnetic shear (PS) plasmas. It was found that in RS plasmas clear ITBs in Te, Ti and ne profiles remained even under dominant electron heating. Energy confinement in excess of twice the H-mode scaling has been obtained in the regime. On the other hand, it was found that in PS plasmas the Ti ITB was degraded by dominant electron heating.

This paper reports results on the progress in steady-state high-βp ELMy H-mode discharges in JT-60U. A fusion triple product, nD(0)τETi(0), of 3.1 × 1020 m−3 s keV under full non-inductive current drive has been achieved at Ip = 1.8 MA, which extends the record value of the fusion triple product under full non-inductive current drive by 50%. A high-beta plasma with βN ~ 2.7 has been sustained for 7.4 s (~60τE), with the duration determined only by the facility limits, such as the capability of the poloidal field coils and the upper limit on the duration of injection of neutral beams. Destabilization of neoclassical tearing modes (NTMs) has been avoided with good reproducibility by tailoring the current and pressure profiles. On the other hand, a real-time NTM stabilization system has been developed where detection of the centre of the magnetic island and optimization of the injection angle of the electron cyclotron wave are done in real time. By applying this system, a 3/2 NTM has been completely stabilized in a high-beta region (βp ~ 1.2, βN ~ 1.5), and the beta value and confinement enhancement factor have been improved by the stabilization.

We report on the electrical transport measurements of single wall carbon nanotubes (CNs) formed in a single crystal of zeolite (IUPAC Cord AFI) at room temperature. Photo-irradiation effects associated with the resonant photo-excitation of CNs at 900 nm are found in the I–V characteristics. The conductive character changes from ohmic to somewhat semiconducting. The character does not depend on the polarity of the bias voltage. Subtracting the ohmic component, an energy threshold to give nonlinear conduction is found at 0.25 eV. A possibility of the metal–semiconductor junction formation caused by the photo-irradiation is discussed.

A substantial reduction of the L-H transition threshold power was observed under the W-shaped divertor in JT-60U, in comparison with the open divertor. Radiation power in the main region was also decreased in this case. These differences of radiation power from the main plasma between open and W-shaped divertor were not large enough to account for the apparent reduction of threshold power. In this paper, the emphasis is on the edge plasma parameters which relate to the L-H transition. It was found that edge plasma pressure just before the L-H transition in the W-shaped divertor case became smaller than that in the open divertor case. After the modification of divertor geometry, edge ion collisionality just before the L-H transition also became lower and it was established with lower auxiliary power input. It was suggested that reduction of edge ion collisionality arose from neutral particles near the X-point by the analysis of poloidal profile of neutral density. Therefore, neutral particles near the X-point prevented the threshold power for the L-H transition from more reduction at low plasma density, and caused the shift of plasma density which gave minimum threshold power.

Internal and edge transport barriers are widely observed in a number of devices, including tokamak and helical systems, with various heatings. The E×B flow shear suppression of turbulence seems to explain formation and subsequent propagation of transport barriers in many cases. However, the role of the safety factor profile has not been clearly understood. Moreover, electron transport still remains a major issue. Tokamak research is challenging the active control of transport barriers. Steady or quasisteady performance of H-modes with internal barriers has been gradually improving and stable sets of profiles close to the steady-state solution have been demonstrated at high β.

In order to establish the physics basis for the sustainment of the high integrated performance required in the next-step experimental reactor, JT-60U has been optimizing the discharge control scenarios of high confinement plasmas and extending the operation regimes by utilizing a variety of heating and current drive systems, including the new Electron Cyclotron Heating/Electron Cyclotron Current Drive (ECH/ECCD) system, and the W-shaped pumped divertor. For the integration, current and pressure profile controls are essential and high triangularity is beneficial. The range of plasma current with high values of H-factor, N and bootstrap fraction was extended to 1.5 MA with nearly full non-inductive current drive by Negative-ion-based Neutral Beam (N-NB) injection into the high-p H-mode with Edge Localized Modes (ELMy H-mode). High current drive efficiency of 1.3 × 1019 A W-1 m-2 was demonstrated for N-NB. High triangularity (0.4-0.5) operation extended the long pulse (~3 s) high N (2.5-2.7) region to low-q95 (~3). For the reversed shear (RS) mode, feedback control of neutron production rate and stored energy enabled reproducible achievement of the DT equivalent fusion gain QeqDT>1 and sustainment of QeqDT~0.4-0.5 for ~1 s. Using the RS mode and the high high-p ELMy H-mode, the high confinement region has been extended to higher Te/Ti and also to higher density. Electron heating by Lower Hybrid Range of Frequency (LHRF) and N-NB extended the high confinement region to Te/Ti>1. Argon gas puffing improved confinement in the high density regime. With the W-shaped pumped divertor, the threshold heating power for the L-H transition was reduced by 30% as compared with the open divertor. Using the pumped divertor, a multiple parameter feedback control including both core and divertor plasma parameters was demonstrated.

The ITER Threshold Database, which at present comprises data from nine divertor tokamaks, is described. The main results are presented and discussed. The properties and dependences of the power threshold in individual devices are reviewed. In particular, the analysis shows a rather general linear dependence on magnetic field, but a non-monotonic density dependence that varies from device to device. Investigation of the combined database suggests that the threshold dependence Pthres approximately=0.3neBT2.5 shows reasonable agreement with the data. This expression yields Pthres approximately=150 MW at a density of 0.5*1020 m-3 for ITER. Other expressions with weaker size dependence, and therefore lower threshold power for ITER, are also discussed. Their agreement with the present data is poorer than that of the above expression. In addition, the database is investigated by statistical discriminant analysis. The edge data included at present are described and discussed. Finally, there is a discussion of the implications of the results for ITER

The method for reducing a poloidal magnetic flux consumption of external coils is investigated to reduce the size of the central solenoid (CS) in the DEMO reactor. The reduction of the poloidal magnetic flux consumption during a plasma current ramp-up phase by electron cyclotron (EC) heating is investigated using an integrated modeling code suite, TOPICS. A strongly reversed shear q profile tends to be produced if intense off-axis EC heating is applied to obtain a large reduction of the flux consumption. In order to overcome this tendency, we find a method to obtain the optimum temperature profile which minimizes the poloidal flux consumption for a wide range of the q profile. We try to reproduce the optimum temperature profile for a weakly reversed shear q profile using six EC rays of 20 MW. As a result, the resistive flux consumption during the current ramp-up can be reduced by 63% from the estimation using the Ejima constant of 0.45 and the total flux consumption can be reduced by 20% from the conventional estimation. In addition, we find that the resistive flux consumption is closely related to the volume averaged electron temperature and not to the profile shape. Using this relation, the required heating power is estimated to be 31 MW based on a well established global confinement scaling, ITER L-89P. As a result, it is clarified that the poloidal magnetic flux consumption can be reduced by 20% using 20–31 MW of EC heating for a weakly reversed shear q profile. This reduction of the flux consumption accounts for 10% reduction of the CS radius.

Photo-induced electric current modulation is studied in AFI-SWCNs systems by visible and near infra-red laser light at room temperature. Remarkable polarization anisotropy in the modulation is observed. λ-dependence of the polarization anisotropy is analyzed by a simple model based on the optical anisotropy observed in absorption spectra.

The central magnetohydrodynamic (MHD) activities are strongly affected by hydrogen pellet injection. The change in the sawtooth characteristics (viz., crash time, crash mechanisms, and sawtooth period) seem to be dependent on the density (and pressure) peakedness. With deepening pellet penetration, the sawtooth frequency becomes longer. At the sawtooth emerging after the deep pellet penetration into high‐Ip limiter discharges, only a small amount of the central kinetic energy is released and the crash does not follow the fully reconnecting style. The sawtooth crash after the pellet injection tends to have more ideal‐like characteristics for higher density and pressure peaking factors. At each sawtooth, the m=1 rotation frequency changes suddenly to the ion‐diamagnetic direction or the codirection (parallel to the plasma current).

The pressure at the top of the edge transport barrier (or ‘pedestal height’) strongly impacts fusion performance, while large edge localized modes (ELMs), driven by the free energy in the pedestal region, can constrain material lifetimes. Accurately predicting the pedestal height and ELM behavior in ITER is an essential element of prediction and optimization of fusion performance. Investigation of intermediate wavelength MHD modes (or ‘peeling–ballooning’ modes) has led to an improved understanding of important constraints on the pedestal height and the mechanism for ELMs. The combination of high-resolution pedestal diagnostics, including substantial recent improvements, and a suite of highly efficient stability codes, has made edge stability analysis routine on several major tokamaks, contributing both to understanding, and to experimental planning and performance optimization. Here we present extensive comparisons of observations to predicted edge stability boundaries on several tokamaks, both for the standard (Type I) ELM regime, and for small ELM and ELM-free regimes. We further discuss a new predictive model for the pedestal height and width (EPED1), developed by self-consistently combining a simple width model with peeling–ballooning stability calculations. This model is tested against experimental measurements, and used in initial predictions of the pedestal height for ITER.

The magnetic susceptibility χ and resistivity ϱ were examined in La 2− x M x CuO 4 (M = Sr, Ba). The T -dependent part χ s of χ below x ⋍ x c ( x c = 0.15 and 0.18 for M = Ba and Sr) consists of a term arising from antiferromagnetically correlated Cu-spins χ s F ( T ), which is characterized by a broad peak at T max . On the other hand, χ s above x = x c consists of χ s F ( T ) and a Curie term ( C/T ) which is small but grows with increasing x . The superconducting transition temperature T c decreases rapidly with increasing x above x c . The relation between T c and C can be explained in terms of the magnetic depairing due to local moments. T ma x and the g -value of the Cu-spins are almost constant above x ⋍ 0.2 , although these are strongly x -dependent for x p ⋍ 0.2 go onto electronic states which differ largely in character from those below p ⋍ 0.2 . It is also found that a significant change in the ϱ - T curve occurs at x ⋍ 0.2 .

Recent lower hybrid current drive (LHCD), and heating (LHH) experiments on JT-60 are reported. The current drive product of neRpIRF approximately 12.5*1019 m-2 MA was achieved at the LH power of approximately 4.5 MW, and the CD efficiency, the energy confinement, the global power balance and the heat load on divertor plates were investigated in high power LHCD plasmas. Nearly steady state H-mode discharges were found during LHCD with two different frequency injections. Sawtooth suppression in NB heated plasmas by LHCD have shown an improvement in confinement near the plasma center. Parametric instabilities in LH heating experiments were significantly reduced by increasing the plasma current, and the stored energy increased linearly with heating power of up to approximately 9 MW at ne approximately 7*1019 m-3 and Ip=2.75 MA. Parametric instabilities near the plasma edge in the ion heating regime were also reduced in peaked density plasmas produced by pellet injection and LH waves increased the central plasma pressure at ne(0) > 1.4*1020 m-3.

A spontaneous transition phenomena between two states of a plasma with an internal transport barrier (ITB) is observed in the steady-state phase of the magnetic shear in the negative magnetic shear plasma in the JT-60U tokamak. These two ITB states are characterized by different profiles of the second radial derivative of the ion temperature inside the ITB region (one has a weak concave shape and the other has a strong convex shape) and by different degrees of sharpness of the interfaces between the L mode and the ITB region, which is determined by the turbulence penetration into the ITB region.

Our understanding of the physics of internal transport barriers (ITBs) is being advanced by analysis and comparisons of experimental data from many different tokamaks worldwide. An international database consisting of scalar and two-dimensional profile data for ITB plasmas is being developed to determine the requirements for the formation and sustainment of ITBs and to perform tests of theory-based transport models in an effort to improve the predictive capability of the models. Analysis using the database indicates that: (a) the power required to form ITBs decreases with increased negative magnetic shear of the target plasma, and: (b) the E x B flow shear rate is close to the linear growth rate of the ion temperature gradient (ITG) modes at the time of barrier formation when compared for several fusion devices. Tests of several transport models (JETTO, Weiland model) using the two-dimensional profile data indicate that there is only limited agreement between the model predictions and the experimental results for the range of plasma conditions examined for the different devices (DIII-D, JET, JT-60U). Gyrokinetic stability analysis (using the GKS code) of the ITB discharges from these devices indicates that the ITG/TEM growth rates decrease with increased negative magnetic shear and that the E x B shear rate is comparable to the linear growth rates at the location of the ITB.

A conceptual design study of a steady-state tokamak, JT-60 Super Upgrade, is being carried out. The capability of the present JT-60 facility will be fully utilized for this upgrade. The mission of JT-60SU is to establish integrated basis of physics and technology for steady-state tokamak reactors. In JT-60SU, steady-state physics will be evaluated in the intermediate parameter region between the present tokamaks and steady-state tokamak reactors. Technology development for long pulse operation and research for engineering safety will also be pursued.

The nuclear shielding design for the vacuum vessel of the JT-60SU has been studied for deuterium operation. A low activation material, Ti-6Al-4V, is proposed through a design study. A 25-40 cm thick water tank-type vacuum vessel using Ti-6Al-4V and boron (1% of /sup 10/B) doped water are shown to be the best choice. This design concept gave results consistent with the design requirements: the one-turn resistivity of the vacuum vessel is /spl sim/50 /spl mu//spl Omega/, and the dose inside the vessel is /spl sim/15 mrem/h a year after the shutdown. The total nuclear heating of the whole TF coil can be reduced to /spl sim/11 kW for a D-D fusion neutron rate of 1/spl times/10/sup 18//s.

It was found that activated carbon possessed remarkably higher heat of adsorption for ATF additives (automatic transmission fluid additives) than any other materials contained within the clutch facing. Friction performance tests in low sliding velocity range were conducted for two kinds of clutch facings, containing activated carbon powder or graphite one as key ingredients for adsorption. As the additives mixed to the base oil to form the testing oils, polybutenyl succinimide (SI) was selected as a dispersant, and tricresyl phosphate and dialkyl phosphite were selected as a friction modifier. The results showed that the combination in activated carbon with SI was most effective to improve the initial anti-shudder performance of continuous slip torque converter clutch (CSTCC). It is considered that this effect was due to the strong adsorption of SI to the surface of activated carbon in the clutch facings.

Recent results from the DIII-D tokamak have provided significant contributions to the understanding of many of the elements of tokamak physics and the application of this understanding to the design of next generation devices including ITER and CIT. The limitations of magnetohydrodynamic stability on the values of plasma beta (the ratio of kinetic pressure to the containing pressure of the magnetic field) that can be attained has been experimentally demonstrated and found to be described by existing theory. Values of beta (10.7%) well in excess of those required for proposed devices (ITER and CIT) have been demonstrated. Regimes of confinement (H-mode) have been established that scale favorably to proposed next generation devices, and experiments demonstrating the dependence of the energy confinement on plasma size have been completed. Understanding of confinement is rapidly developing especially in the areas of bulk transport and the role of turbulence in the plasma edge. Key experimental results in areas of plasma transport and edge plasma phenomena are in agreement with theories based on short wavelength turbulence. Control of the divertor heat loads and impurity influx has been demonstrated, and new progress has been made in the understanding of plasma edge phenomena. Experiments with ion Bernstein wave heating have not found regimes in which these waves can produce effective central ion heating. Electron cyclotron current drive experiments have demonstrated 70 kA of driven current in 400 kA discharges.

The characteristics of plasmas with the safety factor q between that of tokamaks and that of reversed field pinches were studied on the OHTE device, with toroidal fields of up to 3.9 kG and plasma currents of up to 140 kA. Plasmas in this regime (0 < q < 1) are called ultra-low q plasmas. The discharges are globally stable and last much longer than the resistive decay time of the current. A unique feature of most discharges is a series of plateaus in the current (or equivalently q) as the current decreases in a 'staircase' fashion. The plateau q is related to the major rational ratios (1/4, 1/3, 1/2, 1, etc.), suggesting plasma self-organization. Profile measurements show that the q-profile is somewhat different from those of the Taylor relaxed states. The soft X-ray temperature measurement indicates that the plasma has burned through the radiation barrier.

The characteristics of a toroidal discharge with qa ≈ 1/2 is studied on TORIUT-6. Ultra low q equilibria that are stable against the global kink mode are realized through MHD relaxation and are shown to be characterized by dq/dr < 0 and a hollow current profile. The stable state lasts for 600 μs, which is of the order of the classical diffusion time. These experimental observations are consistent with non-linear MHD simulation results.

With use of a multimachine pedestal database, essential issues for each regime of ELM types are investigated. They include (i) understanding and prediction of pedestal pressure during type I ELMs, a reference operation mode of a future tokamak reactor; (ii) identification of the operation regime of type II ELMs, which have small ELM amplitude with good confinement characteristics; (iii) identification of the upper stability boundary of type III ELMs for access to the higher confinement regimes with type I or II ELMs; (iv) understanding the relation between core confinement and pedestal temperature in conjunction with the confinement degradation in high density discharges. Both scaling and model based approaches for expressing pedestal pressure are shown to roughly scale the experimental data well and could be used to make initial predictions for a future reactor. q and delta are identified as important parameters for obtaining the type II ELM regime. A theoretical model of type III ELMs is shown to reproduce the upper stability boundary reasonably well. It is shown that there exists some critical pedestal temperature below which the core confinement starts to degrade. It is also shown that it is possible to obtain improved pedestal conditions for good confinement in high density discharges by increasing the plasma triangularity.

The maximum normalized beta achieved in long-pulse tokamak discharges at low collisionality falls significantly below both that observed in short pulse discharges and that predicted by the ideal MHD theory. Recent long-pulse experiments, in particular those simulating the International Thermonuclear Experimental Reactor (ITER) [M. Rosenbluth et al., Plasma Physics and Controlled Nuclear Fusion (International Atomic Energy Agency, Vienna, 1995), Vol. 2, p. 517] scenarios with low collisionality nu(e)*, are often limited by low-m/n nonideal magnetohydrodynamic (MHD) modes. The effect of saturated MHD modes is a reduction of the confinement time by 10%-20%, depending on the island size and location, and can lead to a disruption. Recent theories on neoclassical destabilization of tearing modes, including the effects of a perturbed helical bootstrap current, are successful in explaining the qualitative behavior of the resistive modes and recent results are consistent with the size of the saturated islands. Also, a strong correlation is observed between the onset of these low-m/n modes with sawteeth, edge localized modes (ELM), or fishbone events. consistent with the seed island required by the theory. We will focus on a quantitative comparison between both the conventional resistive and neoclassical theories, and the experimental results of several machines, which have all observed these low-min nonideal modes. This enables us to single out the key issues in projecting the long-pulse beta limits of ITER-size tokamaks and also to discuss possible plasma control methods that can increase the soft beta limit, decrease the seed perturbations, and/or diminish the effects on confinement. (C) 1997 American Institute of Physics.

Key parts of the ITER scenarios are determined by the capability of the proposed poloidal field (PF) coil set. They include the plasma breakdown at low loop voltage, the current rise phase, the performance during the flat top (FT) phase and a ramp down of the plasma. The ITER discharge evolution has been verified in dedicated experiments. New data are obtained from C-Mod, ASDEX Upgrade, DIII-D, JT-60U and JET. Results show that breakdown for E axis < 0.23–0.33 V m−1 is possible unassisted (ohmic) for large devices like JET and attainable in devices with a capability of using ECRH assist. For the current ramp up, good control of the plasma inductance is obtained using a full bore plasma shape with early X-point formation. This allows optimization of the flux usage from the PF set. Additional heating keeps l i(3) < 0.85 during the ramp up to q 95 = 3. A rise phase with an H-mode transition is capable of achieving l i(3) < 0.7 at the start of the FT. Operation of the H-mode reference scenario at q 95 ∼ 3 and the hybrid scenario at q 95 = 4–4.5 during the FT phase is documented, providing data for the l i (3) evolution after the H-mode transition and the l i (3) evolution after a back-transition to L-mode. During the ITER ramp down it is important to remain diverted and to reduce the elongation. The inductance could be kept ⩽1.2 during the first half of the current decay, using a slow I p ramp down, but still consuming flux from the transformer. Alternatively, the discharges can be kept in H-mode during most of the ramp down, requiring significant amounts of additional heating.