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A review of the development of advanced tokamak scenarios at the Joint European Torus (JET) is presented. It has been established that the current profile plays an important role in these regimes, and the presentation of the experimental achievement has been organized with this in mind. The main achievements are discussed: from high beta plasmas starting with a fully diffused plasma current; from hybrid scenarios with a flat current profile and central q around unity; from pellet-enhanced modes where the role of reversed magnetic shear (transiently) was first established; from optimized shear configurations with weakly reversed shear allowing the establishment of internal transport barriers in D-T plasmas for the first time, including the production of 8.2 MW of fusion power; and from strongly reversed shear and steady-state scenarios. The required development of the control techniques for these advanced scenarios is also described. The results obtained have significantly contributed to the development of advanced scenarios for ITER operation. The prospects for further development of hybrid and steady-state scenarios for ITER are indicated in view of the ongoing upgrades to additional heating systems in JET.

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

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.

Significant progress has been made in the area of advanced modes of operation that are candidates for achieving steady state conditions in a fusion reactor. The corresponding parameters, domain of operation, scenarios and integration issues of advanced scenarios are discussed in this chapter. A review of the presently developed scenarios, including discussions on operational space, is given. Significant progress has been made in the domain of heating and current drive in recent years, especially in the domain of off-axis current drive, which is essential for the achievement of the required current profile. The actuators for heating and current drive that are necessary to produce and control the advanced tokamak discharges are discussed, including modelling and predictions for ITER. The specific control issues for steady state operation are discussed, including the already existing experimental results as well as the various strategies and needs (qψ profile control and temperature gradients). Achievable parameters for the ITER steady state and hybrid scenarios with foreseen heating and current drive systems are discussed using modelling including actuators, allowing an assessment of achievable current profiles. Finally, a summary is given in the last section including outstanding issues and recommendations for further research and development.

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

In this paper we present the result of a study carried out at the Frascati Tokamak Upgrade (FTU), on a set of full non-inductive current driven, electron heated, L-mode discharges aimed at investigating the parametric dependence of the electron density profile on the electron temperature and safety factor gradients as predicted by quasi-linear drift-turbulence transport theory. Experiments in FTU allow the extension of similar studies carried out on other tokamaks to plasmas with higher density and higher magnetic field. Magnetic shear and electron temperature gradients are found to drive opposite turbulent particle flows in the gradient region (0.3 < r/a ? 0.5), while inward thermo-diffusion alone is found in the plasma core (r/a ? 0.3). Density profiles at midradius appear to be controlled by a convective term proportional to the density and independent of the gradient of temperature and magnetic shear. A linear increase in density peaking versus effective collisionality is found, differing from the scaling observed in other FTU plasma regimes.

Experiments designed for the attainment of the hybrid advanced mode of operation in JET revealed the existence of a new form of reconnecting modes with toroidal number n = 1 and poloidal number m = 1, which have the same radial structure as sawtooth precursors but are characterized by small growth rate and low saturation amplitude. These modes develop in repetitive cycles with typical growth times of 100 ms. Temperature perturbations are not in the form of sawteeth; in fact, there is slow and mild erosion of the central temperature without any crash. Comparison with theory of m = 1 modes indicates that the slow growth observed in the hybrid regime is caused by low magnetic shear (a measure of the radial variation of field line inclination) and possibly by diamagnetic effects in the plasma core.

The Frascati Tokamak Upgrade (FTU) is a high field and high density device which allows one to study confinement and transport at plasma parameters generally not accessible to other tokamaks and is relevant to next generation experiments (ITER). In this paper we study the effect of different density, temperature and q profiles on confinement and transport in various types of plasmas including ohmic, externally heated and pellet-fuelled discharges. Peaking of the electron density profile in pellet-fuelled discharges has been found to enhance the energy confinement time on FTU to values as high as 120 ms at densities of the order of those expected in ITER. Experiments designed to further improve confinement by injecting pellets at higher densities show the existence of a second threshold for saturation of confinement when the local electron heat conductivity corresponds to the same order of the local ion neoclassical conductivity in the region of maximum temperature gradient. Reflectometry measurements show turbulence suppression in pellet-injected discharges (as predicted by microstability analysis) and give insights to the change in turbulence in electron internal transport barrier plasmas. Finally, particle transport has been studied in experiments with full LH current drive.

A tokamak plasma with internal transport barriers (ITBs) is the best candidate for a steady ITER operation, since the high energy confinement allows working at plasma currents (Ip) lower than the reference scenario. To build and sustain an ITB at the ITER high density (>=1020 m-3) and largely dominant electron (e-) heating is not trivial in most existing tokamaks. FTU can instead meet both requests, thanks to its radiofrequency heating systems, lower hybrid (LH, up to 1.9 MW) and electron cyclotron (EC up to 1.2 MW). By the combined use of them, ITBs are obtained up to peak densities ne0 > 1.3 × 1020 m-3, with central e- temperatures Te0 ? 5.5 keV, and are sustained for as long as the heating pulse is applied (>35 confinement times, ?E). At ne0 ? 0.8 × 1020 m-3 Te0 can be larger than 11 keV. Almost full current drive (CD) and an overall good steadiness is attained within about one ?E, 20 times faster than the ohmic current relaxation time. The ITB extends over a central region with an almost flat or slightly reversed q profile and qmin ? 1.3 that is fully sustained by off-axis lower hybrid current drive. Consequent to this is the beneficial good alignment of the bootstrap current, generated by the ITB large pressure gradients, with the LH driven current. Reflectometry shows a clear change in the turbulence close to the ITB radius, consistent with the reduced e- transport. Ions (i+) are significantly heated via collisions, but thermal equilibrium with electrons cannot be attained since the e--i+ equipartition time is always 4-5 times longer than ?E. No degradation of the overall ion transport, rather a reduction of the i+ heat diffusivity, is observed inside the ITB. The global confinement has been improved up to 1.6 times over the scaling predictions. The ITB radius can be controlled by adjusting the LH power deposition profile that is affected mostly by the q value of the discharge, while the ITB strength can be varied through central EC heating. FTU experiments have shown that ITER-like e-ITBs are achievable.

We present integrated modelling of steady-state and hybrid scenarios for ITER parameters using several predictive transport codes. These employ models for non-inductive current drive sources in conjunction with various theory-based and semi-empirical transport models. In conjunction with the simulation effort, the current drive models are being evaluated in a series of cross-code and code-experiment comparisons under ITER-relevant conditions. New benchmark evaluations of current drive from injection of neutral beams (NBCD), electron cyclotron waves (ECCD) and lower hybrid waves (LHCD) are reported. Simulations using several transport modelling codes self-consistently calculate the heating and current drive sources using ITER design parameters. Operating constraints are also taken into account, although the calculations reported here still require further refinement. The modelling addresses both the final stationary state and dynamic access to it. The simulations indicate that generation and control of internal and edge barriers to access and maintain high confinement will be a major undertaking for future simulations, as well as a challenge for the ITER steady-state and hybrid experimental programme.

This chapter reports the main physics results obtained with three radio-frequency-injection systems. The frequency of 8 GHz for the lower hybrid (LH) current drive (CD) (LHCD) system was chosen to explore CD at high density: full CD has been demonstrated for central densities up to 1.4 x 10{sup 20} m{sup -3} at 0.5 MA with an applied power up to 2.0 MW. The Frascati Tokamak Upgrade (FTU) database shows CD efficiencies from 0.1 to 0.3 x 10{sup 20} AW{sup -1} m{sup -2}. In combined experiments with electron cyclotron (EC) waves (140 GHz, up to 1.2 MW), a suprathermal absorption by the fast electron tail generated by LHCD has been observed in both downshifted and upshifted interaction regimes, with the resulting electron cyclotron current drive (ECCD) ranging from 20 to 100 kA, depending on experimental conditions. With pure EC resonance heating, the narrowness of the radial power deposition profile has been exploited, resulting in strong local electron heating. Results in high-density regimes are also presented. The third system (433 MHz, 0.5 MW) is the first to test ion Bernstein wave (IBW) coupling with a waveguide antenna. The experiment operates at high frequency, avoiding the occurrence of nonlinear phenomena at the edge.more » Improved confinement regimes resulting in a central peaking of the pressure profiles have been achieved with P{sub IBW} up to 0.4 MW. Modeling and experimental results are summarized.« less

Transport studies are presented in this chapter. Global scaling studies have been performed using several transport codes. Ohmic plasmas are found to follow the ITER97 L-mode scaling. Transport coefficients are discussed for improved confinement scenarios achieved in the Frascati Tokamak Upgrade (FTU): the repetitive pellet enhanced plasma mode, showing neoclassical confinement with H-factors up to 1.6, and the electron internal transport barriers (ITBs) with large transport barriers and H-factors up to 1.3. Heat transport models have been tested using electron cyclotron resonance heating (ECRH), steady or modulated, as a probe. The electron temperature stiffness observed in the main bulk of steady FTU plasmas can be interpreted both with a critical gradient transport model and with a model based on the existence of canonical profiles. ECRH has also been used to benefit from the improved confinement generally associated with low or negative magnetic shear, and large electron temperatures have been achieved in these conditions. Profile resiliency is observed so that heat transport is not consistent with a constant thermal diffusivity. Experimental optimization is discussed together with the analysis of transport coefficients. Thorough discussions of impurity transport are given, both for intrinsic and injected (from laser blow-off) impurities. Code simulation and experimentalmore » data are compared for a series of FTU experiments focusing on the improved confinement modes (pellets and ITBs). A moderate inward pinch velocity is generally required to reproduce the data.« less

Strong electron internal transport barriers (ITBs) are obtained in the Frascati Tokamak Upgrade (FTU) with the combined injection of lower hybrid (LH) (up to 1.9 MW) and electron cyclotron (EC) (up to 0.8 MW) radio-frequency waves. ITBs occur during either the current plateau or the ramp-up phase, both in full and partial current drive (CD) regimes, up to n e0 > 1.4 × 10 20 m -3 , relevant to ITER operation. Central electron temperatures T e0 > 8 keV, at n e0 0.8 X 10 20 m -3 , are sustained for up to 36 confinement times. The ITB extends over a region where a slightly reversed magnetic shear is established by off-axis LHCD and can be even larger than r/a = 0.5. EC power is used either to benefit from this improved confinement by heating inside the ITB or to enhance the peripheral LH power deposition and CD with off-axis resonance. Collisional ion heating is also observed, but thermal equilibrium with the electrons is not attained since the electron-ion equipartition time is always 4 to 5 times longer than the energy confinement time. An extensive transport modeling of these discharges, performed by means of the ASTRA code, is also presented. During the ITB phase, the ion diffusivity is close to the neoclassical value while the electron shear-dependent Bohm-gyro-Bohm model accounts quite well for T e (r, t), The Ray Tracing Fokker-Planck model, used to describe the LHCD physics, appears satisfactory to analyze and interpret the experimental results. It turns out that the barrier radius is mainly influenced by the LHCD deposition. In particular, a wider barrier is obtained the lower q a is and the larger the plasma density is.

The main physics results achieved in the recent years in the Frascati Tokamak Upgrade (FTU) are reviewed. The main focus of research has been the development of performance plasmas at high densities (up to 4 x 10{sup 20} m{sup -3}), high magnetic field (up to 8 T) and plasma current (up to 1.6 MA), that are therefore in a domain of relevance for burning physics experiments such as ITER. The main tools consist in the development of plasma conditioning techniques and the use of various electron heating and current drive systems. Improved confinement regimes have been developed, including (a) the production of steady electron internal transport barriers at high density and electron temperature (up to central electron temperature of 11 keV at a central density of 0.9 x 10{sup 20} m{sup 3}), (b) the production of repetitive pellet enhanced plasma modes with deep pellet deposition leading to a substantial increase of the neutron yield (and a record FTU value of the fusion product n{sub i}T{sub i}{tau}{sub E} up to 0.8 x 10{sup 20} m{sup -3} keV.s), and (c) the production of radiation improved modes at high magnetic field. Main results on the supporting physics program will also be given inmore » the domain of plasma wave physics (lower hybrid current drive, electron cyclotron resonance frequency, ion Bernstein waves), heat and impurities transport, and magnetohydrodynamic studies.« less

A conceptual study is presented for a substantial upgrade of the Frascati Tokamak Upgrade (FTU) up to B = 8 T, I = 6 MA, and R [approximately equal to] 1.3 m to study burning plasma (BP) issues in deuterium plasmas operating up to an equivalent DT gain close to Q = 2 in the ELMy H-mode and to Q = 5 with an internal transport barrier (ITB). The effect of alpha particles is simulated by {approx}1 MeV fast {sup 3}He minority heating produced by ion cyclotron resonance heating (20 MW). Thanks to the high-density values ([approximately equal to]4 x 10{sup 20} m{sup -3}), the FT3 plasmas are characterized by short electron-ion equipartition time (60 ms in the ELMy H-mode scenario) and slowing-down time (44 ms), with respect to the energy confinement time of {approx}340 ms, a feature characteristic of BP experiments but not always satisfied with present tokamak devices. Advanced scenarios at 5 T with fully noninductive current drive can be investigated with a steady-state current density profile achieved in

Tokamak discharges with improved energy confinement properties arising from internal transport barriers (ITBs) have certain attractive features, such as a large bootstrap current fraction, which suggest a potential route to the steady-state mode of operation desirable for fusion power plants. This paper first reviews the present state of theoretical and experimental knowledge regarding the formation and characteristics of ITBs in tokamaks. Specifically, the current status of theoretical modelling of ITBs is presented; then, an international ITB database based on experimental information extracted from some nine tokamaks is described and used to draw some general conclusions concerning the necessary conditions for ITBs to appear, comparing these with the theoretical models. The experimental situation regarding the steady-state, or at least quasi-steady-state, operation of tokamaks is reviewed and finally the issues and prospects for achieving such operational modes in ITER are discussed. More detailed information on the characteristics of ITBs in some 13 tokamaks (as well as helical devices) appears in the appendix.

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.

A low/negative magnetic shear profile is maintained in the Joint European Torus in a plasma target with plasma current of 2.4 MA using 2.2 MW of lower hybrid (LH) power combined with neutral beam injection and ion cyclotron radiofrequency heating. In this scenario, an internal transport barrier (ITB) with time duration up to about 4 s is produced. The fraction of LH driven current is about 25% the total plasma current. During LH power application, the layer with reversed shear q-profile can be maintained in a suitable radial position to inhibit the onset of turbulence, which might otherwise drive the ITB to collapse. LH power could be used as a tool to drive moderate amounts of noninductive off-axis current and sustain high performance ITBs at high plasma current.

The paper highlights the JET work in physics and technology during the period of the JET Joint Undertaking (1978-1999), with special emphasis on what has been learned for extrapolation to a NEXT STEP device. - Global confinement scaling has been extended to high currents and heating powers. Dimensionless scaling experiments of ELMy H?mode plasmas suggest that bulk plasma transport is gyro-Bohm and predict ignition for a device with ITER-FDR parameters. Experiments in which the plasma elongation and triangularity were varied independently show a strong increase of confinement time with elongation (?E ~ ?a 0.8 ? 0.3), thus supporting a basic design principle of ITER-FEAT. With the Pellet Enhanced Performance (PEP) mode, JET has discovered the beneficial effect of reversed magnetic shear on confinement, opening the possibility of advanced tokamak scenarios. - With a three stage programme of progressively more closed divertors, JET has demonstrated the benefits of divertor closure, in particular, of high divertor neutral pressure which facilitates helium removal. It has also shown that in detached (or semidetached) radiative divertor plasmas the average power load on the target plates of a NEXT STEP device should be tolerable but, in addition, that the transient power loads during ELMs could cause problems. - In 1991 JET has demonstrated the first ever controlled production of a megawatt of fusion power. More extensive D-T experiments in 1997 (DTE1) have established new records in fusion performance: 16?MW transient fusion power with Qin = 0.62 (i.e. close to breakeven, Qin = 1) and 4?MW steady state fusion power with Qin = 0.18 for 4?s. DTE1 has also allowed a successful test of various reactor ICRF heating schemes and a clear demonstration of alpha particle heating, consistent with classical expectations. - JET has developed and tested some of the most important technologies for a NEXT STEP and a reactor, in particular the safe handling of tritium and the remote handling of large equipment as demonstrated by the Remote Divertor Exchange following DTE1.

High density plasmas (n0 ≈ 8 × 1020m-3) featuring steady improved core confinement have been obtained in FTU up to the maximum nominal toroidal field (8 T) by deep multiple pellet injection. These plasmas also feature high purity efficient electron-ion coupling and peaked density profiles sustained for several confinement times. Neutron yields in excess of 1 × 1013 n/s are measured, consistent with the reduction of the ion transport to neoclassical levels.

The role of turbulence stabilization by sheared E × B flow is analysed for different types of internal transport barrier (ITB) in JET discharges. After the ITB formation, the shearing rate ωs is found to consistently exceed a rather crude estimate for the ion temperature gradient driven turbulence. The roles played by toroidal plasma rotation and by density and temperature gradients in both the radial electric field Er and the shearing rate ωs are evaluated. It turns out that the major contribution in triggering the positive feedback leading to ITB onset is given by the toroidal rotation.

Optimized shear (OS) experiments in JET can generate an internal transport barrier (ITB) during a high power heating phase early in the plasma discharge. A strong link is generally observed between the formation of the barrier and the location of an integer q magnetic surface within a low magnetic shear (s = r/q(dq/dr)) region of the plasma. However, if the q-profile for such experiments is modified by applying lower hybrid heating and current drive (LHCD) before the main heating pulse, to provide a region of reduced or negative magnetic shear in the plasma core, ITBs can be formed the location of which does not exhibit any apparent association with a particular internal magnetic surface. Initial results suggest that q-profile modification using an LHCD prelude can also be used to reduce the heating power level required for ITB generation.

Results achieved on JET during the 1997-1999 experimental campaigns in the physics of energetic ions and runaway electrons are reviewed. Heating of deuterium-tritium (DT) plasmas by fusion born alpha particles is found to be similar to that achieved by comparable ICRF heating of deuterium plasmas. The stability of alpha particle driven Alfv?n eigenmodes (AEs) in the highest fusion power ELM-free H mode discharges is shown to be consistent with the existing theoretical analysis for AEs. Direct measurements of the trapped alpha particle and knock-on deuteron distribution functions by a neutral particle analyser (NPA) are described. New ICRF heating scenarios tested in JET DT plasmas are presented in view of the possible use of the ICRF heating of ITER-like DT plasmas on route to ignition. The energetic ion pinch in the presence of toroidally asymmetric ICRF waves is studied experimentally on JET. Recent experimentally observed effects of ICRF accelerated ions on sawteeth in JET are reviewed. Detailed time and space resolved X ray images of the electron runaway beam in flight spontaneously generated by disruptions on JET are described.

High performance discharges are routinely obtained on JET with low or reversed magnetic shear (s = (r/q)dq/dr), and the potential for steady state operation of such discharges is under investigation. With the use of the proper heating and fuelling, these `optimized shear' (OS) discharges exhibit an internal transport barrier (ITB), resulting in a strong peaking of the pressure profile, and thus in high fusion performance. These regimes have been extensively studied during the last (DD and DT) JET campaigns in order to promote this type of scenario as the basis for `advanced tokamak' operation. A review is given of the highest performance achieved on JET OS discharges during the last experimental campaigns, in both DD (up to 5.6 × 1016neutrons/s) and DT operation (fusion power up to 8.2 MW, ni0Ti0τE up to 1021 m-3 keV s). The role of the plasma edge is pointed out, as the power required to trigger an ITB is often higher than the H mode power threshold, leading to double barrier regimes. The presence of an H mode pedestal both modifies the ITB and induces edge bootstrap and ELM activity, which should be controlled to prolong such discharges. The operational procedure of optimization is then discussed, addressing the problems of ITB formation (power threshold, timing of the main heating phase, i.e. optimization of the target q profile, influence of the heating scheme, electron versus ion ITBs), ITB evolution (expansion of the ITB footpoint, H mode formation) and ITB termination (disruptive and/or `soft' MHD limits). Finally, the crucial problem of the route to steady state for such OS discharges is addressed, both in terms of ITB sustainment and control within the stability domain and in terms of edge pedestal control by means of impurity injection. The impurity behaviour is found, and examples of high performance discharges sustained for several energy confinement times are given (βN = 1.95, H89 = 2.3, Pfusioneq~10 MW, QDTeq~0.4 sustained for ~3 s). Extrapolation towards fully non-inductive current drive is discussed.

Wave-particle interaction phenomena in the Alfven Eigenmode (AE) frequency range are investigated at the Joint European Torus [P. H. Rebut and B. E. Keen, Fusion Technol. 11, 13 (1987)] using active and passive diagnostic methods. Fast particles are generated by neutral beam injection, ion cyclotron resonance heating, and fusion reactions. External antennas are used to excite stable AEs and measure fast particle drive and damping separately. Comparisons with numerical calculations lead to an identification of the different damping mechanisms. The use of the active AE diagnostic system to generate control signals based on the proximity to marginal stability limits for AE and low-frequency magnetohydrodynamic (MHD) modes is explored. Signatures of the different nonlinear regimes of fast particle driven AE instabilities predicted by theory are found in the measured spectra. The diagnostic use of AE measurements to get information both on the plasma bulk and the fast particle distribution is assessed.

High power combined NBI + ICRF heating experiments have been carried out in the JET Mark IIa divertor configuration in the hot ion ELM-free H mode regime, both in deuterium (DD) and in deuterium-tritium (DT) plasmas. Results are presented from a wide range of additional heating power levels, ICRF up to 9.5 MW tuned to the fundamental hydrogen minority, NBI up to 22 MW, and for plasma currents up to 4.2 MA and toroidal fields up to 3.6 T. Discharges with combined NBI + ICRF heating show a clear improvement in electron temperature, DD neutron yield and stored energy with respect to NBI only discharges. High energy neutral particle analyser data show that acceleration of the NBI deuterons takes place due to absorption of ICRF power at the second harmonic deuterium resonance. This is confirmed by numerical simulations with the PION code, indicating that up to 40% of the ICRF power is absorbed by bulk and NBI ions. ICRF heating has been an essential ingredient in the DT experiments in the ELM-free hot ion regime, contributing to the achievement of a record fusion power of 16.1 MW and a record stored energy of 17 MJ.

The main limitation to the performance of JET optimized shear (OS) discharges is due to MHD instabilities, mostly in the form of a disruptive limit. The structure of the MHD mode observed as a precursor to the disruption, as measured from SXR and ECE diagnostics, shows a global ideal MHD mode. The measured mode structure is in good agreement with the calculated mode structure of the pressure driven kink mode. The disruptions occur at relatively low normalized beta (1 < βN < 2), in good agreement with calculated ideal MHD stability limits for the n = 1 pressure driven kink mode. These low limits are mainly due to the extreme peaking factor of the pressure profiles. Other MHD instabilities observed in JET OS discharges include, usually benign, chirping modes. These modes, which occur in bursts during which the frequency changes, have the same mode structure as the disruption precursor but are driven unstable by fast particles.

High fusion performance has been achieved in optimized shear discharges in JET with the early appearance of internal transport barriers (ITBs) in the high power phase in both DD and DT plasmas. An ITB has been produced in a large variety of conditions, including LHCD only, ICRH only, NBI only and NBI + ICRH with LHCD preheat. An ITB may coexist with the edge barrier. The high pressure gradient produced by ICRH combined with high power NBI heating starting in a low density plasma with q(0) ≈ 1.5-2 is the main prerequisite for an efficient ITB in high performance discharges. Strong coupling between q profiles, MHD activity, energy confinement and fusion yield has been observed. The transition from L to H mode can be triggered by MHD events. The transition is accompanied by significant momentary changes in the density fluctuation spectrum in the peripheral plasma. The ITB may persist for some time in both ELM-free and ELMy H mode.

The unique roles played by ICRH in the preparation, formation and sustainment of internal transport barriers (ITBs) in high fusion performance JET optimized shear experiments using the Mark II poloidal divertor are discussed. Together with LHCD, low power ICRH is applied during the early ramp-up phase of the plasma current, `freezing in' a hollow or flat current density profile with q(0)>1. In combination with up to ~20 MW of NBI, the ICRH power is stepped up to ~6 MW during the main low confinement (L mode) heating phase. An ITB forms promptly after the power step, revealed by a region of reduced central energy transport and peaked profiles, with the ion thermal diffusivity falling to values close to the standard neoclassical level near the centre of both DD and DT plasmas. At the critical time of ITB formation, the plasma contains an energetic ICRF supported hydrogen minority ion population, contributing ~50% to the total plasma pressure and heating mainly electrons. As both the NBI population and the thermal ion pressure develop, a substantial part of the ICRF power is damped resonantly on core ions (ω = 2 ωcD = 3ωcT), contributing to the ion heating. In NBI step-down experiments, high performance has been sustained by maintaining central ICRH; analysis shows the efficiency of central ICRH ion heating to be comparable to that of NBI. The highest DD fusion neutron rates (RNT = 5.6 × 1016 s-1) yet achieved in JET plasmas have been produced by combining a low magnetic shear core with a high confinement (H mode) edge. In DT, a fusion triple product niTiτE = (1.2 ± 0.2) × 1021 m-3 keV s was achieved with 7.2 MW of fusion power obtained in the L mode and with up to 8.2 MW of fusion power in the H mode phase.

Transport analysis of high performance JET plasmas with optimized magnetic shear (OMS) has revealed many interesting features which cannot be explained by the existing JET empirical transport model (Erba, M., et al., Plasma Phys. Control. Fusion 39 (1997) 261). TRANSP analysis shows that transport coefficients in OMS plasmas are often reduced in the plasma core (Cottrell, G.A., et al., in Controlled Fusion and Plasma Physics (Proc. 24th Eur. Conf. Berchtesgaden, 1997), Vol. 21A, Part I, European Physical Society, Geneva (1997) 81) to the level of ion neoclassical transport. TRANSP analysis and predictive modelling with JETTO show that this region of improved confinement appears near the plasma centre and then expands outwards in a way which does not follow either the evolution of the region with the negative magnetic shear or the propagation of the region with a large shear in plasma rotation. The best agreement with experiment has been achieved by using a transport model which combines the effect of a long wavelength decoupling due to small magnetic shear with its suppression by strong rotational shear. Predictive modelling of some of the characteristic JET OMS plasmas gives quite good agreement between such a model and the experimental data.

A number of experiments with heating of DT plasmas using ICRF waves have been carried out at JET. The results of these experiments have been analysed by comparing experimentally measured quantities with the results of numerical simulations. In particular, four scenarios have been examined: (a) heating of minority (~5-20%) deuterons at the fundamental ion cyclotron frequency, ω = ωcD; (b) second harmonic heating of tritium, ω = 2ωcT; (c) fundamental minority heating of 3He with a few per cent of 3He; (d) second harmonic heating of deuterium, ω = 2ωcD. An important aim of the analysis was to assess whether the present understanding of the ICRF physics is adequate for predicting the performance of ICRF in DT plasmas. In general, good agreement between experimental results and simulations was found which increases the confidence in predictions of the impact of ICRF heating in future reactors. However, when a relatively high deuterium concentration was used in the ω = ωcD scenario, discrepancies were observed. In order to increase confidence in the simulations, the sensitivity of the simulation results to various plasma parameters has been studied.

Steady state high performance with improved core confinement and sustainable plasma edge conditions has been approached on JET in a double barrier (DB) mode. The DB mode combines an internal transport barrier of the optimized shear regime with an edge transport barrier of an ELMy H?mode regime. Improved confinement with an H?factor HITER-89 ? 2 has been maintained for four energy confinement times. Ion and electron temperature profiles remain peaked in the DB mode, while the density profile is broad and similar in shape to the conventional ELMy H?mode profile. The energy confinement improves across the whole plasma cross-section, and the ion heat conductivity falls to the neoclassical level in the core. Particle transport studies show that impurity accumulation can be avoided in the DB mode. In DT discharges the DB mode has attained a fusion gain of Q ? 0.4, producing 6.8?MW of fusion power, compared with Q ? 0.2 in the conventional sawtoothing ELMy H?mode. The ELMs are more benign, with an amplitude an order of magnitude smaller in the DB mode. The DB mode has a high potential to improve performance in reactor relevant conditions.

Reactor relevant ICRH scenarios have been assessed during DT experiments on the JET tokamak using H mode divertor discharges with ITER-like shapes and safety factors. Deuterium minority heating in tritium plasmas was demonstrated for the first time. For 9% deuterium, an ICRH power of 6 MW gave 1.66 MW of fusion power from reactions between suprathermal deuterons and thermal tritons. The Q value of the steady state discharge reached 0.22 for the length of the RF flat-top (2.7 s), corresponding to three plasma energy replacement times. The Doppler broadened neutron spectrum showed a deuteron energy of 125 keV, which was optimum for fusion and close to the critical energy. Thus, strong bulk ion heating was obtained at the same time as high fusion efficiency. Deuterium fractions around 20% produced the strongest ion heating together with a strong reduction of the suprathermal deuteron tail. The ELMs had low amplitude and high frequency and each ELM transported less plasma energy content than the 1% required by ITER. The energy confinement time, on the ITERH97-P scale, was 0.90, which is sufficient for ignition in ITER. 3He minority heating, in approximately 50:50 D:T plasmas with up to 10% 3He, also demonstrated strong bulk ion heating. Central ion temperatures up to 13 keV were achieved, together with central electron temperatures up to 12 keV. The normalized H mode confinement time was 0.95. Second harmonic tritium heating produced energetic tritons above the critical energy. This scheme heats the electrons in JET, unlike in ITER where the lower power density will allow mainly ion heating. The inverted scenario of tritium minority ICRH in a deuterium plasma was demonstrated as a successful heating method producing both suprathermal neutrons and bulk ion heating. Theoretical calculations of the DT reactivity mostly give excellent agreement with the measured reaction rates.

Ion cyclotron resonance heating (ICRH) of JET high-performance (optimized shear and hot-ion H-mode) deuterium discharges, characterized by strong neutral beam injection (NBI), high ion and electron temperatures and high deuterium-deuterium fusion reactivity, is analysed using numerical simulations. The main aim is to investigate under which conditions good bulk ion heating can be obtained with ICRH tuned to the fundamental hydrogen resonance, coinciding with the second-harmonic deuterium resonance, near the plasma centre. The analysis has been facilitated by adding NBI source terms in the ion cyclotron range of frequencies (ICRF) code PION, which calculates the time evolution of the ICRF power deposition and the resonating ions distribution functions in a self-consistent way. The results show that damping ICRF power on deuterons results in a more central bulk ion heating profile than minority hydrogen ions, which is an advantage. The highest bulk ion heating fractions of up to 50% of the total ICRF power were obtained by using multiple frequencies, the effects of which have been simulated in detail for the first time. Furthermore, operating with high plasma densities has been found to be beneficial for maximizing the bulk ion heating.

Formation of core regions in Tore Supra and JET tokamaks with reduced transport coefficients is reported. Characteristics of the enhanced confinement regions and the physics process involved in their formation and maintenance should be considered separately when the electron or ion components are predominantly heated. In Tore Supra and JET, central electron temperature transitions are observed by injecting lower hybrid waves at modest power levels during the current ramp-up phase of the discharges. Transport analyses stress the importance of the low magnetic shear in the core to explain the anomalous electron transport reduction. With high-power dominant ion heating schemes in JET (neutral beam injection and ion cyclotron resonance heating), internal transport barriers have been obtained in plasmas fuelled with a mixture of deuterium-tritium (D-T) ions leading to a successful production of fusion power (8.2 MW) in this regime. Similar additional power levels to those applied in pure deuterium (D-D) plasmas are required to establish internal transport barriers in D-T plasmas. In D-D and D-T plasmas, ion thermal diffusivities are reduced close to their neoclassical levels in the plasma core and electron thermal diffusivities decrease by one order of magnitude at midplasma radius. The combined role of magnetic shear and velocity shear can explain the formation and evolution of plasma core regions with low energy transport coefficients.

In lower hybrid current drive (LHCD) experiments in JET, up to 7.3 MW of lower hybrid (LH) power has been coupled to X point plasmas, resulting in sawtooth suppression and full current drive up to 3 MA. The current drive efficiency reached ηCD = 0.26 × 1020 m-2A/W in these experiments. The LH power deposition and driven current profiles can be quite well reproduced with the ray tracing and Fokker-Planck code in JET, even when multipass absorption of the LH wave is dominant. Profile control with off-axis LHCD has been used for increasing q above 1, thereby suppressing sawteeth before the edge localized mode (ELM)-free hot ion H mode. In optimized shear plasmas, a broad q profile with central negative shear was formed with moderate LH power (~2 MW), applied in the low density phase during the current rampup. An internal transport barrier with improved electron confinement was produced, which resulted in a peaking of the electron temperature profile and Te0 up to 10 keV at ne0 ≤ 1.5 × 1019 m-3.

It has been shown that the core energy confinement in tokamaks can be significantly improved by modifying the plasma current profile. Plasma regimes where the transport is reduced to its neoclassical value in the centre of the plasma within a region called the internal transport barrier (ITB) can be achieved with a zero or negative magnetic shear. In this paper the role of the radio frequency (RF) methods: electron cyclotron resonance, ion cyclotron resonance and lower hybrid current drive in producing these ITBs is reviewed. Combined techniques where RF waves are used to modify the magnetic shear together with a central heating method (neutral beam injection or RF) are described. Application of RF methods to high fusion yield plasmas in JET with modified shear are presented together with a discussion on the prospects of achieving steady-state high-performance plasmas. RF techniques are already playing an important role in these `advanced scenarios' and are likely to become indispensable tools for future steady-state tokamak reactors.

Recent progress in the area of ICRF physics is examined and summarized. A very brief summary of simplified theoretical predictions is given. Illustrative examples of experimental results in the area of ICRF heating and fast-wave current drive are given. ICRF heating is found to be highly effective in heating plasmas to high temperatures in modern tokamaks. Fast-wave current drive has been demonstrated in several experiments and the results are in good agreement with theory. Mode-conversion heating and current drive should be effective for profile control, but experimental results are still preliminary.

Plasmas with the highest deuterium fusion neutron rates yet achieved in JET have been produced by combining a hollow or flat central current profile with a high-confinement (H-mode) edge. In these discharges, lower hybrid current drive (LHCD) and ion cyclotron resonance heating (ICRH) preheating, applied early in the current ramp-up phase, `freezes in' a hollow or flat current density profile. When the combined neutral beam injection (NBI) and ICRH heating power is increased, a region of reduced transport and highly peaked profiles forms during the L-mode phase and persists into the later H-mode phase when the fusion reactivity reaches a maximum. Transport analysis shows the formation of a central region of good confinement (the internal transport barrier or ITB) which expands with radial velocity . The clearest signatures of this region are large gradients in the ion temperature and toroidal rotation profiles. Ion thermal diffusivities in the central region are of the order of the neoclassical value. The position and rate of expansion of the ITB radius correlates well with the calculated rational q = 2 surface. The confinement improvement can also be seen in electron density, and, to a lesser extent, in electron temperature. The ITB can persist in combination with the edge transport barrier characteristics of the H-mode.

The observation of internal transport barriers (ITBs) in which ion thermal diffusivity is reduced to a neoclassical level has been made for the first time in tokamak plasmas fueled with deuterium and tritium ions using a broad current density profile. The heating and current profiles required to obtain an ITB are similar in D-T and D-D plasmas. Central ion temperatures of 40 keV and plasma pressure gradients of ${10}^{6}\mathrm{Pa}/\mathrm{m}$ were observed in a D-T plasma, leading to a fusion triple product ${n}_{i}{T}_{i}{\ensuremath{\tau}}_{E}{\phantom{\rule{0ex}{0ex}}=\phantom{\rule{0ex}{0ex}}1.10}^{21}{\mathrm{m}}^{\ensuremath{-}3}\mathrm{keV}\mathrm{s}$ and 8.2 MW of fusion power. There is potential for further optimization as a step towards the development of efficient tokamak fusion reactors.

Heating during the early part of the current rise phase gives a low or negative magnetic shear (= (dq/dr)) in the centre of JET plasmas. Under these conditions the confinement improves with high additional heating power heating during the current ramp-up phase of the discharge. The reduction in the transport manifests itself as a peaking of the profiles with a large gradient region near = 0.55. The best discharges have no transport barrier at the edge of the plasma (L-mode). This allows central power deposition by the neutral beams in JET. A control of the plasma pressure, using feedback of the additional heating power in real-time, minimizes the impact of magnetohydrodynamic instabilities. As a result, these discharges achieve the highest D-D neutron rates in JET; , with , and .

Ion cyclotron resonance heating (ICRH) experiments have been carried out in JET D-T plasmas using scenarios applicable to reactors. Deuterium minority heating in tritium plasmas is used for the first time and produces 1.66 MW of D-T fusion power for an ICRH power of 6 MW. The Q value is 0.22, which is a record for steady state discharges. Fundamental He-3 minority ICRH, in both 50:50 D-T and tritium dominated plasmas, generates strong bulk ion heating and ion temperatures up to 13 keV. Second harmonic tritium ICRH is seen to heat mainly the electrons as expected for JET conditions. All three schemes produce H-mode plasmas.

Experiments using high-power heating during the current ramp-up phase of the discharge have obtained the highest D-D neutron rates in JET; neutrons , with , keV and keV. The best discharges ( MA and tesla) have peaked pressure profiles with a transport barrier located at = 0.55. The pressure peaking is limited by MHD modes and requires active input power control to achieve the best performance. Deuterium neutral beam injection into a tritium-rich target plasma has established internal transport barriers at power levels close to the lowest threshold for pure deuterium plasmas.

An analysis of the complete JET D-D threshold database, from 1990 to 1997 inclusive is presented. A JET scaling is derived for the threshold power through the separatrix, estimated by considering also radiation losses from the bulk plasma. Such scalings are applied to obtain a first qualitative estimate of the amount of power above threshold that is needed to obtain steady-state high confinement (H89 > 1.8). The L-H and H-L transitions are also analysed in terms of local edge parameters. The first results on isotope scaling of the H-mode power threshold in D-T and T-T plasmas are also presented, confirming the scaling.

Ion cyclotron range of frequencies (ICRF) heating experiments with the third harmonic deuterium resonance in the plasma centre have been carried out at JET. These experiments were the first to demonstrate third harmonic damping on initially Maxwellian plasmas. A record deuterium-deuterium (DD) fusion reaction rate for ICRF-only heating on JET was achieved. The discharges have been simulated with the PION code (a self-consistent calculation of the evolution of the distribution function and the ICRF power deposition). To obtain reasonable agreement between the measurements and the simulations, three additions to the PION code were necessary: (a) parasitic absorption at the plasma edge, (b) a particle loss term that removes particles above 4 MeV, i.e. the particles that were not confined in the discharges, and (c) a sawtooth model that accounts in a simplified way for the redistribution of fast ions at the sawtooth crashes. According to the simulations, high energy tail formation on the distribution function of the resonating ions plays a crucial role in the power absorption.

The full lower hybrid current drive (LHCD) system on the Joint European Torus (JET) has been operational since the start of the pumped divertor phase of JET in 1994. By a campaign of vacuum conditioning in the beginning of operation, the power transmission and conditioning of the fuli launcher was speeded up, as compared with the prototype launcher used in 1990/91. After vents of the tokamak, the conditioning can be recovered by -300 vacuum pulses over 3-4 days. Even after extensive conditioning and with efficient pumping, the power handling of the fuli LHCD system on plasma is limited by the maximum electric field in the waveguides, the limit being 400-500 kV/m.

All major JET systems have been fully commissioned for D-T and the DTE1 series of experiments has started with the D-T fuel mixture and operating conditions foreseen for ITER. In the area of ITER physics, significant results have been produced in both D-D and D-T. In D-D, the LH threshold power database has been extended, the bounds on edge-electron temperature and density in ELMy H-modes have been defined and the advantages of Types I and III ELMy discharges have been compared. In D-T plasmas, the isotope effect on H-mode threshold power and transport has been determined so that a more accurate assessment can be made of the ignition margin and heating requirements for ITER. Trace tritium experiments have provided first particle transport measurements and an assessment of the ITER reference ion-cyclotron resonance-frequency heating scenarios has been started, In the area of fusion performance, record D-D neutron yields have been obtained by controlling the plasma and current profiles in hot ion ELM-free H-modes and optimized shear modes. In D-T, internal transport barriers have been readily established in optimized shear discharges and Alfven eigenmodes have been observed.

The fraction of the coupled lower hybrid (LH) power absorbed in divertor plasmas in JET has been determined experimentally with a method utilizing the time derivative of the total stored energy (plasma and magnetic). This method can account for the power absorbed inside a normalized flux coordinate . The experimental LH absorption coefficient reaches 100% at low plasma densities, , and decreases to 25% at . The LH wave accessibility to the plasma core has been found to play an important role in determining the power absorption and the radial deposition profile. The decreasing absorption is correlated with a gradual shift of the LH power deposition profile, as determined by the hard x-ray profiles, towards the plasma periphery. Similar behaviour is found in ray-tracing + Fokker - Planck code calculations. The frequency spectrum of the LH pump wave, as determined by a probe outside the tokamak vessel, broadens strongly as the wave accessibility is reduced and the absorption drops.

The theoretical basis of the dimensionless parameter scaling technique is derived and the limitations in its application are discussed. The use of the technique is illustrated by the production on JET of a steady-state ITER similarity pulse having the same and collisionality as the ignited ITER. The key issue of the scaling of the transport with the main dimensionless parameter is discussed in detail. Finally, possible shortcomings of the technique are examined.

The results of a series of ITER simulation experiments on JET are described. A series of H-mode threshold experiments are shown to reproduce one of the standard power threshold scaling expressions that is being used to predict the power threshold in ITER. Then, from a series of experiments in which the Larmor radius scaling of ELMy H-modes is examined, it is concluded that the scaling of the confinement is gyro-Bohm-like provided the power levels are well above the threshold. Finally, we show that at high there is a dramatic reduction in the confinement at .