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International audience; The nuclear fusion research goal is to demonstrate the feasibility of fusion power for peaceful purposes. In order to achieve the conditions similar to those expected in an electricity-generating fusion power plant, plasmas with a temperature of several hundreds of millions of degrees must be generated and sustained for long periods. For this purpose, RF antennas delivering multi-megawatts of power to magnetized confined plasma are commonly used in experimental tokamaks. In the gigahertz range of frequencies, high power phased arrays known as " Lower Hybrid " (LH) antennas are used to extend the plasma duration. This paper reviews some of the technological aspects of the LH antennas used in the Tore Supra tokamak and presents the current design of a proposed 20 MW LH system for the international experiment ITER.

Alfven cascade (AC) eigenmodes excited by energetic ions accelerated with ion-cyclotron resonance heating in JET reversed-shear discharges are studied experimentally in high-density plasmas fuelled by neutral beam injection (NBI) and by deuterium pellets. The recently developed O-mode interferometry technique and Mirnov coils are employed for detecting ACs. The spontaneous improvements in plasma confinement (internal transport barrier (ITB) triggering events) and grand ACs are found to correlate within 0.2 s in JET plasmas with densities up to ∼5 × 10^19 m−3. Measurements with high time resolution show that ITB triggering events happen before ‘grand’ ACs in the majority of JET discharges, indicating that this improvement in confinement is likely to be associated with the decrease in the density of rational magnetic surfaces just before qmin(t) passes an integer value. Experimentally observed ACs excited by sub-Alfvenic NBI-produced ions with parallel velocities as low as V_||NBI ≈ 0.2 · V_A are found to be most likely associated with the geodesic acoustic effect that significantly modifies the shear-Alfven dispersion relation at low frequency. Experiments were performed with a tritium NBI-blip (short time pulse) into JET plasmas with NBI-driven ACs. Although considerable NBI-driven AC activity was present, good agreement was found both in the radial profile and in the time evolution of DT neutrons between the neutron measurements and the TRANSP code modelling based on the Coulomb collision model, indicating the ACs have at most a small effect on fast particle confinement in this case.

Fast protons can react with tritons in an endothermic nuclear reaction which can act as a source of neutrons in magnetically confined fusion plasmas. We have performed an experiment to systematically study this reaction in low tritium concentration (approximate to 1%) plasmas in the Joint European Torus. A linear dependence is found between excess neutron rate and tritium concentration when the DT fusion rate is low. We discuss the properties of the neutron emission, including anisotropy, from the proton-triton reaction in a fusion reactor environment and derive simple models for the calculation of the neutron yield from this reaction in terms of tritium density, fast ion temperature and fast ion energy content.

Results are presented from the JET Trace Tritium Experimental (TTE) campaign using minority tritium (T) plasmas (n(T)/n(D) 2 MA) and monotonic q-profiles. In CH discharges the gamma-ray emission decay times are much lower than classical (tau(Ts) + tau(alpha s)), indicating alpha confinement degradation, due to the orbit losses and particle orbit drift predicted by a 3-D Fokker-Planck numerical code and modelled using TRANSP.

In preparation for next step burning plasma devices such as ITER, experimental studies of instabilities and confinement of energetic ions were performed on Joint European Torus (JET) and on Mega-Amper Spherical Tokamak (MAST) with innovative diagnostic techniques, in conventional and shear-reversed plasmas, exploring a wide range of effects for energetic ions. A compendium of recent results testing capabilities of the present-day facilities for burning plasma relevant study is presented in this paper. 'Alpha tail' production using 3rd harmonic ion-cyclotron resonance heating (ICRH) of He-4 beam ions has been employed on JET for studying He-4 of the megaelectronvolt energy range in a 'neutron-free' environment. The evolution of ICRH-accelerated ions of He-4 with E >= 1.7 MeV and D with E >= 500 keV was assessed from nuclear gamma-ray emission born by the fast ions colliding with Be and C impurities. A simultaneous measurement of spatial profiles of fast He-4 and fast D ions relevant to ITER was performed for the first time in positive and strongly reversed magnetic shear discharges. Time-resolved gamma-ray diagnostics for ICRH-accelerated He-3 and H minority ions allowed changes in the fast ion distribution function to be assessed in the presence of unstable toroidal Alfven eigenmodes (TAEs) and sawteeth. A significant decrease of gamma-ray intensity from protons with E >= 5 MeV was detected during the 'tornado' modes. This was interpreted as 'tomado'-induced loss of fast ions with the drift orbit width, Delta(f), comparable to the minor radius of tokamak a. Experiments performed in the opposite case, Delta f/a = 500 keV, have shown excitation of numerous Alfven eigenmodes without a significant degradation of the fast ion confinement. The stabilizing effect of fast particles on 'monster' sawteeth was experimentally found to fail in low-density plasmas with high power ion cyclotron resonance frequency (ICRF)-heating. The transition from the 'monster' to short-period 'grassy' sawteeth was investigated with different ICRF phasing, which controls the pinch-effect and radial distribution of ICRF-accelerated ions. Instabilities excited by super-Alfvenic beam ions were investigated on the spherical tokamak MAST. Due to higher values of beta and a higher proportion of fast ions on MAST than on JET, a wider variety of modes and nonlinear regimes for the Alfven instabilities were observed, including the explosive TAE-regimes leading to the formation of hole-clump pairs on the fast ion distribution function. The MAST and START data showed that TAE and chirping modes decrease both in their mode amplitudes and in the number of unstable modes with increasing beta.

Experiments have been carried out on the JET tokamak to investigate fast 3He and hydrogen minority ion populations accelerated by ion cyclotron range of frequencies (ICRF) waves launched with multiple frequencies (i.e. up to four frequencies separated by up to ≈15%). This 'polychromatic' heating is compared with single-frequency, 'monochromatic', ICRF heating of reference discharges with similar power levels. Information on the fast ion populations is provided by two-dimensional gamma-ray emission tomography and the measurements are compared with numerical modelling. Polychromatic heating with resonances in the plasma centre (Rres ≈ R0) and on the low magnetic-field side (LFS) (Rres > R0) is found to produce predominantly high-energy standard trapped ions, while resonances on the high magnetic-field side (Rres < R0) increase the fraction of high-energy passing ions. Monochromatic heating with a central resonance produces stronger gamma-ray emission with the maximum emission in the midplane close to, and on the LFS of, the resonance, in agreement with the calculated radial distribution of fast ion orbits. Both the fast ion tail temperature and energy content are found to be lower with polychromatic waves. Polychromatic ICRF heating has the advantage of producing smaller-amplitude and shorter-period sawteeth, consistent with a lower fast ion pressure inside the q = 1 surface, and higher ion to electron temperature ratios.

Current hole plasmas in JET are those in which the current density within r/a < 0.3 is close to zero. Tritium ions injected quasi-tangentially into such plasmas can fulfil a stagnation condition whereby their vertical drift is cancelled by the poloidal component of their parallel velocity. These ions remain trapped at approximately 0.2 m from the plasma axis and can be detected by a distortion in the neutron emission profile. Numerical modelling of the steady-state distribution reproduces the experimental results while the decay of neutron emission after the cessation of injection is found to be sensitive to small changes in the q-profile.

Efficient coupling of lower hybrid (LH) waves in conditions close to those foreseen in ITER has been obtained in advanced scenario plasmas in the JET tokamak. Up to 3 MW of lower hybrid current drive (LHCD) power has been coupled at a distance between the separatrix and the launcher of 11 cm, in the presence of edge localized mode activity. The key to the improved LH wave coupling is local control of the Scrape-Off-Layer (SOL) density through gas injection in the region magnetically connected to the launcher. This increases the electron density in front of the launcher so as to improve the coupling of the LH waves, i.e. reduce the reflected power in the launcher. The average power reflection coefficient was 5.7% with gas injection, at 11 cm distance between the separatrix and the launcher. A change in the gas injection design has made the gas puffing more efficient, making the use of D2 injection possible as an alternative to CD4. At similar injected electrons/s rates, D2 gives higher electron density in the SOL than CD4, resulting in better LH coupling with D2. The possibility of using D2 instead of the earlier used CD4 is an encouraging result in view of ITER operation, as CD4 may not be compatible in ITER due to the problem of tritium retention in deposited carbon layers.

Low aspect ratio plasmas in devices such as the mega ampere spherical tokamak (MAST) are characterized by strong toroidicity, strong shaping and self fields, low magnetic field, high beta, large plasma flow and high intrinsic E × B flow shear. These characteristics have important effects on plasma behaviour, provide a stringent test of theories and scaling laws and offer new insight into underlying physical processes, often through the amplification of effects present in conventional tokamaks (e.g. impact of fuelling source and magnetic geometry on H-mode access). The enhancement of neoclassical effects makes MAST ideal for the study of particle pinch processes and neoclassical resistivity corrections, which can be assessed with unique accuracy. MAST data have an important influence on scaling laws for confinement and H-mode threshold power, exerting strong leverage on the form of these scaling laws (e.g. scaling with aspect ratio, beta, magnetic field, etc). The high intrinsic flow shear is conducive to transport barrier formation by turbulence suppression. Internal transport barriers are readily formed in MAST with both co- and counter-NBI, and electron and ion thermal diffusivities have been reduced to the ion neoclassical level. The strong variation in toroidal field (~ × 5 in MAST) between the inboard and outboard plasma edges, provides a useful test of edge models prompting, for example, a comparison of inboard and outboard scrape-off-layer transport to highlight magnetic field effects. Low aspect ratio plasmas are also an ideal testing ground for plasma instabilities, such as neoclassical tearing modes, edge localized modes (ELMs) and Alfven eigenmodes, which are readily generated due to the supra-Alfvenic ion population. Examples of how MAST is providing new insights into such instabilities (e.g. ELM structure) are described.

This paper is an overview of recent results relating to turbulent particle and heat transport, and to the triggering of internal transport barriers (ITBs). The dependence of the turbulent particle pinch velocity on plasma parameters has been clarified and compared with experiment. Magnetic shear and collisionality are found to play a central role. Analysis of heat transport has made progress along two directions: dimensionless scaling laws, which are found to agree with the prediction for electrostatic turbulence, and analysis of modulation experiments, which provide a stringent test of transport models. Finally the formation of ITBs has been addressed by analysing electron transport barriers. It is confirmed that negative magnetic shear, combined with the Shafranov shift, is a robust stabilizing mechanism. However, some well established features of internal barriers are not explained by theory.

The large physical size of the JET tokamak, its heating systems and diagnostics, and its capability to operate with full deuterium-tritium (D-T) plasmas, including high-power tritium neutral beam injection (NBI), give it unique possibilities in fast particle research in fusion plasmas. These have already been used to generate significant (2-3 MW level) power in fusion a-particles in the 1997 D-T campaign. Recent JET experiments have concentrated on two important scenarios of relevance to next-step tokamak devices: the ELMy H-mode plasmas and plasmas with strong internal transport barriers (ITBs). The achieved progress will help in preparation for a possible second D-T experiment on JET. Fast particle studies have also been carried out recently using ion cyclotron resonance heating (ICRH)-accelerated particles and external-excitation methods to study Alfven eigenmodes (AEs). Looking towards the future, the capability of JET will be enhanced by upgrades to the NBI system, ICRH system and various diagnostics. Results of the first JET D-T experiment (DTE1) form a basis on which to elaborate a second D-T experiment (DTE2) which could be proposed after these enhancements. The alpha-physics part of this programme would be divided between the investigation of alpha-particle confinement, heating and loss processes in the 'integrated scenarios' (where the discharge is as close as possible to an ITER-relevant scenario), and dedicated 'alpha-physics' experiments, with specially prepared plasmas. In ELMy H-mode plasmas the fusion performance could roach Q(=P-fusion/P-input) of similar to0.33 at the highest combined heating powers, corresponding to similar to 6x10(-4), allowing a test of the margins of TAE stability in quasi-steady-state conditions. The integrated-scenario fast particle programme could concentrate on the instabilities and heating in plasma regimes with strong steady-state ITBs, with expected Q values similar to0.58 and similar to2x10(-3), demonstrating the compatibility of these operating scenarios with alpha-effects. Excitation of TAEs by alpha-particles in the plasma core could also be studied in such integrated scenarios. An issue which will receive attention is the confinement of MeV energy ions in the centre of ITB plasmas with strongly reversed shear, where the low current density in the centre may lead to the alpha-particles entering loss orbits. In preparation for a D-T campaign, studies of triton burn-up in deuterium ITB plasmas will begin in the 2002 experimental campaigns. Special 'afterglow' experiments to measure TAEs after the termination of the (stabilizing) NBI have already been explored in JET deuterium ITB scenarios and would be planned for DTE2. It is intended to develop special versions of ITB plasmas with dominant ion heating which would maximize the sensitivity to degradation of alpha-heating effects.

An overview of the recent results of Lower Hybrid (LH) experiments at JET with the ITER-like wall (ILW) is presented. Topics relevant to LH wave coupling are addressed as well as issues related to ILW and LH system protections. LH wave coupling was studied in conditions determined by ILW recycling and operational constraints. It was concluded that LH wave coupling was not significantly affected and the pre-ILW performance could be recovered after optimising the launcher position and local gas puffing. SOL density measurements were performed using a Li-beam diagnostic. Dependencies on the D2 injection rate from the dedicated gas valve, the LH power and the LH launcher position were analysed. SOL density modifications due to LH were modelled by the EDGE2D code assuming SOL heating by collisional dissipation of the LH wave and/or possible ExB drifts in the SOL. The simulations matched reasonably well the measured SOL profiles. Observations of arcs and hotspots with visible and IR cameras viewing the LH launc...

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.

A linear empirical threshold condition ωE×B/γITG>0.68s-0.095 has been found for the onset of the ion internal transport barriers in the JET optimised shear database. Here, s is the magnetic shear, ωE×B the flow shearing rate and γITG is an approximate of the linear growth rate of the ion temperature gradient instability. The present empirical threshold condition for the ITB formation will provide a first clear indication of the strong correlation of s and ωE×B at the ITB transition. The empirical analysis consists of ITB discharges from a wide plasma parameter range; the toroidal magnetic field varies between 1.8-4.0 T, the auxiliary heating power between 10-30 MW and the diamagnetic energy between 3-12 MJ. The predictive simulations of several ITB discharges with the empirical ITB threshold condition reproduce the experiments with time averaged prediction errors of the order of 10-25% in Ti and Te profiles and 10-15% in ne profiles as well as the toroidal flow velocity with errors of the order of 10-20%. The simulated times of the onset of the ITB compared to the experimental ones are typically within 0.4 s and the simulated ITB widths within 0.1 in r/a throughout the whole simulations.

Modelling of LHCD with transport calculations is performed with the JETTO transport code, which has been upgraded by implementing the Fast Ray Tracing Code to calculate self-consistent LH power deposition profiles. Heat and particle transport models that are able to reproduce the experimental JET temperature and density profiles are used in JETTO for predictive high performance modelling. Application of 3.5 MW LHCD power provides an inverted q profile across 50-70% of the plasma radius whereas, without LHCD, the q profile is monotonic during the flat-top phase. The results predict that the fusion power is about 60% higher for high performance DT plasmas in the optimized shear scenario with 3.5 MW LHCD applied during the high performance phase than without LHCD at Bt = 3.4 T and Ip = 3.9 MA on JET. In addition, the width of the internal transport barrier (ITB) is 0.25-0.30 m larger and the ITB can be sustained for a longer time with LHCD.

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.

Ideal MHD limits to beta and bootstrap fraction are computed for pressure profiles similar to those of JET discharges with both internal and H mode transport barriers. Several non-monotonic current profiles are tested, and the pressure profile is constrained so that no negative current drive is required in steady state. Calculations are made both with and without an ideally conducting wall. The main result is that, for such peaked pressure profiles, the limits to beta and bootstrap fraction improve at low internal inductance, in particular when wall stabilization is taken into account. The highest limits to beta, and often also to normalized beta, occur for the maximum plasma current. There is also a weak dependence on qmin, for which three favourable regions have been identified. A highly advantageous region is found at qmin 1.6, where the limits to β* are 7.0% with, and 4.8% without, wall stabilization. The corresponding limits are 68 and 50%, respectively, for the bootstrap fraction and 4.0 and 2.9 for the normalized beta. These equilibria have low internal inductance, li = 0.62. For higher inductance, an optimum occurs when qmin 1.2, where the limit to β* is 5.3% with a wall and 4.7% without. The corresponding bootstrap fractions are about 46 and 38%, respectively. A third type of equilibrium that is interesting for steady state operation has qmin 2.1 and low inductance. Here the β* limits are lower, 4.9 and 3.4%, but the bootstrap fractions are higher, 77 and 60%.

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.

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.

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.

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.

An important effect observed in ILW discharges is the change of q = 1 MHD activity in hybrid discharges. The q = 1 activity with the JET C-wall had usually shown no effect on plasma confinement, except for NTM triggering, its main experimental signature being a rotating kink activity interleaving sawteeth crashes. In the C-Wall hybrid scenario sawteeth had low amplitude and frequency, and the rotating precursor had often fishbone characteristic. The radial profile of the mode fluctuation has been calculated using a coherence technique between one pick up coil and 96 ECE measurements along the red L.o.S in Figure 1. As explained in ref.[1], the coherence phase radial profile shows a -jump whenever an island is crossed moving along the plasma radius, this effect is due to the radial displacement of magnetic field lines produced by the rotating island, which has opposite directions on the outer side of the island x-point than on the inner side of the x-point. In the C-Wall hybrid pulses the ECE fluctuation at the q = 1 magnetic frequency did not show any radial phase jump, as can be seen in Figure 2, and therefore had kink characteristic.

In this paper important aspects of Lower Hybrid (LH) operation with the ITER Like Wall (ILW) [1] at JET are reported. Impurity release during LH operation was investigated and it was found that there is no significant Be increase with LH power. Concentration of W was analysed in more detail and it was concluded that LH contributes negligibly to its increase. No cases of W accumulation in LH-only heating experiments were observed so far. LH wave coupling was studied and optimised to achieve the level of system performance similar to before ILW installation. Measurements by Li-beam were used to study systematic dependencies of the SOL density on the gas injection rate from a dedicated gas introduction module and the LH power and launcher position. Experimental results are supported by SOL transport modelling. Observations of arcs in front of the LH launcher and hotspots on magnetically connected sections of the vessel are reported. Overall, a relatively troublefree operation of the LH system up to 2.5MW of coupled Radio Frequency (RF) power in L-mode plasma was achieved with no indication that the power cannot be increased further., Comment: 21 pages, 9 figures. This is an author-created, un-copyedited version of an article accepted for publication in Plasma Physics & Controlled Fusion. IOP Publishing Ltd is not responsible for any errors or omissions in this version of the manuscript or any version derived from it

A new code has been developed and benchmarked with data from lower hybrid current drive (LHCD) experiments in JET. The model includes stochastic radial diffusion of electrons and scattering of lower hybrid (LH) waves. The code is validated by comparing calculated and experimental current drive efficiencies as well as calculated and measured hard X ray bremsstrahlung emission radial profiles. Reasonable agreement has been found for full LH current driven cases and LH plus moderate DC electric field

Large and infrequent collapse events have been observed in high βN advanced tokamak (AT) plasmas in JET. Although they have features similar to large ELMs, they were triggered by core MHD. They caused a considerable loss of the plasma thermal and fast particle energy (∼10% of the total stored energy), but the heat load in the divertor due to these collapse events was small as a fraction of the plasma energy loss compared with regular type-I ELMs. Instead, significant heating of the main chamber wall was observed. A large, toroidally asymmetric, increase in the neutral gas pressure outside the plasma was observed after such events, which caused arcs in the lower hybrid (LH) and ion cyclotron (IC) heating systems and increased reionization in the neutral beam (NB) injectors. The collapses resulted in a reduction in the electron and ion temperatures and toroidal rotation of the whole plasma, a rise in Z eff; and a sufficiently large increase in the peripheral electron density to completely black-out the ECE emission from the plasma core. These features have been modelled to gain an understanding of the plasma behaviour associated with these collapse events and the implication for the operation of AT plasma scenarios with high additional heating power will be discussed.

Lower hybrid (LH) wave coupling is modified and usually degraded when the system is powered simultaneously with the ion cyclotron range of frequency (ICRF) antennas magnetically connected to the launcher. This has been attributed to scrape-off layer density modifications by the RF sheaths. LH reflection coefficients dependences on various parameters are investigated and shown to be consistent with RF sheath physics. Gas puffing near the launcher has been used to improve the coupling of LH waves.

In this paper, significant results in the heating and current drive domains obtained at JET in the past few years following systems upgrade and dedicated experimental time, will be reviewed. Firstly, an overview of the new Ion Cyclotron Resonance Frequency (ICRF) heating capabilities will be presented i.e. results from the ITER‐Like ICRF antenna (ILA), the use of External Conjugate‐T and 3dB hybrid couplers to increase the ICRF power during ELMy H‐mode. Furthermore, experiments to study the influence of the phasing of the ICRF antenna on power absorption and coupling will be described. Looking at Low Hybrid (LH) issues for ITER, the effect of the location of gas injection on the LH coupling improvement at large launcher‐separatrix distances will be discussed as the possibility to operate at ITER‐relevant power densities. Experiments to characterise the LH power losses in the Scrape‐Off‐Layer (SOL) and to determine the LH wave absorption and current drive using power modulation will be shown. Finally, plas...

Good coupling of lower hybrid (LH) waves has been demonstrated in different H-mode scenarios in JET, at high triangularity (delta similar to 0.4) and at large distance between the last closed flux surface and the LH launcher ( up to 15 cm). Local gas injection of D(2) in the region magnetically connected to the LH launcher is used for increasing the local density in the scrape-off layer ( SOL). Reciprocating Langmuir probe measurements magnetically connected to the LH launcher indicate that the electron density profile flattens in the far SOL during gas injection and LH power application. Some degradation in normalized H-mode confinement, as given by the H98(gamma,2)-factor, could be observed at high gas injection rates in these scenarios, but this was rather due to total gas injection and not specifically to the local gas puffing used for LH coupling. Furthermore, experiments carried out in L-mode plasmas in order to evaluate the effect on the LH current drive efficiency, when using local gas injection to improve the coupling, indicate only a small degradation (Delta I(LH)/I(LH) similar to 15%). This effect is largely compensated by the improvement in coupling and thus increase in coupled power when using gas puffing.

In the 2006 experimental campaign, progress has been made on JET to operate non-inductive scenarios at higher applied powers (31 MW) and density (n(1) similar to 4 x 10(19) m(-3)), with ITER-relevant safety factor (q(95) similar to 5) and plasma shaping, taking advantage of the new divertor capabilities. The extrapolation of the performance using transport modelling benchmarked on the experimental database indicates that the foreseen power upgrade (similar to 45 MW) will allow the development of non-inductive scenarios where the bootstrap current is maximized together with the fusion yield and not, as in present-day experiments, at its expense. The tools for the long-term JET programme are the new ITER-like ICRH antenna (similar to 15 MW), an upgrade of the NB power (35 MW/20s or 17.5 MW/40s), a new ITER-like first wall, a new pellet injector for edge localized mode control together with improved diagnostic and control capability. Operation with the new wall will set new constraints on non-inductive scenarios that are already addressed experimentally and in the modelling. The fusion performance and driven current that could be reached at high density and power have been estimated using either 0D or 1-1/2D validated transport models. In the high power case (45 MW), the calculations indicate the potential for the operational space of the non-inductive regime to be extended in terms of current (similar to 2.5 MA) and density (n(1) > 5 x 10(19) m(-3)), with high beta(N) (beta(N) > 3.0) and a fraction of the bootstrap current within 60-70% at high toroidal field (similar to 3.5 T).

The non-solenoid start-up technique reported here relies on a double mode conversion for electron Bernstein wave (EBW) excitation. It consists of the mode conversion of the ordinary mode, entering the plasma from the low field side of the tokamak, into the extraordinary (X) mode at a mirror-polarizer located at the high field side. The X mode propagates back to the plasma, passes through electron cyclotron resonance and experiences a subsequent X to EBW mode conversion near the upper hybrid resonance. Finally the excited EBW mode is totally absorbed at the Doppler shifted electron cyclotron resonance. The absorption of EBW remains high even in cold rarefied plasmas. Furthermore, EBW can generate significant plasma current giving the prospect of a fully solenoid-free plasma start-up. First experiments using this scheme were carried out on MAST [V. Shevchenko et al, Nuclear Fusion 50, 022004 (2010)]. Plasma currents up to 33 kA have been achieved using 28 GHz 100kW 90ms RF pulses. Recently experimental results were extended to longer RF pulses showing further increase of plasma currents generated by RF power alone. A record current of 73kA has been achieved with 450ms RF pulse of similar power. The current drive enhancement was mainly achieved due to RF pulse extension and further optimisation of the start-up scenario., Comment: 22 pages, 11 figures

A strong ITB was sustained in a reversed shear discharge in JET during a long time interval after a significant reduction in plasma heating power. The observed ITB evolution is reminiscent of the effect of hysteresis, associated in theory with turbulence suppression due to shear of the plasma rotation. The mechanism of ITB sustainment was analysed. Modelling of the plasma heating and current profile was done using the TRANSP and JETTO transport codes. The resulting q-profile evolution was verified by comparison with the pitch angle ? = Bp/Bt deduced from motional Stark effect measurements (where Bp and Bt are the poloidal and toroidal magnetic field, respectively). Turbulence stability was analysed using the gyro-kinetic code Kinezero. It was shown that the strong negative magnetic shear produced by LHCD and sustained mainly by the bootstrap current in the plasma core is responsible for the turbulence stabilization. Stabilization due to shear of the plasma rotation and finite ? -stabilization play a complimentary role. The effect of bootstrap and LH driven currents on magnetic shear in JET discharges was analysed.

Gamma-ray spectra from nuclear reactions between fusion-born alpha ($\ensuremath{\alpha}$) particles and Be impurities were measured for the first time in deuterium-tritium plasmas in the Joint European Torus. The time dependence of the measured spectra allowed the determination of the density evolution of fast $\ensuremath{\alpha}$ particles. Correlation between the decay time of the $\ensuremath{\gamma}$-ray emission and the plasma parameters in different plasma scenarios was established. Results are consistent with classical slowing down of the $\ensuremath{\alpha}$ particles in discharges with high plasma currents and monotonic q-profiles. In low plasma current discharges and in the discharges with large on-axis current holes (extreme reversal central magnetic shear), the $\ensuremath{\gamma}$-ray emission decay times are shorter than the classical slowing down times, indicating an $\ensuremath{\alpha}$-particle confinement degradation in such discharges in line with theoretical predictions.

ELMy H-mode experiments at JET in 2000/mid-2002 have focused on discharges with normalized parameters for plasma density, energy confinement and beta similar to those of the ITER Q(DT) = 10 reference regime (n/n(GW) similar to 0.85, H-98(y,H-2) similar to 1, beta(N) similar to 1.8). ELMy H-mode plasmas have been realized reaching or even exceeding those parameters in steady-state conditions (up to similar to5 s or 12tau(epsilon)) in a reproducible way and only limited by the duration of the additional heating phase. These results have been obtained (a) in highly triangular plasmas, by increasing the average triangularity delta towards the ITER reference value (delta similar to 0.5), and (b) in plasmas at low triangularity (delta similar to 0.2) by seeding of Ar and placing the X-point of the plasma on the top of the septum. Pellet injection from the high field side is a third method yielding high density and high confinement, albeit not yet under steady-state conditions. In highly triangular plasmas the influence of input power, plasma triangularity and impurity seeding with noble gases has been studied. Density profile peaking at high densities has been obtained in (a) impurity seeded low triangularity discharges, (b) ELMy H-modes with low levels of input power and (c) discharges fuelled with pellet injection from the high field side. New ELM behaviour has been observed in high triangularity discharges at high density, opening a possible route to ELM heat load mitigation, which can be further amplified by Ar impurity seeding. Current extrapolations of the ELM heat load to ITER show possibly a window for Type I ELM operation. Confinement scaling studies indicate an increase in confinement with triangularity and density peaking, and a decrease in confinement with the Greenwald number. In addition, experiments in H isotope and He indicate tau(E) proportional to M(0.19)Z(-0.59). The threshold power for the L-H transition in He plasmas shows the same parametric dependence as in D plasmas, but with a 50% higher absolute value.

In JET, advanced tokamak research mainly focuses on plasmas with internal transport barriers (ITBs) that are strongly influenced by the current density profile. A previously developed optimized shear regime with low magnetic shear in the plasma centre has been extended to deeply negative magnetic shear configurations. High fusion performance with wide ITBs has been obtained transiently with negative central magnetic shear configuration: HIPB98(y,2) ∼ 1.9, βN = 2.4 at IP = 2.5 MA. At somewhat reduced performance, electron and ion ITBs have been sustained in full current drive operation with 1 MA of bootstrap current: HIPB98(y,2) ∼ 1, βN = 1.7 at Ip = 2.0 MA. The ITBs were maintained for up to 11 s for the latter case. This duration, much larger than the energy confinement time (37 times larger), is already approaching a current resistive time. New real-time measurements and feedback control algorithms have been developed and implemented in JET for successfully controlling the ITB dynamics and the current density profile in the highly non-inductive current regime.

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 significant step towards demonstrating the feasibility of coupling Lower Hybrid (LH) waves in ITER has been achieved in the latest LH current drive experiments in JET. The local electron density in front of the LH launcher was increased by injecting gas (D2 or CD4) from a dedicated gas injection module magnetically connected to the launcher. PLHCD=3MW was coupled with an average reflection coefficient of 5%, at a distance between the last closed flux surface and the launcher of 10cm, in plasmas with an internal transport barrier (ITB) and H‐mode edge, with type I and type III ELMs. Following a modification of the gas injection system, in order to optimise the gas localisation with respect to the LH launcher, injection of D2 proved to be more efficient than CD4. A D2 flux of 5−8×1021el/s provided good coupling conditions at a clearance of 10cm, while when using CD4, a flux of 12×1021el/s was required at 9cm. The plasma performance (neutron rate, H‐factor, ion temperature) was similar with D2 and CD4. An ...

Experiments have been carried out on the JET tokamak to investigate differences between multiple and single frequency ICRH operation with ICRH power in the range of 3 to 8 MW using H and 3He minority heating. High‐energy neutral particle analysis and gamma‐ray emission tomography are used to measure fast ions, including their radial localisation. For 3 MW of 3He minority heating, the fast 3He ion profile is broader according to the gamma emission data and the fast ion tail temperature Ttail and energy content Wfast are lower with polychromatic ICRH than monochromatic ICRH. Polychromatic ICRH has the advantage of producing smaller‐amplitude and shorter‐period sawteeth, consistent with a lower fast ion pressure inside q = 1, and higher Ti/Te ratios (i.e. similar Ti at lower Te). At high powers with resonances in the centre and/or on the low field side, the data indicates a larger fraction of trapped ions and a lower Ttail for polychromatic ICRH. Experimental results are compared with theoretical predictions.

Quasistationary operation has been achieved on the Joint European Torus tokamak in internal-transport-barrier (ITB) scenarios, with the discharge time limited only by plant constraints. Full current drive was obtained over all the high performance phase by using lower hybrid current drive. For the first time feedback control on the total pressure and on the electron temperature profile was implemented by using, respectively, the neutral beams and the ion-cyclotron waves. Although impurity accumulation could be a problem in steady state ITBs, these experiments bring some elements to answer to it.

RF heating and current drive with ion cyclotron waves and Lower Hybrid waves have been crucial for the development of the Optimized Shear scenario on JET to high performance. Peaked electron temperature profiles and improved energy confinement could be obtained with electron heating both from LHCD and ICRH during plasma current ramp up. ICRH and NBI comparisons allow to separate heating and fueling and suggest a dominant role of core heating in the formation of an Internal Transport Barrier (ITB). ICRH and NBI powers are there equivalent. Pressure profile control by varying the composition of centrally peaked ICRF and broader NBI deposition improves MHD stability. Current profile modifications in a wide range have been obtained with LHCD and in combination with NBI during current ramp up. During the high performance phase, however, LH coupling degrades strongly due to the steep edge density gradient resulting in a drop of the density in front of the LH antenna to the cut-off density. High fusion performance achieving simultaneously high beta values and bootstrap currents is predicted in scenario modeling using pressure and current profile control with ICRF and LHCD.

Reliable coupling of the lower hybrid current drive (LHCD) to H-mode plasmas in JET is made feasible through a dedicated gas injection system, located at the outer wall and magnetically connected to the antenna (Pericoli Ridolfini et al 2004 Plasma Phys. Control. Fusion 46 349, Ekedahl et al 2005 Nucl. Fusion 45 351, Ekedahl et al 2009 Plasma Phys. Control. Fusion 51 044001). An experiment was carried out in JET in order to investigate whether a gas injection from the top of the torus, as is foreseen for the main gas injection in ITER, could also provide good coupling of the LH waves if magnetically connected to the antenna. The results show that a top gas injection was not efficient for providing a reliable LHCD power injection, in spite of being magnetically connected and in spite of using almost twice the amount of gas flow compared with the dedicated outer mid-plane gas puffing system. A dedicated gas injection system, set in the outer wall and magnetically connected to the LHCD antenna, is therefore recommended in order to provide the reliable coupling conditions for an LHCD antenna in ITER.

Lower hybrid current drive (LHCD) experiments performed at a density close to that required for the steady-state (SS) scenario are reported. On C-Mod, FTU and Tore Supra, a strong decay of the bremsstrahlung emission is observed when the density is increased, much faster that the prediction of LHCD modelling. On JET, LH power deposition is also found to be sensitive to the plasma density: LH power modulation indicates that the power deposition moves to the very edge of the plasma (r/a similar to 0.9) when the density approaches the requirement of the JET SS scenario. From this experiment but also from the reconstruction of the electron cyclotron emission spectrum, the decrease in the LHCD efficiency with density is also found. From LHCD modelling of different JET pulses performed at different densities and wave parallel refraction indexes, it is concluded that the wave accessibility condition is not the key parameter for explaining the decrease in the efficiency. C-Mod, FTU and Tore Supra experiments indicate that the plasma edge parameters, namely density and temperature but also fluctuations, are affecting the efficiency via loss mechanisms which are likely to be collisional damping (C-Mod), parametric decay instabilities or wave scattering (FTU/Tore Supra).

A new ITER-relevant lower hybrid current drive (LHCD) launcher, based on the passive-active-multijunction (PAM) concept, was brought into operation on the Tore Supra tokamak in autumn 2009. The PAM launcher concept was designed in view of ITER to allow efficient cooling of the waveguides, as required for long pulse operation. In addition, it offers low power reflection close to the cut-off density, which is very attractive for ITER, where the large distance between the plasma and the wall may bring the density in front of the launcher to low values. The first experimental campaign on Tore Supra has shown extremely encouraging results in terms of reflected power level and power handling. Power reflection coefficient −2, i.e. its design value at f = 3.7 GHz. In addition, 2.7 MW has been coupled at a plasma–launcher distance of 10 cm, with a power reflection coefficient

The lower hybrid (LH) power deposition and the current drive (CD) efficiency were assessed by the application of modulated LH power. Density and magnetic field scans were performed and the response of the electron temperature provided by the available electron cyclotron emission diagnostic was investigated by means of fast Fourier transform analysis. An innovative technique based on a comparison between modelled and experimental data was developed and used in the study. The LH waves are absorbed by fast electrons with energies of a few times the thermal one, causing a modification in the electron distribution function (EDF) by creating a plateau in the parallel direction. The phase of the temperature perturbations, ϕ, as well as the ratio between the amplitudes of the third and the main harmonics, δT e3/δT e1, are found to be strongly affected by the plateau of the EDF as the broader the plateau the larger |ϕ|, (ϕ < 0), and the smaller δT e3/δT e1 are. Transport and Fokker–Planck modelling was used to support this conclusion as well as to interpret the experimental data and hence to assess the LHCD efficiency and deposition profile. The results from the analysis are consistent with broad off-axis LH power deposition profile. For densities between 1 × 1019 and 4 × 1019 m−3, which is the accessibility limit at the highest magnetic field discharges, a gradual shift of the maximum of the power deposition to the periphery and a degradation of the CD efficiency was observed.

Trace tritium experiments (TTE) on JET were analysed using Monte Carlo modelling of the neutron emission resulting from the neutral beam injection (NBI) of short (~300 ms) tritium (T) beam blips into reversed shear, hybrid ELMy H-mode and L-mode deuterium plasmas for a wide range of plasma parameters. The calculated neutron fluxes from deuterium–tritium (DT) reactions could only be made consistent with all plasmas by applying an artificial reduction of the T beam power in the modelling of between 20% and 40%. A similar discrepancy has previously been observed in both JET (Gorini et al 2004 Proc. 31st EPS Conf. on Plasma Physics (London, UK) vol 28G (ECA)) and TFTR (Ruskov et al 1999 Phys. Rev. Lett. 82 924), although no mechanism has yet been found that could explain such a difference in the measured T beam power. Applying this correction in the T beam power, good agreement between calculated and measured DT neutron emission profiles was obtained in low to moderate line averaged density ELMy H-Mode plasmas assuming that the fast beam ions experience no, or relatively small, anomalous diffusion (Dan 0.5 m2 s−1). However, the modelled neutron profiles do not agree with measurements in higher density plasmas using the same assumption and the disagreement between the measured and calculated shape of the neutron profile increases with plasma density. In this paper it is demonstrated that large anomalous losses of fast ions have to be assumed in the simulations to improve agreement between experimental and simulated neutron profiles, characterized by the goodness of fit. Various types of fast ion losses are modelled to explain aspects of the data, though further investigation will be required in order to gain a more detailed understanding of the nature of those anomalous losses.

Tritium transport in edge localized mode (ELM) high confinement (H-mode) plasmas is analyzed here as a function of density for discharges from the recent trace tritium experimental campaign performed on Joint European Torus. In this campaign small amounts of tritium have been puffed or injected (with neutral beam injectors) into deuterium plasmas [K.-D. Zastrow, J. M. Adams, Yu. Baranov , Plasma Phys. Controlled Fusion 46, B255 (2004)]. Information about the tritium has been obtained from the evolution of the profiles of neutron emission simulated via the TRANSP [R. J. Goldston, D. C. McCune, H. H. Towner, S. L. Davis, R. J. Hawryluk, and G. L. Schmidt, J. Comput. Phys. 43, 61 (1981)] and SANCO (L. Lauro-Taroni, B. Alper, R. Giannella, K. Lawson, F. Marcus, M. Mattioli, P. Smeulders, and M. Von Hellermann, Proceedings of the 21st European Conference on Controlled Fusion and Plasma Physics, Montpelier, France, 1994) codes. A strong inverse correlation of tritium transport with plasma density is found in this analysis. The low tritium transport at high density is close to neoclassical values while the transport becomes strongly anomalous in low density plasmas. The thermal transport does not exhibit such a strong density dependence, leading to a varying ratio of thermal to tritium transport in these discharges. An interpretation of the density effects on the trace tritium transport, partially based on the test particle simulations in plasmas with stochastic magnetic field, is proposed. A simple model for the tritium diffusion coefficient and convective velocity, which includes the modification of the neoclassical particle diffusion in presence of electromagnetic turbulence [A. I. Smolyakov and P. N. Yushmanov, Nucl. Fusion 35, 383 (1993)] completed with an empirical density dependence, is developed. This model has positive beta dependence in agreement with the results of the similarity experiments performed for trace tritium transport. (c) 2005 American Institute of Physics.

Data for the calculation of lower-hybrid wave ray trajectories in a tokamak plasma are quoted. It is demonstrated that under actual conditions the refractive index parallel to the total magnetic field varies considerably as the wave is propagated along the trajectory. This implies that there is a considerable change in the conditions of accessibility of the lower-hybrid resonance in a tokamak, as compared with a one-dimensional inhomogeneous plasma model. The accessibility of the resonance area is a strong function of the current distribution in the plasma. When the distribution is highly peaked the central area of the column becomes inaccessible to waves with any initial retardation. When the current profile is flatter, the conditions for accessibility are made easier to fulfil.

The stereospecific fragmentation of 3-(2,2-dichlorovinyl)-2,2-dimethylcyclopropanecarboxylic acid α-cyano-3-phenoxybenzyl ester was found under conditions of resonant capture of electrons of epithermal energies with dissociation of HCN. The stereospecific collection of fragment ions allows not only the identification of one of the diastereoisomers, but also the quantitative (≥5%) determination of it in a mixture of racemates.