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The impact of the parallel flow shear on the tokamak plasma stability and turbulent transport driven by the ion temperature gradient (ITG) modes is analyzed by means of local gyrokinetic numerical analyses. It is shown that the parallel flow shear increases the ITG growth rate in the linear regime, and induces a broadening and shift of the radial spectrum. Then, the different effects of the finite parallel shear on the ITG turbulence characteristics are deeply analyzed in the nonlinear regime. These studies highlight that a reduction of the thermal-ion turbulent heat flux is induced by a complex mechanism involving the nonlinear generation of an enhanced zonal flow activity. Indeed, the turbulent sources of the zonal flows are increased by the introduction of the finite parallel flow shear in the system, beneficially acting on the saturation level of the ITG turbulence. The study has been carried out for the Waltz standard case below the critical threshold of the destabilization of the parallel velocity gradient instability, and then generalized to a selected pulse of a recent JET scenario with substantial toroidal rotation in the edge plasma region. It is, thus, suggested that the investigated complex mechanism triggered by the finite parallel flow shear reducing the ITG turbulent heat fluxes could be complementary to the well-established perpendicular flow shear in a region with sufficiently large plasma toroidal rotation.

The analysis of the current ramp-down phase of JET plasmas has revealed the occurrence of additional magnetic oscillations in pulses characterized by large magnetic islands. The frequencies of these oscillations range from 5 to 20 kHz , being well below the toroidal gap in the Alfvén continuum and of the same order as the low-frequency gap opened by plasma compressibility. The additional oscillations only appear when the magnetic island width exceeds a critical threshold, suggesting that the oscillations could tap their energy from the tearing mode (TM) by a non-linear coupling mechanism. A possible role of fast ions in the excitation process can be excluded, being the pulse phase considered in the observations characterized by very low additional heating. The calculation of the coupled Alfvén–acoustic continuum in toroidal geometry suggests the possibility of beta-induced Alfvén eigenmodes (BAEs) rather than beta-induced Alfvén–acoustic eigenmodes. As a main novelty compared to previous work, the analysis of the electron temperature profiles from electron cyclotron emission has shown the simultaneous presence of magnetic islands on different rational surfaces in pulses with multiple magnetic oscillations in the low-frequency gap of the Alfvén continuum. This observation supports the hypothesis of different BAE with toroidal mode number n = 1 associated with different magnetic islands. As another novelty, the observation of magnetic oscillations with n = 2 in the BAE range is reported for the first time in this work. Some pulses, characterized by slowly rotating magnetic islands, exhibit additional oscillations with n = 0, likely associated with geodesic acoustic modes (GAMs), and a cross-spectral bicoherence analysis has confirmed a non-linear interaction between TM, BAE and GAM, with the novelty of the observation of multiple triplets (twin BAEs plus GAM), due to the simultaneous presence of several magnetic islands in the plasma.

Changes in global and edge plasma parameters (H 98(y,2), dimensionless collisionality ν *, core density peaking, separatrix density n e,sep) with variations in the D2 fueling rate and divertor configuration are unified into a single trend when mapped to ⟨T e,ot⟩, the spatially averaged spectroscopically derived outer target electron temperature. Dedicated JET with the ITER-like wall (JET-ILW) experiments in combination with an extended JET-ILW database of unseeded low-triangularity H-mode plasmas spanning a wide range of D2 fueling rates, I p, B t and heating power have demonstrated the importance of ⟨T e,ot⟩ as a key physics parameter linking the recycling particle source and detachment with plasma performance. The remarkably robust H 98(y,2) trend with ⟨T e,ot⟩ is connected to a strong inverse correlation between ⟨T e,ot⟩, n e,sep and ν *, thus directly linking changes in the divertor recycling moderated by ⟨T e,ot⟩ with the previously established relationship between ν *, core density peaking and core pressure resulting in a degradation in core plasma performance with decreasing ⟨T e,ot⟩ (increasing ν *). A strong inverse correlation between the separatrix to pedestal density ratio, n e,sep/n e,ped, and ⟨T e,ot⟩ is also established, with the rise in n e,sep/n e,ped saturating at ⟨T e,ot⟩ > 10 eV. A strong reduction in H 98(y,2) is observed as ⟨T e,ot⟩ is driven from 30 to 10 eV via additional D2 gas fueling, while the divertor remains attached. Consequently, the pronounced performance degradation in attached divertor conditions has implications for impurity seeding radiative divertor scenarios, in which H 98(y,2) is already low (∼0.7) before impurities are injected into the plasma since moderate gas fueling rates are required to promote high divertor neutral pressure. A favorable pedestal pressure, p e,ped, dependence on I p has also been observed, with an overall increase in p e,ped at I p = 3.4 MA as ⟨T e,ot⟩ is driven down from attached to high-recycling divertor conditions. In contrast, p e,ped is reduced with decreasing ⟨T e,ot⟩ in the lower I p branches. Further work is needed to (i) clarify the potential role of edge opacity on the observed favorable pedestal pressure I p scaling; as well as to (ii) project the global and edge plasma performance trends with ⟨T e,ot⟩ to reactor-scale devices to improve predictive capability of the coupling between recycling and confined plasma fueling in what are foreseen to be more opaque edge plasma conditions.

The fusion performance of ELMy H-mode 50–50 deuterium–tritium (DT) plasmas with 50–50 DT NBI injection and q 95 ≈ 3 and β N ≈ 1.8 (also referred to as medium-β N baseline scenario in the rest of this paper) are predicted with the JINTRAC suite of codes and the QuaLiKiZ transport model. The predictions are based on the analysis of plasmas from the first DT campaign on JET in 1997 (DTE1) and pure deuterium plasmas developed at JET in preparation for the DT experimental campaign in 2021 (DTE2), after the installation of a Be/W ITER-like wall in 2011. The sensitivity of the predictions to plasma parameters such as current, toroidal field, pedestal confinement and impurity content are analysed together with the sensitivity to the amount of auxiliary heating power available. The simulations indicate that a fusion power of 10 MW should be achievable under a fairly wide range of assumptions, provided that the auxiliary heating power is around or above 38 MW. Higher fusion power approaching 15 MW could be achievable for this value of β N only for particularly pure plasmas and with 40 MW of additional heating power.

The fusion fuel ion ratio, n T / n D , is an important plasma parameter that needs to be tuned to maximize the power of a tokamak type fusion reactor. It is recognized as a parameter required for optimizing several ITER operating scenarios, and will likely be continuously monitored in future high-performance fusion devices such as DEMO. Tritium was recently introduced in the Joint European Torus (JET) plasma for the first time since the 1997 DTE1 and 2003 TTE campaigns, enabling the possibility to investigate fuel ion ratios. We present a method for measuring n T / n D using neutron time-of-flight (TOF) spectrometry. By fitting the measured neutron spectral features, the relative reaction rate intensities between different ion species can be inferred, from which the fuel ion ratio can be extracted for a corresponding modeled reactivity. Unlike previous measurements of n T / n D using neutron spectrometry, we utilize the neutron energy continuum produced in the three-body TT reaction to determine the fuel ion ratio for plasmas with large concentrations of tritium. Furthermore, the use of neutron TOF spectrometry has never previously been demonstrated for evaluating n T / n D . The method is applied to TOF spectra acquired with TOFOR (JET name KM11) and shown to be consistent with the optical JET diagnostic KT5P which uses optical spectroscopy of a modified Penning gauge plasma to measure tritium and deuterium concentrations in the divertor exhaust gas.

In recent years, a strong effort has been dedicated to the development of tokamak plasma regimes alternative to the standard high confinement mode (H-mode) with type-I edge localized mode (ELM), i.e. ELM-free and small-ELM regimes, given the associated hardly sustainable energy and particle fluxes on plasma facing components. In this work, we will focus on new H-mode regimes with small-ELMs, the so-called baseline small-ELMs (BSE), characterized by high thermal confinement and low core impurity accumulation, which have been recently found at JET. In order to characterize the micro-turbulence at play at the top of the pedestal, an extensive local linear gyrokinetic analysis with the GKW code has been carried out. In particular, a comparison between a reference type-I ELM (#97395) and two BSE plasmas (#96994 and #94442) has been performed. The ion-scale (0.1 ≤ k θ ρ i ≤ 2) micro-turbulence is found to have different characteristics in the two regimes. Indeed, kinetic-ballooning modes (KBM) are destabilized in the type-I ELM regime at k θ ρ i ∼ 0.1, while they are stable in BSE regimes. In addition, negative (i.e. electron-diamagnetic-direction) frequency modes, identified as electron-temperature-gradient (ETG) modes, are destabilized at k θ ρ i ∼ 1.5 in the type-I ELM regime while BSE regimes are characterized by positive (i.e. ion-diamagnetic-direction) frequency modes. Meanwhile, at electron-scale (10 ≤ k θ ρ i ≤ 700) ETG modes are the dominant micro-instabilities in both regimes. Then, since BSE regimes are characterized by a higher impurity concentration at the pedestal, particular attention has been given to the role played by them. We found that impurities represent a critical player in the linear dynamics, strongly affecting the nature of micro-instabilities at ion-scale.

Achieving high-performance conditions and maximizing the fusion yield of plasma discharges have been one of the main goals of recent Joint European Torus (JET) campaigns in preparation for its second deuterium–tritium (D–T) campaign. The simulations shown in this work delve into the role of external heating using neutral beam injection (NBI) and radio-frequency waves in the ion cyclotron range of frequencies (ICRF) in order to optimise high fusion performance in the JET tokamak. A baseline discharge with record neutron rate is used as a reference in order to perform a D–T prediction, which considers the NBI+RF synergy. In this work, our focus is on JET’s two main minority schemes, H and 3He. This study tackles the heating mechanisms by which these schemes reach high-performance conditions. The H scheme typically boosts the ICRF fusion enhancement through the second D harmonic heating, whereas 3He minority is characterised by its strong bulk ion heating. Both features are beneficial for increasing the fusion yield. Nevertheless, the minority concentration is a relevant parameter, which needs to be assessed to understand in which concentration ranges the benefits of each particular minority scheme are met. Therefore, the main objective of this work is to assess in which concentration range the ICRF fusion enhancement and bulk ion heating are optimal, for H and 3He, respectively. Under these conditions, our prediction suggests 3He concentration should stay above 1.2% and H should remain below 2.2%.

The ITER plasma control system (PCS) has successfully completed its final design for the first plasma (FP) and engineering operations (EOs) phase and plant system commissioning has begun as ITER prepares for this first operation phase. Commissioning of the essential plant systems will continue as each plant system is completed and made ready for operation. Tokamak assembly has begun with the base and lower cylinder of the cryostat and the lower most poloidal field (PF) coil installed in the tokamak pit. The first vacuum vessel (VV) sector and accompanying two toroidal field (TF) coils are being prepared for transfer to the pit. Once the tokamak is assembled, the cryostat top lid is closed and pump down begins, this will start approximately one year of integrated commissioning (IC) to prepare all of the relevant plant systems for FP operation. After a scheduled one month of plasma operation with the goal of achieving a plasma current > 100 kA for at least 100 ms, there will be about six months of EO to complete commissioning of the superconducting central solenoid, PF, and TF magnet systems to full current, without plasma, to complete this initial ITER operation phase. The PCS final design for FP will be described as well as the IC sequence of the main plant systems required for this operations phase. The plans for the FP operation campaign will be described, including specific challenges present on ITER due to large VV eddy currents, issues associated with electron cyclotron heating (ECH) assist, neutral pressure and impurities. This will be followed by the EO phase to commission full current operation of the superconducting magnets, possibly including plasma operation at full TF of 5.3 T to have improved conditions both for Ohmic plasma initiation and ECH absorption.

Sustained operation of high-performance, ITER-baseline scenario plasmas at the high levels of input power ( ≲ 40 MW ) required to achieve ∼15 MW of D-T fusion power in JET-ILW requires careful optimisation of the fuelling to avoid an unacceptable disruption rate due to excessive radiation, primarily from W impurities, which are sputtered by edge-localised modes (ELMs) from the divertor targets. By using a train of ELM-pacing pellets from a high-frequency pellet injector to promote regular ELMs, which flush W and other impurities from the confined plasma, such high-performance plasmas can be sustained (for ∼5 s) while maintaining a high normalised confinement factor H 98,y2 ∼ 1, which would otherwise be degraded by reducing the pedestal confinement if a higher rate of D2 gas fuelling were used instead of the pellets to mitigate the W contamination. The causes underlying the improved performance and energy confinement obtained using this combined, gas and pellet fuelling scheme is investigated here in some detail.

Control of a Tokamak requires operating many sophisticated control loops in a dynamic sequence of activities. To take appropriate actions in case technical or physics conditions occur unexpectedly, the continuous control must be backed up by Exception Handling (EH) logic. To mature the Conceptual Design of the ITER Plasma Control System (PCS) with such logic for the Preliminary Design, we studied, in a formal process, how the plasma will be controlled during the 1 st plasma and early operation phases of ITER and analysed the required control and EH functions and dependencies: Three classes of Exception Handling were identified which cover all use cases: modification of the control behaviour of a single control function, modification of the control structure of connected controllers, and change of the control goal which modifies the overall control system. The three classes form a EH hierarchy from low to high impact responses that can be implemented as local EH in the Pulse Continuous Control layer and as central EH in the Pulse Supervision layer of the PCS.

I. Introduction Disruptive events still represent one of the main concerns for the protection of in-vessel components of large size tokamaks, imposing several constraints on the design of the next step experimental devices such as ITER and DEMO. This work aims at summarizing the efforts in the development of an innovative machine learning approach, based on a generative model, towards the implementation of a disruption prediction and avoidance system. To this end, a general-purpose tool based on the Generative Topographic Mapping (GTM) algorithm [1] has been developed [2] and is being upgraded adding new features for a more advanced investigation of the mapped parameter space. GTM performs an unsupervised mapping from a low dimensional latent space, which is usually assumed to be two or three dimensional for visualization purposes, into the high dimensional original data space through radial basis functions, preserving the topology of the data space. This means that operating points close to each other in the data space will be mapped still close in the latent space. The algorithm produces a density model defining probability distributions over the data and the manifold properties, providing at the same time a quantification of the uncertainty of the model fitted to the data. In addition to some global 0-D plasma parameters, where some of them have already been employed for disruption prediction purposes in the past, the original multidimensional space has been described by a set of dimensionless, machine-independent, plasma features. These latter have been synthesized extracting the information associated to 1-D spatial distribution of kinetic quantities and radiated power, which are suitable to describe several physics mechanisms characterizing disruptions and allow a more robust extrapolation to operational

Local beryllium (Be) I and Be II line intensities were measured in the plasma–wall interaction region near an ICRH antenna in JET. The intent was to use these intensities as a measure of the formation of local Radio Frequency (RF) sheath potentials, through RF sheath rectification and potential build up at the end of field lines passing in front of the antenna. Experimentally, it was found that the Be I and Be II emission increase when using the antenna local to the spectroscopic measurement, and increase even more when using a remote antenna that is magnetically connected to the observation point. Magnetic field mapping indicates a magnetic connection between the observation location and the top corner region of the remote antenna and/or its protection limiter. These measurements can be used in support of RF sheath modeling that is an important part of the optimization of antenna design for next generation fusion energy devices, including ITER.

Ion Cyclotron Resonance Heating (ICRH) in the hydrogen minority scheme provides central ion heating and acts favorably on the core tungsten transport. Full wave modeling shows that, at medium power level (4MW), after collisional redistribution, the ratio of power transferred to the electrons in the core (r/a

A specific experiment was performed in Tore Supra, dedicated to the search for consistency between wall inventories estimated from gas balance and post-mortem analysis and to the characterization of the D-retention mechanism. More than 160 similar discharges were performed, representing a cumulative time of 18 000 s of plasma with no intermediate conditioning procedure. The only significant operational issue was linked to the ejection of flakes from the plasma facing components, whose frequency increased dramatically during the campaign, triggering a plasma detachment phase followed by a disruption in number of cases. In-vessel inventory was increased by ∼3.1 × 1024 D, and constant retention rate (∼2.3 × 1020 D/s averaged over discharge duration) was measured, with no indication of wall saturation. First measurements on dismantled tiles of the Toroidal Pump Limiter (TPL) show [D]/[C] ratios ranging from 0.04 to 0.20, on a typical thickness of 15–20 μm. Extrapolated to the whole TPL, this yields an amount of ∼(1.5 ± 0.3) × 1024 D, i.e. ∼50% of the estimated in-vessel inventory.

A prototype of ICRF antenna based on the load-resilient electrical layout anticipated for ITER has been built at CEA-Cadarache. It consists of two toroidally adjacent resonant double loops (RDLs) based on the conjugate-T concept proposed for the ITER ICRF array. This prototype has been recently validated in Tore Supra plasmas exhibiting fast density perturbations in front of the antenna. This paper reports on the load resilience properties of the antenna prototype, as well as the RF modeling and the commissioning. Significant effort on modeling, coupled to an extensive low power campaign, has allowed characterization of the antenna in both vacuum and plasma loading conditions. Plasma load modelings computed with the code TOPICA – very helpful to set up the matching points on plasmas – are found to be in good agreement with the experimental results. The main studies focus on load resilience properties have been carried out in L-mode plasmas. Supersonic molecular beam injection (SMBI), able to launch a series of very short/dense gas jets at Mach number up to 5, was used to mimic the sudden increases of the antenna coupling provoked by ELMs. The results are found to be in good agreement with RF circuit calculations that include 3D modeling.

A set of discharges performed with the JET ITER-like wall is investigated with respect to control capabilities on tungsten sources and transport. In attached divertor regimes, increasing fueling by gas puff results in higher divertor recycling ion flux, lower divertor tungsten source, higher ELM frequency and lower core plasma radiation, dominated by tungsten ions. Both pedestal flushing by ELMs and divertor screening (including redeposition) are possibly responsible. For specific scenarios, kicks in plasma vertical position can be employed to increase the ELM frequency, which results in slightly lower core radiation. The application of ion cyclotron radio frequency heating at the very center of the plasma is efficient to increase the core electron temperature gradient and flatten electron density profile, resulting in a significantly lower central tungsten peaking. Beryllium evaporation in the main chamber did not reduce the local divertor tungsten source whereas core radiation was reduced by approximately 50%.

Since it is uncertain if ITER operation is compatible with type-l edge localized modes (ELMS), the study of alternative ELM regimes is an urgent issue. This paper reports on experiments on JET aiming to find scenarios with small ELMS and good confinement, such as the type-II ELMS in ASDEX Upgrade, the enhanced D-alpha H-mode in Alcator C-mod or the grassy ELMS in JT-60U. The study includes shape variations, especially the closeness to a double-null configuration, variations of q(95), density and beta poloidal. H-mode pedestals without type-I ELMs have been observed only at the lowest currents (

Recently a new recipe for developing and deploying real-time systems has become increasingly adopted in the JET tokamak. Powered by the advent of x86 multi-core technology and the reliability of the JET’s well established Real-Time Data Network (RTDN) to handle all real-time I/O, an official Linux vanilla kernel has been demonstrated to be able to provide realtime performance to user-space applications that are required to meet stringent timing constraints. In particular, a careful rearrangement of the Interrupt ReQuests’ (IRQs) affinities together with the kernel’s CPU isolation mechanism allows to obtain either soft or hard real-time behavior depending on the synchronization mechanism adopted. Finally, the Multithreaded Application Real-Time executor (MARTe) framework is used for building applications particularly optimised for exploring multicore architectures. In the past year, four new systems based on this philosophy have been installed and are now part of the JET’s routine operation. The focus of the present work is on the configuration and interconnection of the ingredients that enable these new systems’ real-time capability and on the impact that JET’s distributed real-time architecture has on system engineering requirements, such as algorithm testing and plant commissioning. Details are given about the common real-time configuration and development path of these systems, followed by a brief description of each system together with results regarding their real-time performance. A cycle time jitter analysis of a user-space MARTe based application synchronising over a network is also presented. The goal is to compare its deterministic performance while running on a vanilla and on a Messaging Real time Grid (MRG) Linux kernel.

Recent progress in experimental and theoretical studies of edge localized mode (ELM) physics is reviewed for the reactor relevant plasma regimes, namely the high confinement regimes, that is, H-modes and advanced scenarios.Theoretical approaches to ELM physics, from a linear ideal magnetohydrodynamic (MHD) stability analysis to non-linear transport models with ELMs are discussed with respect to experimental observations, in particular the fast collapse of pedestal pressure profiles, magnetic measurements and scrape-off layer transport during ELMs.High confinement regimes with different types of ELMs are addressed in this paper in the context of development of operational scenarios for ITER. The key parameters that have been identified at present to reduce the energy losses in Type I ELMs are operation at high density, high edge magnetic shear and high triangularity. However, according to the present experimental scaling for the energy losses in Type I ELMs, the extrapolation of such regimes for ITER leads to unacceptably large heat loads on the divertor target plates exceeding the material limits. High confinement H-mode scenarios at high triangularity and high density with small ELMs (Type II), mixed regimes (Type II and Type I) and combined advanced regimes at high βp are discussed for present-day tokamaks. The optimum combination of high confinement and small MHD activity at the edge in Type II ELM scenarios is of interest to ITER. However, to date, these regimes have been achieved in a rather narrow operational window and far from ITER parameters in terms of collisionality, edge safety factor and βp.The compatibility of the alternative internal transport barrier (ITB) scenario with edge pedestal formation and ELMs is also addressed. Edge physics issues related to the possible combination of small benign ELMs (Type III, Type II ELMs, quiescent double barrier) and high performance ITBs are discussed for present-day experiments (JET, JT-60U, DIII-D) in terms of their relevance for ITER. Successful plasma edge control, at high triangularity (~0.5) and high density (~0.7nGR), in ITB scenarios in JET is reported.Active control of ELMs by edge current, pellet injection, impurities and external magnetic perturbations creating an ergodic zone localized at the separatrix are discussed for present-day experiments and from the perspective of future reactors.

A set of dedicated JET experiments is described where the plasma elongation (k) and triangularity (δ) were varied separately in order to study the influence of plasma magnetic topology on the internal transport barrier (ITB). With low δ, type III ELMs were observed and ITBs readily generated. At the highest δ, large type I ELMs and ELM free phases were observed but, at best, only marginal ITBs. At fixed δ the increase of the elongation of internal magnetic surface have a beneficial effect on the transport quality of the ITB.

In optimized shear plasmas in the Joint European Torus [P. H. Rebut and B. E. Keen, Fusion Technol. 11, 13 (1987)], safety factor (q) profiles with negative magnetic shear are produced by applying lower hybrid (LH) waves during the plasma current ramp-up phase. These plasmas produce a barrier to the electron energy transport. The radius at which the barrier is located increases with the LH wave power. When heated with high power from ion cyclotron resonance heating and neutral beam injection, they can additionally produce transient internal transport barriers (ITBs) seen on the ion temperature, electron density, and toroidal rotation velocity profiles. Due to recent improvements in coupling, q profile control with LH current drive in ITB plasmas with strong combined heating can be explored. These new experiments have led to ITBs sustained for several seconds by the LH wave. Simulations show that the current driven by the LH waves peaks at the ITB location, indicating that it can act in the region of low magnetic shear.

After the change over from the C-wall to the ITER-like Be/W wall (ILW) in JET, the radiation losses during ICRF heating have increased and are now substantially larger than those observed with NBI at the same power levels, in spite of the similar global plasma energies reached with the two heating systems. A comparison of the NBI and ICRF performances in the JET-ILW experiments, based on a statistical analysis of ∼3000 L-mode discharges, will be presented.

Some models for the suppression of turbulence in the L to H transition, suggest that the width of the H-mode edge barrier is either proportional or is of the order of the thermal or the fast-ion poloidal Larmor radius. This would require that the width of the edge barrier should depend on the plasma current. This dependence has been clearly verified at JET in experiments designed to control the edge MHD stability of ELM-free hot-ion H-mode plasmas. The effects of isotopic mass and the applicability of several edge barrier models to the hot-ion H-mode plasmas were analysed in (Guo H Y et al 2000 Edge transport barrier in JET hot-ion H-modes Nucl. Fusion 40 69) using a large database containing both deuterium-only and deuterium-tritium plasmas. This database has now been enlarged to include discharges from a plasma shape scan, allowing one to study the dependence of the pedestal height on the edge shear. In addition, the range of plasma currents was extended up to 6 MA. It is shown that the edge data are best described by a model where the edge barrier width is determined by the fast ions weighted towards the components with largest poloidal Larmor radii. However, it is not possible to conclusively eliminate the thermal ion model.

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.

Recent experimental studies of H-mode plasmas in JET have revealed an n = 0 m = 1 magnetic oscillation starting immediately at the L to H transition (called M-mode for brevity). While the oscillation is present a weak ELM-less H-mode regime is obtained (medium confinement), with a clear increase of pedestal density (ne,ped) and a weak temperature pedestal (Te,ped). In ICRH heated plasmas or low power, low density NBI plasmas the mode and the pedestal pressure can remain steady for the duration of the heating phase, of order 10s or more. The observed axisymmetric MHD oscillation has period ~1­2ms, and odd parity across the mid-plane: m = 1 with sin(q) dependency. Electron Cyclotron Emission, interferometry, reflectometry, SXR and fast Li beam measurements locate the mode in the pedestal region. Da, fast infrared camera and Langmuir probe measurements show that the mode modulates heat and particle flux to the target. Similar phenomena are described elsewhere, lately described as an "I-phase". From a physics point of view, our fundamental observation is that the M-mode is magnetic, not electrostatic, and its frequency scales with pedestal poloidal Alfvén velocity.

After the change over from the C-wall to the ITER-like Be/W wall (ILW) in JET, the radiation losses during ICRF heating have increased and are now substantially larger than those observed with NBI at the same power levels, in spite of the similar global plasma energies reached with the two heating systems. A comparison of the NBI and ICRF performances in the JET-ILW experiments, based on a statistical analysis of ~3000 L-mode discharges, will be presented.