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2021 has culminated with the completion of the JET-ILW DTE2 experimental campaign. This contribution summarizes Ion Cyclotron Resonance Heating (ICRH) operations from system and physics point of view. Improvements to the (ICRH) system, to operation procedures and to real time RF power control were implemented to address specific constraints from tritium and deuterium-tritium operations and increase the system reliability and power availability during D-T pulses. ICRH was operated without the ITER-Like Antenna (ILA) because water leaked from an in-vessel capacitor into the vessel on day-2 of the D-T campaign. Three weeks were required to identify and isolate the leak and resume plasma operations. Dedicated RF-Plasma Wall Interaction (PWI) experiments were conducted; tritium plasmas exhibit a higher level of Be sputtering on the outer wall and impurity content when compared to deuterium or hydrogen plasmas. The JET-DTE2 campaigns provided the opportunity to characterize ICRH schemes foreseen for the ITER operation, in the ITER like wall environment in ELMy H-mode scenarios aiming at maximizing fusion performance. The second harmonic tritium resonance heating and to a lesser extent minority 3He heating (ITER D-T ICRH reference schemes) lead to improved ion temperature and fusion performance when compared to hydrogen minority ICRH. However, these discharges suffered from a lack of stationarity and gradual impurity accumulation potentially because of a deficit of ICRH power when using JET antennas at lower frequencies. Fundamental deuterium ICRH was used in tritium-rich plasmas and with deuterium Neutral Beam Heating; this ICRH scheme proved to be very efficient boosting ion temperature and fusion performance in these plasmas. [ABSTRACT FROM AUTHOR]

Beam-target reactions are responsible for a substantial fraction of the fusion power generated in D-T plasmas in JET-ILW (Be/W-wall), with ion temperatures of 10-12keV and large neutral-beam injection (NBI) power. It is known that injecting D beam ions with energies of ∼100-150keV in T-rich plasmas has a larger potential for beam-target fusion than in 50:50 D:T plasmas, but such a scenario was never developed in the past D-T experiments performed in JET-C (Carbon-wall) and in TFTR in the 90's. On top of the intrinsic advantages of using D beams in T-rich plasmas for D-T neutron production, simulations have shown that fundamental ion-cyclotron resonance heating (ICRH) of the D ions can significantly boost the net fusion reactivity, since both the thermalized D ions and the fast D-NBI ions are accelerated to energy ranges that are optimal for the D-T reaction cross-section. The beneficial effect of fundamental D ICRH on thermal D minorities in tritium plasmas (without NBI) was identified in the JET-C D-T experiments, but was not tested in high performance H-mode discharges with D-NBI heating. In 2021, dedicated JET-ILW DTE2 [1] experiments confirmed - for the first time - the improved fusion performance of T-rich plasmas with high D-NBI power and highlighted the key impact of fundamental D ICRH on the fusion reactivity. This new scenario lead to the world-wide 5s averaged fusion power (and energy) record in D-T tokamak plasmas with dominant beam-target reactions. A brief experimental overview followed by detailed RF wave / Fokker-Planck simulations including NBI-ICRH synergy will be presented, to disentangle the different components contributing to the high neutron yield achieved in these experiments. [ABSTRACT FROM AUTHOR]

This work studies the impact of the Ion Cyclotron Resonance Heating (ICRH) of the fast Neutral Beam Injected (NBI) D and T ions on the fusion performance in recent JET DTE2 campaign. The effect of the Radio Frequency (RF) wave on fast NBI D and T ion dynamics was analyzed by means of TRANSP simulations. Moderate increase, 5-10%, in DT reaction rates was predicted by TRANSP for the case of ICRH heating of fast NBI D ions at n=2 harmonic of ion cyclotron resonance. The simulations have shown that synergistic interaction between fast NBI T ions and RF waves have little or no impact on the fusion performance in the experimental conditions discussed here. Contribution of various heating and fast ion sources have been assessed and discussed. The absence of strong interaction between fast NBI T ions and RF waves was attributed to the poorer penetration of the T beams leading to lower fast T ion densities in the core and the lower perpendicular velocity of the fast NBI T ions resulting in diminished intensity of fast particle – RF wave interaction. [ABSTRACT FROM AUTHOR]

The European Integrated Modelling effort (EU-IM) provides the European Transport Simulator (ETS) [1] which was designed to simulate arbitrary tokamak plasma discharges. Two new 1D Fokker-Planck solvers have recently been implemented within ETS: StixRedist [3] and FoPla [4]. To ensure the CPU time remains acceptable, the latter was parallelized with a generic and easy to implement method. In this paper, it will be shown how these modules were integrated in the ETS workflow in particular a first approach adopted to reach a consistency between wave and Fokker-Planck equation resolution. Also, the Verification and Validation efforts will be discussed. JET shots were analyzed and the ETS predictions were cross-checked against earlier validated codes external to the EU-IM effort, TRANSP [5] in particular, as well as against experimental neutron yield data. A good agreement was obtained, both when comparing the predictions with other codes for cases within their reach (minority or beam populations) and with experimental neutron yield data. Simulations illustrating the exploitation of the nonlinear collision operator when solving a set of coupled Fokker-Planck equations for cases when majority species play a key role will be also shown. [ABSTRACT FROM AUTHOR]

Ion cyclotron resonance heating (ICRH) plays an important role in high performance JET-ILW plasma discharges, both for keeping the heavy impurities away from the plasma centre as for increasing the core ion temperature to boost fusion performance. While the former is needed in all high-performance discharges for steady state operation, the latter will be particularly important in the next-coming JET Deuterium-Tritium campaign (DTE2). Currently, the workhorse for impurity control in high power D plasmas is fundamental H minority ICRH (with simultaneous ω=2ωc harmonic D heating), which leads to localized core electron heating that induces turbulence (flatter density profiles) as well as peaked electron temperatures. For fusion power enhancement, dominant bulk ion heating and RF acceleration of the NBI ions to appropriate energies would be preferable and theoretical predictions suggest that ion heating is also effective for core impurity screening. In this paper, we discuss the basic modeling results of different ICRF scenarios available for the DTE2 campaign in JET-ILW, highlighting their main properties in terms of the RF absorption of the various species, their slowing-down properties and their impact on high-Z impurity transport. Correctly modeling the wave absorption, slowing-down and collisional energy redistribution of the simultaneously RF-heated species in a DT plasma mix with important neutral-beam injection (NBI) is numerically challenging and is outside the scope of this paper. The simplified calculations presented here are rather intended to give the reader an overview of the ICRH options for JET-DTE2 with references to the state-of-the-art ICRH modeling given throughout the paper. [ABSTRACT FROM AUTHOR]

The paper discusses the characterization of the three high power steady-state and load-resilient ICRF launchers of WEST before their installation in the tokamak. These launchers have been characterized and validated in low-power experiments (milliwatt range) as well as in experiments at the nominal RF voltages and currents in the TITAN vacuum chamber (∼30 kV and 915 A peak). The successful commissioning of two of the launchers during the WEST C3 campaign at ∼1 MW power level is illustrated. Manual and real-time controlled impedance-matching of the launchers are discussed, as well as the validation of their load-resilience. Furthermore, several redundant and complementary protection systems have been validated and are reviewed in the paper. [ABSTRACT FROM AUTHOR]

The effect of the novel 'three-ion' D-(3He)-H minority ICRH heating scheme on the behavior of the metallic impurities at JET-ILW is discussed. The reported experiment was performed in L-mode plasmas at a magnetic field BT = 3.2 T, plasma current Ip = 2 MA and central plasma densities ne(0)≈4×1019m-3. ICRH power was delivered with dipole or +π/2 antenna phasing at f ≈ 32.2-33MHz, placing the 3He cyclotron resonance at the plasma core. The edge isotopic ratio H/(H+D) was varied between 73 and 92%, and 3He concentration in the range of 0.1-1.5% to assess the sensitivity of the scheme to the detailed plasma composition. The results of our analysis show a linear increase of the plasma effective charge Zeff, radiated power Prad, bulk and content of metallic impurities with ICRF power. The observed scattering of the points reflects the difference in the plasma composition and ICRF antenna phasing. For discharges heated with similar ICRH power level ∼4MW, our analysis indicates that for a large range of H/(H+D) the novel scenario effectively heats the plasma with reduced content of metallic impurities. The impurities are shown to be concentrated mainly around the mid-radius region of the plasma. We conclude this paper with a discussion of the effect of the long-period sawteeth on the observed dynamics of metallic impurities in the plasma core. [ABSTRACT FROM AUTHOR]

A tritium (T) campaign is planned in preparation for the deuterium-tritium (D-T) campaign at the Joint European Torus (JET). These experiments will be the first experiments involving T with the ITER-like plasma-wall facing components materials. They will give a unique opportunity to test one of the most promising ion cyclotron resonance frequency (ICRF) heating schemes for ITER plasmas: the 2nd tritium (T) harmonic resonance (ω = 2ωT). This paper provides two key contributions related to modelling of the performance of this scheme at JET. First, we assess the heating performance of the 2nd T harmonic resonance and, second, we model different ICRF schemes for the T campaign in support for the D-T campaign, i.e., identify differences and similarities from the heating point of view between T and D-T plasmas in order to predict the performance of ω = 2ωT in the D-T scenario. In our modelling we use a selected hybrid record discharge as reference, i.e., using its experimental profiles. We consider two ICRF schemes (6 MW), i.e. ω = ω3He = 2ωT (no 3He), with a central resonance and three NBI power outputs (15, 25 and 35 MW) in two plasma compositions (100% T and 50%:50% D-T). Note that isotope effects are not taken into account in these simulations. For this study, the ICRF and NBI heating are modelled with the ICRF code PION and the beam code PENCIL which take into account the ICRF+NBI synergy. The analysis of the T velocity distribution function shows that a stronger tail is formed in those plasmas with lower tritium density. This fact has an important impact on the slowing-down process of fast tritons with the background species. The T plasma shows a higher and more peaked ion-ion collisional power density at the plasma centre as compared to D-T plasma. In the T plasma, ICRF heating drives fast tritons at the plasma centre with an average energy substantially lower than in the other case. On the other hand, the use of 3He as a minority makes the fast ion T energy considerably lower due to strong 3He absorption. Fast ion average energies reached at the plasma centre are similar in all species mixture cases. As a result, there is a strong heating similarity between T and D-T. However, it is crucial to study the generation of a strong T tail as a result of particle-wave interaction which is not possible under this scheme and needs to be studied in plasmas without 3He in preparation of JET D-T campaign and ITER. Therefore, both schemes (ω = ω3He = 2ωT and ω=2ωT) need to be tested. [ABSTRACT FROM AUTHOR]

In 2018, Ion Cyclotron Resonant Frequency (ICRF) waves were for the first time applied to the WEST plasma scenarios. In ICRF-only plasmas at medium density, or on top of a low level of Lower Hybrid (LH) power, the coupled ICRF power increases the plasma energy content. However in discharges with large LH power at high core density, nearly all the applied ICRF power gets radiated, mainly in the plasma bulk. Both the energy content and conducted power decrease. Two peculiarities of WEST, that may combine, are presently invoked to explain this phenomenology:1) Fast ion ripple losses, evidenced between TF coils on the baffle, likely degrade the ICRF heating efficiency. 2) RF-enhanced W sources, evidenced on several Plasma-Facing Components (PFCs), likely over-contaminate the high-power plasmas. [ABSTRACT FROM AUTHOR]

The studies of recent JET experiments in H/D≈0.85/0.15 plasma (2.9T/2MA) in which neutron rate was enhanced by applying 2.5MW of ICRH using D-(DNBI)-H three-ion scheme are reported. An extensive analysis of this novel heating scenario has been carried out by means of integrated TRANSP/TORIC modelling, and a comprehensive validation of the computed Fast Ion Distribution Function (FI DF) with a range of fast ion diagnostics available at JET is presented. The predicted acceleration of D Neutral Beam Injection (NBI) ions beyond their injection energies and the associated changes in FI DF by RF waves are found to be in good agreement with measured neutron yield and TOFOR neutron spectrometer measurements, as well as with multi-channel neutron camera observations and neutral particle analyser diagnostic. An outlook of the possible applications of the developed technique for future DTE2 studies on JET has been highlighted. Controlled acceleration of TNBI ions in D-rich and DNBI ions in T-rich plasmas to optimal energies can be applied to maximise BT fusion rates and contribute to the success of future DT experiments at JET and ITER as illustrated in this study. [ABSTRACT FROM AUTHOR]

This study compares experimental observations on the two challenges of Ion Cyclotron Resonance Heating (ICRH), namely the coupling of waves to the plasma and the enhanced impurity sputtering in EAST and WEST medium size tokamaks. In WEST, Lower Hybrid (LH) power helps improving ICRH coupling. In both machines experiments reveal that fueling from the midplane not only helps to couple waves from nearby antennas like in other devices, but also has an impact on the scrape-off layer (SOL) density in regions that are not magnetically connected to the valves. Localized midplane nozzle valves allow similar (in WEST) or better (in EAST) coupling compared to poloidally distributed valves. Core density control requirements for long-pulse operation, in particular in L-mode regime, however limit the amount of gas that can be injected. During ICRH impurities can contaminate the plasma up to a level detrimental for the operation, e.g. up to 100% of ICRH power can be radiated on WEST in certain high power conditions. In WEST, tungsten (W) production measured by visible spectroscopy increases on all the observed objects during ICRH compared to a reference phase without ICRH. On some components (antenna side limiters, baffle, divertor) the rise is larger than with a similar LH power. The relative contribution of each object and physical process (RF-sheaths, fast ion ripple losses) to core contamination yet remains poorly known. Comparing antenna limiters with W-coating vs low-Z materials would help quantifying the role of these components. In EAST, the core W content, measured by EUV spectroscopy in presence of divertor sources only, is correlated with the total injected power, either from ICRH or LH. Since 2018 the LH guard limiter tiles were W-coated. Their contribution to the core W content appears more important than divertor sources when two-strap ICRH antennas magnetically connected to W components at the midplane is powered, compromising high performance operations. [ABSTRACT FROM AUTHOR]

Heating single ion species plasmas with ICRF is a challenging task: Fundamental ion cyclotron heating (w = wci) suffers from the adverse polarization of the RF electric fields near the majority cyclotron resonance while second harmonic heating (w = 2wci) typically requires pre-heating of the plasma ions to become efficient. Recently, w = 2wci ICRF heating was tested in JET-ILW hydrogen plasmas in the absence of neutral beam injection (L-mode). Despite the lack of pre-heating, up to 6MW of ICRF power were coupled to the plasma leading to a transition to H-mode for PICRH>5MW in most discharges. Heating efficiencies between 0.65-0.85 were achieved as a combination of the low magnetic field adopted (enhanced finite Larmor radius effects) and the deliberate slow rise of the ICRF power, allowing time for a fast ion population to gradually build-up leading to a systematic increase of the wave absorptivity. Although fast ion tails are a common feature of harmonic ICRF heating, the N=2 majority heating features moderate tail energies (<500keV) except at very low plasma densities (ne0<3x1019/m³), where fast H tails in the MeV range developed and fast ion losses became significant, leading to enhanced plasma wall interaction. The main results of these experiments will be reported. [ABSTRACT FROM AUTHOR]

Ion cyclotron resonance frequency (ICRF) heating has been an essential component in the development of high power H-mode scenarios in JET-ILW. The steps that were taken for the successful use of ICRF heating in terms of enhancing the power capabilities and optimizing the heating performance in view of core impurity mitigation in these experiments will be reviewed. [ABSTRACT FROM AUTHOR]

As follow-up of the benchmark activity of ICRF full-wave codes within the EUROfusion Code Development for Integrated Modelling project (WPCD), a simple-to-complex approach has been devised for verification of the European ICRF codes, imported in the European-Integrated Modelling infrastructure, which represents a unique environment for input-data sharing and result analysis. This benchmark activity has been recently extended to non-European codes, in particular the ICRF full-wave AORSA code. Here we discussed the results of this benchmark. [ABSTRACT FROM AUTHOR]

The Joint European Torus (JET, Culham, UK) is the largest tokamak in the world. It is devoted to nuclear fusion experiments of magnetic confined Deuterium (D) or Deuterium-Tritium (DT) plasmas. JET has been upgraded over the years and recently it has also become a test facility of the components designed for ITER, the next step fusion machine under construction in Cadarache (France). JET makes use of many different diagnostics to measure the physical quantities of interest in plasma experiments. Concerning D or DT plasmas neutron production, various types of detectors are implemented to provide information upon the neutron total yield, emission profile and energy spectrum. The neutron emission profile emitted from the JET plasma poloidal section is reconstructed using the neutron camera (KN3). In 2010 KN3 was equipped with a new digital data acquisition system capable of high rate neutron measurements (<0.5 MCps). A similar instrument will be implemented on ITER and it is currently in its design phase. Various types of neutron spectrometers with different view lines are also operational on JET. One of them is a new compact spectrometer (KM12) based on organic liquid scintillating material which was installed in 2010 and implements a similar digital data acquisition system as for KN3. This article illustrates the measurement results of KN3 neutron emission profiles and KM 12 neutron energy spectra from the latest JET D experimental campaign C31. [ABSTRACT FROM AUTHOR]

The ability to non-inductively drive a large fraction of the toroidal plasma current in magnetically confined plasmas is an essential requirement for steady state fusion reactors such as DEMO. Besides neutral beam injection (NBI), electron-cyclotron resonance heating (ECRH) and lower hybrid wave heating (LH), ion-cyclotron resonance heating (ICRH) is a promising candidate to drive current, in particular at the high temperatures expected in fusion plasmas. In this paper, the current drive (CD) efficiencies calculated with coupled ICRF wave / CD numerical codes for the DEMO-1 design case (R0=9m, B0=6.8T, ap=2.25m) [1] are presented. It will be shown that although promising CD efficiencies can be obtained in the usual ICRF frequency domain (20-100MHz) by shifting the dominant ion-cyclotron absorption layers to the high-field side, operation at higher frequencies (100-300MHz) has a stronger CD potential, provided the parasitic RF power absorption of the alpha particles can be minimized. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

The most recent JET campaign has focused on characterizing operation with the "ITER-like" wall. One of the questions that needed to be answered is whether the auxiliary heating methods do not lead to unacceptably high levels of impurity influx, preventing fusion-relevant operation. In view of its high single pass absorption, hydrogen minority fundamental cyclotron heating in a deuterium plasma was chosen as the reference wave heating scheme in the ion cyclotron domain of frequencies. The present paper discusses the plasma behavior as a function of the minority concentration X[H] in L-mode with up to 4MW of RF power. It was found that the tungsten concentration decreases by a factor of 4 when the minority concentration is increased from X[H] ≈ 5% to X[H] ≈ 20% and that it remains at a similar level when X[H] is further increased to 30%; a monotonic decrease in Beryllium emission is simultaneously observed. The radiated power drops by a factor of 2 and reaches a minimum at X[H] ≈ 20%. It is discussed that poor single pass absorption at too high minority concentrations ultimately tailors the avoidance of the RF induced impurity influx. The edge density being different for different minority concentrations, it is argued that the impact ICRH has on the fate of heavy ions is not only a result of core (wave and transport) physics but also of edge dynamics and fueling. [ABSTRACT FROM AUTHOR]

ICRF (ion cyclotron range of frequencies) heating experiments were performed in D-H plasmas at various H concentrations on ASDEX Upgrade. The rf power was modulated to measure the electron power deposition profile from electron temperature modulation. To minimize the contribution from indirect collisional heating and the effect of radial transport, the rf power was modulated at 50 Hz. However, peaking of electron temperature modulation was still observed around the hydrogen cyclotron resonance indicating collisional heating contribution. Time dependent simulation of the hydrogen distribution function was performed for the discharges, using the full-wave code AORSA (E.F. Jaeger, et al., Phys. Plasmas, Vol. 8, page 1573 (2001)) coupled to the Fokker-Planck code CQL3D (R.W. Harvey, et al., Proc. IAEA (1992)). In the present experimental conditions, it was found that modulation of the collisional heating was comparable to that of direct wave damping. Impact of radial transport was also analyzed and found to appreciably smear out the modulation profile and reduce the phase delay. [ABSTRACT FROM AUTHOR]

The grand aim of the Integrated Tokamak Modelling (ITM) task-force is to provide a ..exible, modular and reliable plasma simulator in view of planning and analyzing ITER discharges. Since radio-frequency (rf) heating in the ion cyclotron range of frequencies (ICRF) is foreseen as one of the main additional heating systems in ITER, physics modules that simulate ICRF wave propagation and absorption are necessary for the ITM project. Here, we report on the status of the benchmark activity of ICRF codes, already imported in ITM environment platform. We consider various scenarios for ITER, limiting the comparisons to wave propagation and absorption in Maxwellian plasmas. [ABSTRACT FROM AUTHOR]

Magnetically confined plasmas, such as those produced in the tokamak JET, contain measurable amounts of impurity ions produced during plasma-wall interactions (PWI) from the plasma-facing components and recessed wall areas. The impurities, including high- and mid-Z elements such as tungsten (W) from first wall tiles and nickel (Ni) from Inconel structure material, need to be controlled within tolerable limits, to ensure they do not significantly affect the performance of the plasma. This contribution focuses on documenting W and Ni impurity behavior during Ion Cyclotron Resonance Heating (ICRH) operation with the new ITER-Like Wall (ILW). Ni- and W-concentration were derived from VUV spectroscopy and the impact of applied power level, relative phasing of the antenna straps, plasma separatrix - antenna strap distance, IC resonance position, edge density and different plasma configuration, on the impurity release during ICRH are presented. For the same ICRH power the Ni and W concentration was lower with dipole phasing than in the case of -π/2 phasing. The Ni concentration was found to increase with ICRH power and for the same NBI power level, ICRH-heated plasmas were characterized by two times higher Ni impurity content. Both W and Ni concentrations increased strongly with decreasing edge density which is equivalent to higher edge electron temperatures and more energetic ions responsible for the sputtering. In either case higher levels were found in ICRH than in NBI heated discharges. When the central plasma temperature was similar, ICRH on-axis heating resulted in higher core Ni impurity concentration in comparison to off-axis ICRH in L-mode. It was also found that the main core radiation during ICRH came from W. [ABSTRACT FROM AUTHOR]

During the initial operation of the JET ITER-like wall, particular attention was given to the characterization of the Ion Cyclotron Resonance Frequency (ICRF) heating in this new metallic environment. In this contribution we compare L-modes plasmas heated by ICRF or by Neutral Beam Injection (NBI). ICRF heating as expected led to a much higher centrally peaked power deposition on the electrons and due to the central fast ion population to stronger sawtooth activity. Surprisingly, although a higher bulk radiation was observed during the ICRF phase, the thermal plasma energy was found similar for both cases, showing that a higher radiation inside the separatrix was not incompatible with an efficient central heating scheme. The higher radiation was attributed to the presence Tungsten (W). Tomographic inversion of SXR emissions allowed a precise observation of the sawtooth effect on the radiation pattern. W concentration profiles deconvolved from SXR emission showed the flattening of the profiles due to sawtooth for both heating and the peaking of the profiles in the NBI case only hinting for extra transport effect in the ICRF case. [ABSTRACT FROM AUTHOR]

The aim of driving a sufficient amount of plasma current with an appropriate radial current density profile is considered as one of the key challenges for a tokamak fusion power plant in steady state operation. Furthermore, efficient heating to enable transition to regime of enhanced confinement and to achieve breakeven plasma temperatures as well as MHD control and plasma breakdown assistance are required. In the framework of the EFDA Power Plant Physics and Technology (PPPT) activities, the ability of the Electron cyclotron (EC), Ion Cyclotron (IC) and Lower Hybrid (LH) systems to fulfil these requirements, was studied for a demonstration fusion power plant (DEMO). As boundary condition, a 1D description of the plasma for a pulsed DEMO based on system code studies combined with transport analysis was developed. The predicted 1D plasma parameters were used to calculate the current drive (CD) efficiency of each system and eventually optimised it. As an example, the EC current drive efficiency could be increased strongly by top launch compared to equatorial launch at least by a factor of two. For the IC system, two possible windows of operation for standard and higher frequencies were highlighted, whereby again top launch leads to higher CDefficiencies. The efficiencies predicted for DEMO for the RF current drive systems will be presented. Finally, gaps in the feasibility of RF systems under DEMO relevant conditions will be identified. [ABSTRACT FROM AUTHOR]

In 2011/12, JET started operation with its new ITER-Like Wall (ILW) made of a tungsten (W) divertor and a beryllium (Be) main chamber wall. The impact of the new wall material on the JET Ion Cyclotron Resonance Frequency (ICRF) operation was assessed and also the properties of JET plasmas heated with ICRF were studied. No substantial change of the antenna coupling resistance was observed with the ILW as compared with the carbon wall. Heat-fluxes on the protecting limiters close the antennas quantified using Infra-Red (IR) thermography (maximum 4.5 MW/m² in current drive phasing) are within the wall power load handling capabilities. A simple RF sheath rectification model using the antenna near-fields calculated with the TOPICA code can well reproduce the heat-flux pattern around the antennas. ICRF heating results in larger tungsten and nickel (Ni) contents in the plasma and in a larger core radiation when compared to Neutral Beam Injection (NBI) heating. Some experimental facts indicate that main-chamber W components could be an important impurity source: the divertor W influx deduced from spectroscopy is comparable when using RF or NBI at same power and comparable divertor conditions; the W content is also increased in ICRF-heated limiter plasmas; and Be evaporation in the main chamber results in a strong and long lasting reduction of the impurity level. The ICRF specific high-Z impurity content decreased when operating at higher plasma density and when increasing the hydrogen concentration from 5% to 20%. Despite the higher plasma bulk radiation, ICRF exhibited overall good plasma heating efficiency; The ICRF power can be deposited at plasma centre and the radiation is mainly from the outer part of the plasma. Application of ICRF heating in H-mode plasmas started, and the beneficial effect of ICRF central electron heating to prevent W accumulation in the plasma core could be observed. [ABSTRACT FROM AUTHOR]

The non-active operation phase of ITER will be done in H and 4He plasmas at half the nominal magnetic field, B0 = 2.65T. At this field and for the given frequency range of the ICRF system (f = 40-55MHz), three ICRF heating scenarios are available a priori: (i) Fundamental ICRH of majority H plasmas at f≈40MHz, (ii) second harmonic (N= 2) 3He ICRH in H plasmas at f≈53MHz and (iii) fundamental minority H heating in 4He plasmas at f≈40MHz. While the latter is expected to perform well for not too large H concentrations, the heating scenarios available for the Hydrogen plasmas are less robust. Recent JET experiments performed in similar conditions to those expected in ITER's half-field phase confirmed the low performance of these two scenarios and numerical simulations have shown that the situation is not much improved in ITER, mainly because of the rather modest plasma temperature and density expected in its initial operation phase. A summary of the main experimental results obtained at JET followed by numerical predictions for ITER's half-field ICRF heating scenarios will be presented. [ABSTRACT FROM AUTHOR]

The constructive interference effect described by Fuchs et al. [1] shows that the mode conversion and thereby the overall heating efficiency can be enhanced significantly when an integer number of fast wave wavelengths can be folded in between the high field side fast wave cutoff and the ion-ion hybrid layer(s) at which the ion Bernstein or ion cyclotron waves are excited. This effect was already experimentally identified in (3He)-D plasmas [2] and was recently tested in (3He)-H JET plasmas. The latter is an 'inverted' scenario, which differs significantly from the (3He)-D scenarios since the mode-conversion layer is positioned between the low field side edge of the plasma and the ion-cyclotron layer of the minority 3He ions (whereas the order in which a wave entering the plasma from the low field side encounters these layers is inverted in a 'regular' scenario), and because much lower 3He concentrations are needed to achieve the mode-conversion heating regime. The presence of small amounts of 4He and D in the discharges gave rise to an additional mode conversion layer on top of the expected one associated with 3He-H, which made the interpretation of the results more complex but also more interesting: Three different regimes could be distinguished as a function of X[3He], and the differing dynamics at the various concentrations could be traced back to the presence of these two mode conversion layers and their associated fast wave cutoffs. Whereas (1-D and 2-D) numerical modeling yields quantitative information on the RF absorptivity, recent analytical work by Kazakov [3] permits to grasp the dominant underlying wave interaction physics. [ABSTRACT FROM AUTHOR]

Heat loads from ICRF and LH wave absorption in the SOL are characterized on JET from the de-convolution of surface temperatures measured by infrared thermography. The spatial localization, quantitative estimates, parametric dependence and physical origin of the observed heat fluxes are documented. Implications of these observations are discussed for the operation of JET with an ITER-Like Wall, featuring Beryllium tiles with reduced power handling capability. [ABSTRACT FROM AUTHOR]

The first series of experiments with the ITER-like wall (ILW) will start mid-2011 with D plasmas and will continue through 2012-13 with H, 4He and D plasmas, and up to 2014-15, when a DT campaign is proposed. In this paper, the previous experience at JET is reviewed to set the scene for the future challenges of ICRF operation including change in the ICRF coupling, W impurity production and evaluation of localized power loads due the RF sheaths. development in a Beryllium/Tungsten environment of ICRF heating schemes for the non activated and the DT phases of ITER. [ABSTRACT FROM AUTHOR]

The current planning of IC systems in W7-X foresees in a first phase the construction of a single strap IC wall conditioning (ICWC) antenna, fed by one generator providing max 1MW at frequencies between 4 and 26 MHz. While this system is expected to be well suited for ICWC applications, we show that it could also be used for ICRF plasma heating and current drive (H&CD) upon making modest modifications to the original design and adding a second generator. A two strap antenna design is proposed (i) with at least the same performance for wall conditioning as the original design; (ii) allowing flexible relative strap phasings for ICRF heating, fast particle generation and/or current drive; (iii) using the full ICRH power from the two generators available; (iv) with a compact design, such that it can be embedded in the wall panels, causing minimal perturbations for various magnetic configurations. The use of such an antenna would allow exploring experimentally various H&CD scenarios, and generate the necessary data sets for validating numerical tools. Experimental tests of various H&CD scenarios now, would thus offer important advantages for the design of a dedicated H&CD antenna later. [ABSTRACT FROM AUTHOR]

A set of simulation/interpretation tools based on transmission line theory and on the RF model developed by M. Vrancken [2] has been developed to study the ITER-like Antenna (ILA) at JET. For given tuning element settings, the unique solution of the equations governing the ILA circuit requires solving a system of coupled linear equations relating the voltages and currents at the antenna straps and other key locations. This computation allows cross-checking predicted values against measured experimental ones. Further more, a minimization procedure allows improving the correspondence with the quantities measured in the circuit during shots, thus coping with unavoidable errors arising from uncertainties in the measurements or from inaccuracies in the adopted RF model. Typical applications are e.g. fine-tuning of the second-stage of the ILA circuit for increased ELM-resilience, cross-checking the calibration of the measurements throughout the circuit and predicting the antenna performance and matching conditions in new plasma scenarios. [ABSTRACT FROM AUTHOR]

The Scattering Matrix Arc Detection System (SMAD) has been fully deployed on all 4 sets of Resonant Double Loop (RDL), Vacuum Transmission Line (VTL) and Antenna Pressurised Transmission Lines (APTL) of the JET ICRF ITER-Like Antenna (ILA) and this has been indispensable for operating at low (real) T-point impedance values to investigate ELM tolerance. This paper describes the necessity of the SMAD vs VSWR (Voltage Standing Wave Ratio) protection system, SMAD commissioning, problems and a number of typical events detected by the SMAD system during operation on plasma. [ABSTRACT FROM AUTHOR]

Using the JET A2 ICRF antennas [1], experiments were carried out to assess the performance of three different dipole phasing configurations relevant for the operation of the ICRF antenna in ITER. Three similar discharges with dipole (0π0π), “symmetric dipole” (0ππ0) and “super dipole” (00ππ) phasings were compared. In the “super dipole” case, higher coupling was confirmed but lower heating efficiency and much stronger plasma wall interaction were observed, as corroborated, respectively, by the analysis of the diamagnetic energy response to the ICRF power steps and by the observation of a considerable temperature rise of the antenna limiters and septa in the 00ππ case. These observations were found to be in line with simulations of the ICRF wave absorption and with High Frequency Structure Simulator (HFSS) modelling of the RF fields near the antenna [2]. [ABSTRACT FROM AUTHOR]

W sputtering during ICRF on ASDEX Upgrade (AUG) and temperature rise on JET A2 antenna septa are considered in connection with plasma conditions at the antenna plasma facing components and E| near-fields. Large antenna-plasma clearance, high gas puff and low light impurity content are favorable to reduce W sputtering in AUG. The spatial distribution of spectroscopically measured effective W sputtering yields clearly points to the existence of strong E| fields at the antenna box (“feeder fields”) which dominate over the fields in front of the antenna straps. The picture of E| fields, obtained by HFSS code, corroborates the dominant role of E| at the antenna box on the formation of sheath-driving RF voltages for AUG. Large antenna-plasma clearance and low gas puff are favorable to reduce septum temperature of JET A2 antennas. Assuming a linear relation between the septum temperature and the sheath driving RF voltage calculated by HFSS, the changes of the temperature with dipole phasing (00ππ, 0ππ0 or 0π0π) are well described by the related changes of the RF voltages. Similarly to the AUG antenna, the strongest E| are found at the limiters of the JET A2 antenna for all used dipole phasings and at the septum for the phasings different from 0π0π. A simple general rule can be used to minimize E| at the antenna: image currents can be allowed only at the surfaces which do not intersect magnetic field lines at large angles of incidence. Possible antenna modifications generally rely either on a reduction of the image currents, on their short-circuiting by introducing additional conducting surfaces or on imposing the E| = 0 boundary condition. On the example of AUG antenna, possible options to minimize the sheath driving voltages are presented. [ABSTRACT FROM AUTHOR]

The commissioning of the ITER-Like ICRF Antenna (ILA) [1] on JET plasmas from May 2008 to April 2009 in various conditions (33, 42 and 47 MHz, L- and H-mode, antenna strap-plasma separatrix distances of ∼9 to 17 cm) has provided relevant information for future antenna design and operation. The maximum power density achieved was 6.2 MW/m2 in L-mode with strap to plasma separatrix distance of ∼9–10 cm at 42 MHz on the lower half of the ILA extrapolating to 8 MW/m2 if the full generator power had been available. Efficient (trip-free operation) ELM tolerance was obtained both at 33 and 42 MHz on a large range of ELMs with strap voltages up to 42 kV and a maximum power density of 4.1 MW/m2. The paper reviews these achievements as well as remaining issues. [ABSTRACT FROM AUTHOR]

Ion Cyclotron Resonance Frequency (ICRF) antennas achieve maximum power transfer from RF generator to plasma load by establishing an impedance match. The JET ITER-Like Antenna (ILA) consists of 4 mutually coupled resonators that need to be matched simultaneously and whose resonant states need to be accurately controlled by several available actuators to couple maximum power and to achieve optimal ELM resilience. The operation of the matching system is described. Experimental operation revealed some phenomena and sensitivities that did not surface from simulation alone and that should be taken into account for the design of future ICRF antennae systems. [ABSTRACT FROM AUTHOR]

Most of today's tokamak experiments rely on auxiliary heating of the plasma for achieving reactor-relevant temperatures. On top of the theoretical efforts to estimate the power deposition profiles of the various adopted heating schemes (NBI, ICRH or ECRH) by numerical modelling, there are various techniques used to infer the power profiles experimentally, usually based on electron/ion temperature data. Amongst these methods, the break-in-slope (BIS) and the Fast Fourier Transform (FFT) analyses of the experimental temperature signals responding to a discontinuity (or modulation) on the applied power have been systematically used in JET in the last years. The FFT method requires a periodic modulation of the auxiliary power and provides information on both the power deposition profile and the diffusive heat transport. The standard BIS (linear fit), a priori, does not distinguish between local power deposition and heat transport, but has the advantage of needing only a single discontinuity on the auxiliary power level. In this work we present an improved BIS procedure that not only considers the time delays between the RF power change and the respective temperature responses at the several plasma radii, but also extends the original linear fit of the experimental signals to an exponential one, in a tentative to capture part of the transport phenomena. The results of the improved BIS analysis are compared with the FFT results and the main advantages/restrictions of the two methods are discussed. [ABSTRACT FROM AUTHOR]

Recent experiments in JET have provided information on the potential of using majority RF heating schemes in large plasmas. Adopting a wide range of available diagnostics, the plasma behaviour was monitored. The main results of the experiments are that—due to the poor antenna coupling at low frequency, the low (Ohmic) plasma temperature and the reduced RF electric field amplitude near the ion-cyclotron resonance layer of the majority ions—ICRH alone is barely capable of heating the plasma. On the other hand, when preheating the plasma using neutral beam injection, the wave-plasma coupling is noticeably improved and considerable plasma heating, followed by increased neutron yield were observed in several diagnostics. This effect is not only attributed to the lower collisionality of the pre-heated plasma but also to the Doppler-shifted IC absorption of the fast beam ions. By studying the response of the plasma to sudden changes in the RF power level, the experimental power deposition profiles were determined and compared to theoretical predictions. The numerical modelling was done adopting a coupled wave/Fokker-Planck code that enables accounting for the non-Maxwellian distributions of the RF heated particles and the injected beam ions in the wave equation, and for the actual local RF fields in the Fokker-Planck description. The theoretical results confirm the experimental finding that the beam ions do play a crucial role in this heating scheme. [ABSTRACT FROM AUTHOR]

D plasmas with 3He minorities have sharp, thin ion-ion hybrid layers that enable to efficiently excite short wavelength branches that are subsequently damped by fairly well localized electron Landau and TTMP absorption. Depending on the minority concentration chosen, ion minority heating or electron mode conversion damping is dominant. Recent experiments have been devoted to the study of (3He)-D JET plasmas. One aspect of those experiments—using RF heating as a tool—is the study of the response of the plasma to RF power modulation, allowing to examine the fate of the RF power and to diagnose particle and energy transport. The present paper gives a very brief summary of a subset of these experiments. The focus will largely but not exclusively be on understanding ITB physics. The adopted probing methods are more generally applicable, though. [ABSTRACT FROM AUTHOR]

Recent progress on heating and current drive on JET is reported. Topics discussed are: high power coupling of ICRF/LH at ITER relevant antenna/launcher-separatrix distances, succesfull demonstration of 3 dB couplers for ELM tolerance of the ICRF system, influence of ICRF on LH operation, rotation studies in plasma without external momentum with standard and enhanced JET toriodal field ripple, studies of different ICRF heating schemes and of NTM avoidance schemes using Ion Cyclotron Current Drive. A brief outlook on future plans for experiments at JET is given. [ABSTRACT FROM AUTHOR]

Results of the experimental studies of ICRH at the fundamental cyclotron frequency of the majority deuterons in JET plasmas with near-tangential deuteron neutral beam injection (NBI) are presented. 1D, 2D and 3D ICRH modeling indicated that several ITER relevant mechanisms of heating may occur simultaneously in this heating scheme: fundamental ion cyclotron resonance heating of majority and beam D ions, impurity ion heating and electron heating due to Landau damping and TTMP. These mechanisms were studied in JET experiments with a ∼90% D, 5% H plasma including traces of Be and Ar. Up to 2MW of ICRH power was applied at 25 MHz to NBI heated plasmas. In most of the discharges the toroidal magnetic field strength was 3.3T, but in one it was equal to 3.6T. The E+ component of the electric field governs the ion cyclotron heating of not too fast particles. The Doppler shifted RF absorption of the beam deuterons away from the cold resonance at which E+ is small was exploited to enhance the RF power absorption efficiency. Fundamental ICRH experiments were also carried out in LHD hydrogen plasma with high energy hydrogen NBI. ICRH was performed at 38MHz with injected power <1 MW. The effect of fundamental ICRH was clearly demonstrated in both machines. [ABSTRACT FROM AUTHOR]

In the initial phase of ITER H plasmas will be used in order to avoid activating the machine. The reference ICRF healing scenarios rely on minority species such as Helium (3He) or deuterium (D). These schemes’ distinctive feature comes from the presence of the fast magnetosonic wave ion-ion hybrid resonance/cut-off pair, between the antennas and the minority cyclotron layer. In order to document these unusual heating schemes, ICRF experiments were carried out recently on JET. First, the use of 3He ions in H plasmas was investigated with a sequence of discharges in which 5 MW of ICRF power was coupled to the plasma and the 3He concentration was varied from below 1% up to 10%. The inverted minority heating regime was observed at low concentrations (up to ∼2%). Energetic tails in the 3He distribution were observed with effective temperatures up to 300 keV and central electron temperatures up to 6 keV. At around 2%, a sudden transition was reproducibly observed to the mode conversion regime, in which the ICRF fast wave couples to short wavelength modes, leading to efficient direct electron heating and central electron temperature up to 8 keV. All these experiments systematically used power modulation techniques to assess the radial profiles of the wave absorption by the electrons. Secondly, experiments to study the ICRF heating of D minority ions in H were performed. This heating scheme proved much more difficult since modest quantities of C6+ impurity, which has the same Z/A ratio than the D minority ions, led us directly into the mode conversion regime. This effect preventing any absorption by D ions at minority cyclotron layer, could make the (D)H scenario not suitable for the non-active phase of ITER. © 2005 American Institute of Physics [ABSTRACT FROM AUTHOR]

A brief review of the Alfvén Wave Excitation System (AWES) designed for the TCABR tokamak and the first experimental results on RF plasma heating are presented. One of four antenna modules has been completely installed in the vacuum chamber and the initial experiments were carried out in the low power regime using the four-phase RF generator. The main objectives were the antenna tuning according to the typical plasma parameters of TCABR and the evaluation of the antenna parasitic loading, as well as the calibration of the RF diagnostic tools in real discharge conditions. The first results have been obtained with standard diagnostics and with the RF signals measured using high sampling rate digital oscilloscopes. They showed that daily antenna cleaning and correct wave helicity excitation reduce significantly the parasitic loading and are crucial for efficient plasma coupling. [ABSTRACT FROM AUTHOR]

A new regime of high-density runaway discharges has been found in TCABR (Tokamak Chauffage Alfvén in Brazil). This regime is obtained by initiating the discharge with low filling pressure and, after the initial current rise, maintaining a large filling rate. The line density keeps a constant value, ∼10[sup 19]m[sup -3], during the current ramp-up phase and then drops by a factor around 3 or 4 in the quasi-stationary phase of the discharge, when the new regime is formed. The average value of the line density is then kept approximately constant, almost independent of the intensity of the neutral gas injection. The Hα emission increases substantially and shows strong spikes, of the same type observed in the H-mode ELMs, possibly triggered by the relaxation instability typical of runaway discharges. As the filling rate increases, the spike period decreases and its behavior changes as in ELM type I to type III transition observed in experiments in large tokamaks with detached plasmas. Spikes well correlated with the Hα ones are also observed in the time traces of the line average density, MHD fluctuations, ion saturation current in the shadow of the limiter, loop voltage and X-ray emission. [ABSTRACT FROM AUTHOR]