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The tokamak à configuration variable (TCV) continues to leverage its unique shaping capabilities, flexible heating systems and modern control system to address critical issues in preparation for ITER and a fusion power plant. For the 2019–20 campaign its configurational flexibility has been enhanced with the installation of removable divertor gas baffles, its diagnostic capabilities with an extensive set of upgrades and its heating systems with new dual frequency gyrotrons. The gas baffles reduce coupling between the divertor and the main chamber and allow for detailed investigations on the role of fuelling in general and, together with upgraded boundary diagnostics, test divertor and edge models in particular. The increased heating capabilities broaden the operational regime to include T e/T i ∼ 1 and have stimulated refocussing studies from L-mode to H-mode across a range of research topics. ITER baseline parameters were reached in type-I ELMy H-modes and alternative regimes with ‘small’ (or no) ELMs explored. Most prominently, negative triangularity was investigated in detail and confirmed as an attractive scenario with H-mode level core confinement but an L-mode edge. Emphasis was also placed on control, where an increased number of observers, actuators and control solutions became available and are now integrated into a generic control framework as will be needed in future devices. The quantity and quality of results of the 2019–20 TCV campaign are a testament to its successful integration within the European research effort alongside a vibrant domestic programme and international collaborations.

In magnetic confinement thermonuclear fusion the exhaust of heat and particles from the core remains a major challenge. Heat and particles leaving the core are transported via open magnetic field lines to a region of the reactor wall, called the divertor. Unabated, the heat and particle fluxes may become intolerable and damage the divertor. Controlled ‘plasma detachment’, a regime characterized by both a large reduction in plasma pressure and temperature at the divertor target, is required to reduce fluxes onto the divertor. Here we report a systematic approach towards achieving this critical need through feedback control of impurity emission front locations and its experimental demonstration. Our approach comprises a combination of real-time plasma diagnostic utilization, dynamic characterization of the plasma in proximity to the divertor, and efficient, reliable offline feedback controller design., The exhaust of heat and particles is an important challenge in future nuclear fusion devices. Here the authors report the use of carbon emission as indicator for plasma detachment in a tokamak and its real-time feedback control.

In this paper we present a novel algorithm to extract the optical plasma boundary and radiation front for detached divertor plasmas. We show that reliable detection of the divertor leg and radiation front is possible using lightweight image processing tools. Using a non-tomographic approach, the detected divertor leg and radiation front can be mapped to the poloidal plane. This approach is fast and accurate enough for real-time control purposes, allowing in particular real-time plasma shape and detachment control, and post-shot detachment physics and dynamics analysis.

A model-based approach to real-time reconstruction of the particle density profile in tokamak plasmas is presented, based on a dynamic state estimator. Traditionally, the density profile is reconstructed in real-time by solving an ill-conditioned inversion problem using a measurement at a single point in time. This approach is sensitive to diagnostics errors and failure. The inclusion of a dynamic model in a real-time estimation algorithm allows for reliable reconstruction despite diagnostic errors. Predictive simulations show that the model can reproduce the density evolution of discharges on TCV and ASDEX-Upgrade after tuning of a few parameters. Offline reconstructions using experimental data from TCV show accurate estimation of the density profile and show examples of fault detection of interferometry signals.

Old and recent experiments show that there is a direct response to the heating power of transport observed in modulated ECH experiments both in tokamaks and stellarators. This is most apparent for modulated experiments in the Large Helical Device (LHD) and in Wendelstein 7 advanced stellarator (W7-AS). In this paper we show that: (1) this power dependence can be reproduced by linear models and as such hysteresis (in flux) has no relationship to hysteresis as defined in the literature; (2) observations of 'hysteresis' (in flux) and a direct response to power can be perfectly reproduced by introducing an error in the estimated deposition profile as long as the errors redistribute the heat over a large radius; (3) non-local models depending directly on the heating power can also explain the experimentally observed Lissajous curves (hysteresis); (4) how non-locality and deposition errors can be recognized in experiments and how they affect estimates of transport coefficients; (5) from a linear perturbation transport experiment, it is not possible to discern deposition errors from non-local fast transport components (mathematically equivalent). However, when studied over different operating points non-linear-non-local transport models can be derived which should be distinguishable from errors in the deposition profile. To show all this, transport needs to be analyzed by separating the transport in a slow (diffusive) time-scale and a fast (heating/non-local) time-scale, which can only be done in the presence of perturbations.

This work presents a novel, real-time capable, 10-channel Multispectral Advanced Narrowband Tokamak Imaging System installed on the TCV tokamak, MANTIS. Software and hardware requirements are presented together with the complete system architecture. The image quality of the system is assessed with emphasis on effects resulting from the narrowband interference filters. Some filters are found to create internal reflection images that are correlated with the filters' reflection coefficient. This was measured for selected filters where significant absorption (up to 65% within similar to 70 nm of the filter center) was measured. The majority of this was attributed to the filter's design, and several filters' performance is compared. Tailored real-time algorithms exploiting the system's capabilities are presented together with benchmarks comparing polling and event based synchronization. The real-time performance is demonstrated with a density ramp discharge performed on TCV. The behavior of spectral lines' emission from different plasma species and their interpretation are qualitatively described.

For control of neoclassical tearing modes (NTMs) and the resulting rotating magnetic islands in tokamak plasmas, the frequency and phase of the magnetic islands need to be accurately tracked in real-time. In previous experiments on TEXTOR, this was achieved using a phase-locked loop (PLL). For ASDEX Upgrade however, the desired frequency range in which the islands are to be tracked (100 Hz-10 kHz) is much larger than is possible with a PLL. In this contribution, an extended Kalman filter (EKF) and an unscented Kalman filter (UKF) are proposed for real-time frequency, phase and amplitude tracking of sinusoidal signals, based on noisy measurements. Compared to PLLs, the EKF and UKF are able to track sinusoidal signals in a much larger frequency range. The filters are applied on synthetic data and on experimental data from the TEXTOR and TCV tokamaks, from which we conclude that the UKF can be useful for real-time control of magnetic islands on ASDEX Upgrade. (C) 2013 Elsevier B.V. All rights reserved.

The development of the control system for a tokamak demonstration fusion reactor (DEMO) faces unprecedented challenges. First, the requirements for control reliability and accuracy are more stringent than on existing fusion devices: any loss of plasma control on DEMO may result in a disruption which could damage the inner wall of the machine, while operating the device with larger margins against the operational limits would lead to a reduction of the electrical output power. Second, the performance of DEMO control is limited by space restrictions for the implementation of components (optimization of the tritium breeding rate), by lifetime issues for the front-end parts (neutron and gamma radiation, erosion and deposition acting on all components) and by slow, weak and indirect action of the available actuators (plasma shaping, heating and fuelling). The European DEMO conceptual design studies include the development of a reliable control system, since the details of the achievable plasma scenario and the machine design may depend on the actual performance of the control system. In the first phase of development, an initial understanding of the prime choices of diagnostic methods applicable to DEMO, implementation and performance issues, the interrelation with the plasma scenario definition, and the planning of necessary future R&D have been obtained.

The development of scenarios characterized by H-mode confinement and internal transport barriers (ITBs) in high triangularity, δ ~ 0.4–0.5, discharges is of particular interest for ITER advanced tokamak operation. Previous JET experiments have shown that high triangularity favours H-modes which are ELM-free or develop type I edge localized mode (ELM) activity, which inhibits long lasting ITBs. The recent experiments reported here concentrate on integrated optimization of edge and core conditions. The stability of the edge pedestal was controlled using gas injection, deuterium or light impurities, and plasma current ramps. Both methods yield more ITB-friendly edge pedestal conditions, varying from small type I to type III ELMs and, in extreme cases, resulting in L-mode. In parallel, the conditions for triggering and sustaining ITBs encompassing a large proportion of the plasma volume (outer ITBs) were optimized, as opposed to less performing ITBs located closer to the plasma centre (inner ITB). These plasmas have deeply reversed target current profiles with qmin ~ 3 and a narrow inner ITB, located typically at a small normalized radius ρ < 0.5 and close to the reversed shear region, is routinely observed. Large radius outer ITBs are only triggered at an input power in excess of 20 MW, but they do not usually survive the transition into H-mode. The best results, in terms of sustained high confinement, have been obtained with neon injection; an outer ITB is triggered during the phase with L-mode edge and survives into H-mode for about 2 s, corresponding to ~10 times the global energy confinement time τE, at q95 = 7.5, H89βN ~ 3.5–4 and ~60% of the Greenwald density limit. In summary, a high triangularity scenario has been developed, which combines the desirable characteristics of controlled edge, long lasting wide ITBs and high performance at density higher than the low triangularity JET scenarios.

A 800 kW, 140 GHz gyrotron has been mounted on TEXTOR. Initial operation was limited by arcs in the transmission line, and by pulse length restrictions due to the expected limited power handling of the window and the launching mirror. Measurements of power absorption in the window and launching mirror, however, showed them to be capable of handling full power, 3 s pulses, while arcing could be prevented by absorption of stray radiation at strategic position in the transmission line. This resulted in the realisation of full power, 3 s pulses on the plasma. Further measures were taken to prevent gyrotron operation in wrong modes or under conditions of low absorption in the plasma. Future improvements, expected to allow further extension of the pulse length to 10 s include: a CVD diamond window to replace the current quartz torus vacuum window, and a fast and accurate launcher.

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.

We have investigated the effect of different ion cyclotron resonance frequency (ICRF) heating schemes, of error field modes, of the plasma shape and edge magnetic shear, and of the ion delB drift direction on the stability of Alfven eigenmodes (AEs). The use of multi-frequency or 2nd harmonic minority ICRF heating at high plasma density gives rise to a lower fast ion pressure gradient in the plasma core and to a reduced mode activity in the Alfven frequency range. Externally excited low-amplitude error fields lead to a much larger AE instability threshold, which we attribute to a moderate radial redistribution of the fast ions. The edge plasma shape has a clear stabilizing effect on high-n, radially localized AEs. The damping rate of n = 1 toroidal AEs is a factor 3 higher when the ion VB drift is directed towards the divertor. These results represent a useful step towards the extrapolation of current scenarios to the inclusion of fusion-born alpha particles in ITER, with possible application for feedback control schemes for the various ITER operating regimes.

Ion cyclotron resonance frequencies (ICRF) mode conversion has been developed for localized on-axis and off-axis bulk electron heating on the JET tokamak. The fast magnetosonic waves launched from the low-field side ICRF antennas are mode-converted to short-wavelength waves on the high-field side of the He-3 ion cyclotron resonance layer in D and He-4 plasmas and subsequently damped on the bulk electrons. The resulting electron power deposition, measured using ICRF power modulation, is narrow with a typical full-width at half-maximum of approximate to30 cm (i.e. about 30% of the minor radius) and the total deposited power to electrons comprises at least up to 80% of the applied ICRF power. The ICRF mode conversion power deposition has been kept constant using He-3 bleed throughout the ICRF phase with a typical duration of 4-6 s, i.e. 15-40 energy confinement times. Using waves propagating in the counter-current direction minimizes competing ion damping in the presence of co-injected deuterium beam ions.

We demonstrate the use of collective Thomson scattering (CTS) for spatially localized measurements of the isotopic composition of magnetically confined fusion plasmas. The experiments were conducted in the TEXTOR tokamak by scattering millimeter-wave probe radiation off plasma fluctuations with wave vector components nearly perpendicular to the magnetic field. Under such conditions the sensitivity of the CTS spectrum to plasma composition is enhanced by the spectral signatures of the ion cyclotron motion and of weakly damped ion Bernstein waves. Recent experiments on TEXTOR demonstrated the ability to resolve these signatures in the CTS spectrum as well as their sensitivity to the ion species mix in the plasma. This paper shows that the plasma composition can be inferred from the measurements through forward modeling of the CTS spectrum. We demonstrate that spectra measured in plasmas consisting of hydrogen, deuterium and (3)He can be accurately reproduced by theory and yield inferred plasma compositions consistent with expectations. The potential to use CTS for measurements of plasma composition is of significant interest since CTS is well suited for reactor environments and since there is at present no established method to measure the fuel ion density ratio in the core of a burning fusion plasma.

The systematic design of a robust adaptive control strategy for the sawtooth period using electron cyclotron current drive (ECCD) is presented. Recent developments in extremum seeking control (ESC) are employed to derive an optimized controller structure and offer practical tuning guidelines for its parameters. In this technique a cost function in terms of the desired sawtooth period is optimized online by changing the ECCD deposition location based on online estimations of the gradient of the cost function. The controller design does not require a detailed model of the sawtooth instability. Therefore, the proposed ESC is widely applicable to any sawtoothing plasma or plasma simulation and is inherently robust against uncertainties or plasma variations. Moreover, it can handle a broad class of disturbances. This is demonstrated by time-domain simulations, which show successful tracking of time-varying sawtooth period references throughout the whole operating space, even in the presence of variations in plasma parameters, disturbances and slow launcher mirror dynamics. Due to its simplicity and robustness the proposed ESC is a valuable sawtooth control candidate for any experimental tokamak plasma, and may even be applicable to other fusion-related control problems.

A demonstration of real-time feedback control for autonomous tracking and stabilization of m/n = 2/1 tearing modes in a tokamak using Electron Cyclotron Resonance Heating and Current Drive (ECRH/ECCD) is reported. The prototype system on TEXTOR combines in the same sight-line an Electron Cyclotron Emission (ECE) diagnostic for tearing mode sensing and a steer-able ECRH/ECCD antenna. The mode location is retrieved from the ECE measurements and serves as input for a control loop, which aligns the ECRH/ECCD deposition with the tearing modes by steering of a launcher mirror. The alignment is achieved by matching the mode location in the sensor spectrum with the fixed ECRH/ECCD actuator frequency. The control response is dominated by the response of the mechanical launcher. Analysis of the launcher dynamics receives special emphasis in the control design. In addition, the ECRH/ECCD power is modulated in phase with the rotation frequency of the O-point of the tearing modes using a feedback loop, which extracts the modes frequency and phase from the ECE data. The 'line-of-sight' concept has advantages over conventional tearing mode suppression schemes as it provides information of the modes location and phase in the direct vicinity of the EC-deposition. It avoids the use of ray-tracing in the control loop and guarantees accurate alignment in the presence of disturbances. The experimental results demonstrate the capabilities of the control system to track and suppress tearing modes in real-time. A relatively simple control design suffices to meet the performance requirements demanded for effective tearing mode suppression. In addition, a control-oriented model for the tearing mode control problem is being developed. This model includes tearing mode and plasma dynamics, models for the actuators, diagnostics, data-processing and control algorithms. Simulation results for typical TEXTOR conditions will be discussed and validated with experimental results.

High-power millimetre wave beams employed on ITER for heating and Current drive at the 170 GHz electron cyclotron resonance frequency require agile steering and tight focusing of the beams to suppress neoclassical tearing modes. This paper presents experimental validation of the remote steering (RS) concept of the ITER upper port millimetre wave beam launcher. Remote steering at the entrance of the upper port launcher rather than at the plasma side offers advantages in reliability and maintenance of the mechanically Vulnerable steering system. A one-to-one scale mock-up consisting of a transmission line, mitre bends, remote steering unit, vacuum window, square corrugated waveguide and front mirror simulates the ITER launcher design configuration. Validation is based on low-power heterodyne measurements of the complex amplitude and phase distribution of the steered Gaussian beam. High-power (400 kW) short Pulse (10 ms) operation under vacuum, diagnosed by calorimetry and thermography of the near- and far-field beam patterns, confirms high-power operation, but shows increased power loss attributed to deteriorating input beam quality compared with low-power operation. Polarization measurements show little variation with steering, which is important for effective current drive requiring elliptical polarization for O-mode excitation. Results show that a RS range Of LIP to -12 to +12 can be achieved with acceptable beam quality. These measurements confirm the back-up design of the ITER ECRH&CD launcher with future application for DEMO. (C) 2009 Elsevier B.V. All rights reserved.

In nuclear fusion the sawtooth instability is an important plasma phenomenon, having both positive and negative effects on the tokamak plasma. Control of its period is essential in future nuclear fusion reactors. This paper presents a control oriented model of the sawtooth instability, with current drive as input and sawtooth period as output, and validates it with a more extended model and experimental data. This model is numerically implemented and combined with PI-controllers in a feedback loop. Simulations show that closed loop control of the system to a desired sawtooth period is indeed possible, even when time delay is added to the feedback loop. © 2010 AACC.

The impurity seeded type-III ELMy H-mode is proposed as an integrated ITER scenario. At JET this scenario has been demonstrated up to plasma currents of 3 MA with nitrogen as seeding gas. Detached divertor operation is achieved with significantly reduced steady state and transient heat fluxes. By operating in a high triangular magnetic configuration, very high central line-averaged electron densities are reached (up to 1.1 x 10(20) m(-3)). The impurity sources and impurity concentration in the plasma core were investigated for a broad range of plasma conditions in those highly radiating plasmas. The results are compared to an integrated model of the plasma edge and plasma core. On the basis of this modelling extrapolations to ITER are done with neon seeding, giving more confidence that these radiative plasmas will have a power amplification factor in excess of 10. (C) 2009 J. Rapp. Published by Elsevier B.V. All rights reserved.

Neoclassical tearing modes (NTMs) deteriorate high-pressure tokamak plasma confinement and can be suppressed by electron cyclotron current drive (ECCD). In order to obtain efficient suppression, the ECCD power needs to be deposited at the center of an NTM magnetic island. To enhance efficiency, this power also needs to be synchronized in phase with the rotation of the island. The problem is that of real-time detection and precise localization of the island(s) in order to provide the feedback signal required to control the ECCD power deposition area with an accuracy of 1 to 2 cm. Existing schemes based on mode location, equilibrium reconstruction, and plasma profile measurements are limited in positional and temporal accuracy and moreover will become very complex when applied to ITER. To overcome these limitations, it is proposed to provide the feedback signal from electron cyclotron emission (ECE) measurements taken along the identical line of sight as traced by the incident ECCD millimeter-wave beam but in reverse direction. Experiments on TEXTOR have demonstrated a proof of principle. These measurements motivate the further development and the implementation of such an ECCD-aligned ECE system for NTM control in larger fusion machines. Possible implementation of such a system on ASDEX-Upgrade, based on waveguides equipped with a fast directional switch, is presented in this paper. Possible further development for ITER is also discussed.

Tokamak plasmas in nuclear fusion are subject to various instabilities. A clear example is the sawtooth instability, which has both positive and negative effects on the plasma. To optimize between these effects control of the sawtooth period is necessary. This paper presents a simple control oriented model, from current drive (input) to sawtooth period (output), which can mimic the most relevant aspects of the sawtooth instability. It also shows some simulation results, including a static input-output map and a comparison with experimental data.

In ITER, the requirement for the ICRF antenna is to deliver 20 MW in ELMy H‐mode plasmas with an averaged antenna—plasma separatrix distance of 14 cm [1]. Two major problems will have to be solved: the very fast change in antenna loading during ELMs and the decrease of the loading when the plasma is pushed far away from the antenna. JET has the capability to combine these conditions and for the first time, experiments were performed in ELMy H‐mode at antenna—separatrix distance, referred as ROG, varied from 10 to 14 cm. When ROG was increased, the perturbation caused by ELMs was found to decrease significantly and the loading between ELMs was found to deteriorate to very low values. In order to compensate the latter unwanted effect, different levels of deuterium gas were injected in the edge either from the divertor, the midplane or the top of the tokamak. Using this technique, the loading was increased by up to a factor 6 and up to 8 MW of ICRF power were coupled.

The dynamical behavior of the global, two-fluid, electromagneticmodel of a tokamak plasma is explored under conditions corresponding to the Rijnhuizen tokamak project [A. J. H. Donne, Plasma Phys. Rep.20, 192 (1994)] using the CUTIE code [A. Thyagaraja, Plasma Phys. Controlled Fusion42, B255 (2000)]. Simulations of an off-axis electron-cyclotron-heated ( 350 kW ) hydrogen discharge and a purely Ohmic one over several resistive evolution times ( τ res ≃ 15 − 20 ms ) are described. The results illustrate profile-turbulence interactions and the spectral transfer processes implicated in the spontaneous generation and maintenance of mesoscale zonal flows and dynamo currents. Relaxation phenomena, including off- and on-axis sawteeth and periodically repeating edge ballooning instabilities mediated by these mechanisms, are presented. The CUTIEmodel reproduces many observed features of the experiment qualitatively and suggests that global electromagnetic simulations may play an essential role in understanding tokamakturbulence and transport.

A two-fluid computer model of electromagnetic tokamak turbulence, CUTIE, is used to study the dynamic structure and turbulent transport in the Rijnhuizen Tokamak Project tokamak. A discharge with dominant, off-axis electron cyclotron heating is the main focus of the simulations which were extended over several resistive diffusion times. CUTIE reproduces the turbulent transport and MHD phenomena of the experiment. The noninductive components of the current density profile, viz., the dynamo current and the bootstrap current, are identified as key players in the turbulent transport and its suppression and in off-axis MHD events.

ELMy H-modes in helium-4 plasmas provide valuable information on ELMy H-mode physics as well as a possible early low activation operational phase for next-step tokamaks, such as ITER. With this in mind, a series of helium-4 H-mode experiments were performed on JET with pure helium-4 NBI auxiliary heating (up to 12 MW). A set of ELMy H-mode plasmas were produced, in both the Type I ELM regime and a second regime, which showed characteristics similar to the deuterium Type III regime, but with a reverse ELM frequency dependence on power. Sawteeth were also observed, and had similar behaviour to those seen in deuterium. Compared with deuterium plasmas, Type I ELMy H-mode confinement is seen to be 28 +/- 6% poorer in helium-4 plasmas and the L-H power threshold 42 +/- 10% larger. This is the opposite of the behaviour predicted by experimental isotope mass scalings from hydrogenic plasmas.Comparison with a wider hydrogenic database, enables the effects of isotopic charge and mass to be studied independently.

Recently, the dynamic ergodic divertor (DED) of TEXTOR has been studied in an m/n = 3/1 set-up which is characterized by a relatively deep penetration of the perturbation field. The perturbation field creates (a) a helical divertor, (b) an ergodic pattern and/or (c) excitation of tearing modes, depending on whether the DED current is static, rotating in the co-current direction or in the counter-current direction. Characteristic divertor properties such as the high recycling regime or enhanced shielding have been studied. A strong effect of the ergodization is spin up of the plasma rotation, possibly due to the electric field at the plasma edge. Tearing modes are excited in a rather reproducible way and their excitation threshold value, their motion and their reduction due to the ECRH/ECCD have been studied. The different scenarios are characterized by strong modifications of the toroidal velocity profile and by a reduced or enhanced radial transport.

Experiments in the Rijnhuizen Tokamak Project are reported which show bifurcated states of plasmas in the linear Ohmic confinement regime. One branch is the usual Ohmic state. The second, newly discovered, state features a much broader temperature profile, while the density has a very pronounced gradient near the half radius. The bifurcated states form after off-axis electron cyclotron heating is switched off. The new state contradicts theoretical models which predict improved confinement for peaked density profiles or high collisionality. [S0031-9007(98)08046-6].

In the Rijnhuizen Tokamak Project [RTP, A. J. H. Donne, the RTP Team, Plasma Phys. Rep. 20, 192 (1994)], experiments have been carried out with Electron Cyclotron Heating (ECH) at four to seven times the Ohmic power, with a deposition width < 10% of the minor radius. Central deposition leads to sharp electron temperature (T-e) gradients near the q=1 surface. Off-axis deposition results in steady-state hollow T-e profiles. A radial scan of the ECH deposition reveals a discrete, stepwise response of T-e profile. The transitions occur for a displacement of the deposition by less than 1 mm. Each transition is due to the loss of a transport barrier. The transitions occur when the minimum value of q passes through the values 1, 4/3, 5/2, 2, 7/2 and 3. A numerical model featuring transport barriers near those q values gives good reproduction of the experimental data. The possible role of the topological properties of the magnetic field is discussed. (C) 1999 American Institute of Physics. [S1070-664X(99)00812-5].

A family of off-axis, or annular, instabilities has been studied using Thomson scattering, soft X-ray emission, and two electron cyclotron emission diagnostic systems. In the Rijnhuizen tokamak (RTP) [N. J. Lopes Cardozo , Plasma Physics and Controlled Nuclear Fusion Research 1992 (International Atomic Energy Agency, Vienna, 1993), Vol. 1, p. 271] these phenomena are invoked in a controlled way in discharges with specific (off-axis) deposition of electron cyclotron heating (ECH) and persist during most of the heating period, or during many current diffusion times. Based on coherent mode analysis at the crash time, the instabilities are associated with resonant surfaces near simple rational values of q (3/2, 2, and 3). A parameter study shows an increase of reheat rate and a decrease of sawtooth period with increasing ECH power and - in contrast to observations in other experiments - with increasing density as well. (C) 1999 American Institute of Physics. [S1070-664X(99)02210-7].

The flux surfaces' layout and the magnetic winding number q are important quantities for the performance and stability of tokamak plasmas. Normally, these quantities are iteratively derived by solving the plasma equilibrium for the poloidal and toroidal flux. In this work, a fast, non-iterative and magnetics-free numerical method is proposed to estimate the shape of the flux surfaces by an inward propagation of the plasma boundary shape, as can be determined for example by optical boundary reconstruction described in Hommen (2010 Rev. Sci. Instrum. 81 113504), toward the magnetic axis, as can be determined independently with the motional Stark effect (MSE) diagnostic. Flux surfaces are estimated for various plasma regimes in the ITER, JET and MAST tokamaks and are compared with results of CRONOS reconstructions and simulations, showing agreement to within 1% of the minor radius for almost all treated plasmas. The availability of the flux surface shapes combined with the pitch angles measured using MSE allow the reconstruction of the plasma q-profile, by evaluating the contour-integral over the flux surfaces of the magnetic field pitch angle. This method provides a direct and exact measure of the q-profile for arbitrary flux surface shapes, which does not rely on magnetic measurements. Results based on estimated flux surface shapes show agreement with CRONOS q-profiles of better than 10%. The impact of the shape of the flux surfaces on the q-profile, particularly the profiles of elongation and Shafranov shift, and offsets in plasma boundary and the magnetic axis are assessed. OFIT+ was conceived for real-time plasma profile control experiments and advanced tokamak operation, and provides quickly and reliably the mapping of actuators and sensors to the minor radius as well as the plasma q-profile, independent of magnetic measurements.

An intermediate frequency (IF) band digitizing radiometer system in the 100-200 GHz frequency range has been developed for Tokamak diagnostics and control, and other fields of research which require a high flexibility in frequency resolution combined with a large bandwidth and the retrieval of the full wave information of the mm-wave signals under investigation. The system is based on directly digitizing the IF band after down conversion. The enabling technology consists of a fast multi-giga sample analog to digital converter that has recently become available. Field programmable gate arrays (FPGA) are implemented to accomplish versatile real-time data analysis. A prototype system has been developed and tested and its performance has been compared with conventional electron cyclotron emission (ECE) spectrometer systems. On the TEXTOR Tokamak a proof of principle shows that ECE, together with high power injected and scattered radiation, becomes amenable to measurement by this device. In particular, its capability to measure the phase of coherent signals in the spectrum offers important advantages in diagnostics and control. One case developed in detail employs the FPGA in real-time fast Fourier transform (FFT) and additional signal processing. The major benefit of such a FFT-based system is the real-time trade-off that can be made between frequency and time resolution. For ECE diagnostics this corresponds to a flexible spatial resolution in the plasma, with potential application in smart sensing of plasma instabilities such as the neoclassical tearing mode (NTM) and sawtooth instabilities. The flexible resolution would allow for the measurement of the full mode content of plasma instabilities contained within the system bandwidth. (C) 2011 American Institute of Physics.

Real-time plasma electron density profile estimation and control are essential in the operation of future tokamaks. In particular, the robustness against diagnostics failure and disturbances is important for long pulse operation., A model-based approach to profile estimation is implemented on the control systems of ASDEX Upgrade and TCV, which is able to merge information from various diagnostics for both core and edge density, as well as systematically handling diagnostic failure. The model used for profile estimation is employed to tune a feedback controller before an experiment, thereby reducing the experimental time required for tuning., Subsequently, this observer and controller have been employed in scientific experiments on ASDEX Upgrade and TCV. On ASDEX Upgrade, the density profile estimator was used in high-density pellet-fuelled discharges, providing a more reliable real-time estimate of the core density for feedback control than previously achieved. On TCV, in experiments on integrated pressure and safety factor profile control, the density profile estimator and feedback controller provide a constant density despite disturbances from time-varying ECCD power. Additionally, the real-time density profiles provide an essential input for other real-time plasma state estimation codes including Electron Cyclotron ray tracing codes, contributing to a complete real-time estimation of the entire plasma state.

Experimental evidence from the JET tokamak is presented supporting the predictions of a recent theory (Graves et al 2009 Phys. Rev. Lett. 102 065005) on sawtooth instability control by toroidally propagating ion cyclotron resonance waves. Novel experimental conditions minimized a possible alternate effect of magnetic shear modification by ion cyclotron current drive, and enabled the dependence of the new energetic ion mechanism to be tested over key variables. The results have favourable implications on sawtooth control by ion cyclotron resonance waves in a fusion reactor.

A dual, high speed, real-time visible light camera setup was installed on the TCV tokamak to reconstruct optically and in real-time the plasma boundary shape. Localized light emission from the plasma boundary in tangential view, broadband visible images results in clearly resolved boundary edge-features. These projected features are detected in real-time and transformed to the poloidal plane to obtain a measurement of the plasma boundary. Plasma boundary reconstructions of diverted plasma discharges are presented, showing agreement of within 1 cm compared with magnetic equilibrium reconstruction. The resulting real-time plasma shape measurement is applied in a feedback control loop for the plasma position, demonstrating effective stabilization and tracking of the plasma vertical position. © 2014 EURATOM.

The successful performance of a model predictive profile controller is demonstrated in simulations and experiments on the TCV tokamak, employing a profile controller test environment. Stable high-performance tokamak operation in hybrid and advanced plasma scenarios requires control over the safety factor profile (q-profile) and kinetic plasma parameters such as the plasma beta. This demands to establish reliable profile control routines in presently operational tokamaks. We present a model predictive profile controller that controls the q-profile and plasma beta using power requests to two clusters of gyrotrons and the plasma current request. The performance of the controller is analyzed in both simulation and TCV L-mode discharges where successful tracking of the estimated inverse q-profile as well as plasma beta is demonstrated under uncertain plasma conditions and the presence of disturbances. The controller exploits the knowledge of the time-varying actuator limits in the actuator input calculation itself such that fast transitions between targets are achieved without overshoot. A software environment is employed to prepare and test this and three other profile controllers in parallel in simulations and experiments on TCV. This set of tools includes the rapid plasma transport simulator RAPTOR and various algorithms to reconstruct the plasma equilibrium and plasma profiles by merging the available measurements with model-based predictions. In this work the estimated q-profile is merely based on RAPTOR model predictions due to the absence of internal current density measurements in TCV. These results encourage to further exploit model predictive profile control in experiments on TCV and other (future) tokamaks.