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The solution of the problem of heat exhaust has been pointed out as one of the main challenge towards the realization of magnetic confinement fusion. In the last years, two concepts have been proposed in alternative to the conventional divertor solution adopted for ITER: modification of the magnetic topology in the divertor region and liquid metal as plasma facing component. The role of the Divertor Tokamak Test facility (DTT) in the power exhaust implementation strategy is discussed. The evolution of the project, since the original proposal in 2015 to the present design, is shown. The DTT facility is well integrated in the European strategy and the final decision on the divertor configuration will be made, within 2022-23, on the basis of the indication of the Power Exhaust Group constituted by the EUROfusion Consortium. Finally, the main milestones and the timeline of the project are illustrated. © 2019 Elsevier B.V.

The Divertor Tokamak Test (DTT) facility has been launched to investigate alternative power exhaust solutions for DEMO. DTT should offer sufficient flexibility to be able to incorporate the best candidate divertor concept (e.g. conventional, Snowflake, Super-X, Double Null, liquid metals). In this paper, the revised up-down symmetric DTT device is presented. The up-down symmetrisation of the device makes it possible to have the reference values of the plasma current up to 5.5 MA and, at the same time, it has an impact on the costs, for which a slight revision of the main parameters has been considered. The DTT alternative magnetic configurations, such as Double Null, SnowFlake, Super-X, Double Super-X and Single Null with reverse triangularity, guarantee suitable constraints on the plasma-wall distance and the plasma elongation. The feasibility of the configurations is evaluated in terms of maximum vertical forces and currents on the PF coils along the scenarios.

In this paper we analyze fast ion motion in the divertor tokamak test (DTT) device (Albanese et al 2017 Nucl. Fusion 57 016010). It is planned that DTT will be heated through a mix of 45 MW heating power, including 15 MW negative-ion-based neutral beam heating (NNBI) which is currently being developed by Consorzio RFX in Padova, Italy (Agostinetti et al 2019 Fusion Eng. Design 146 441-446). An issue for DTT is that a toroidal field (TF) ripple with a maximum value of about similar to 0.42% (with respect to the on-axis magnetic field B (0)) is expected on the low-field side, and this ripple interacts with fast ions through the rather well-known phenomena of ripple-precession resonances, in addition to prompt losses of ions which do not complete a full orbit in the poloidal plane. We will show that, with the planned geometry of NNBI, prompt losses are negligible, and ripple-precession losses amount to a maximum of 0.15%. The calculations are performed with the guiding center code Orbit using two different equilibria, and a beam with an energy of 400 keV and the injection angle alpha (inj) = 40 degrees (measured w.r.t. the first wall), which corresponds to a pitch of injected particles lambda = v (parallel to)/v approximate to sin alpha (inj) = 0.65. The main resonances are of the form omega (b) - nN omega (d) = 0, omega (b) and omega (d) being the bounce and precession frequency, respectively, N = 18 the ripple periodicity and 3

The Divertor Tokamak Test facility is an Italian experimental facility under design and construction at ENEA C.R. Frascati. The main goal of DTT is to provide an integrated environment, relevant to DEMO, where testing possible solution to the power exhaust problem in a tokamak (like for example: i) Plasma facing components technology ; ii) Plasma and divertor shape; iii) impurity seeding to increase radiation). In this respect, DTT has been designed to be flexible and adopting technologies relevant to DEMO. After its initial inception in 2015, concluded with the publication of the "DTT project proposal", a complete re-baseline has been provided oncluded with the publication of the "DTT Interim Design Report" in 2019, aimed at accommodating the request of flexibility coming from the international fusion community. During 2019, the engineering integration activity has started and the first construction contracts have been signed. This paper provides an overview of the integrated design activity towards the realization of the facility within 2025.

This paper reviews the research on the reversed field pinch (RFP) in the last three decades. Substantial experimental and theoretical progress and transformational changes have been achieved since the last review (Bodin 1990 Nucl. Fusion 30 1717–37). The experiments have been performed in devices with different sizes and capabilities. The largest are RFX-mod in Padova (Italy) and MST in Madison (USA). The experimental community includes also EXTRAP-T2R in Sweden, RELAX in Japan and KTX in China. Impressive improvements in the performance are the result of exploration of two lines: the high current operation (up to 2 MA) with the spontaneous occurrence of helical equilibria with good magnetic flux surfaces and the active control of the current profile. A crucial ingredient for the advancements obtained in the experiments has been the development of state-of-art active feedback control systems allowing the control of MHD instabilities in presence of a thin shell. The balance between achievements and still open issues leads us to the conclusion that the RFP can be a valuable and diverse contributor in the quest for fusion electricity.

Within the EUROfusion MST1 work package, a series of experiments has been conducted on AUG and TCV devices to disentangle the role of plasma fueling and plasma shape for the onset of small ELM regimes. On both devices, small ELM regimes with high confinement are achieved if and only if two conditions are fulfilled at the same time. Firstly, the plasma density at the separatrix must be large enough (), leading to a pressure profile flattening at the separatrix, which stabilizes type-I ELMs. Secondly, the magnetic configuration has to be close to a double null (DN), leading to a reduction of the magnetic shear in the extreme vicinity of the separatrix. As a consequence, its stabilizing effect on ballooning modes is weakened. Peer Reviewed Article escrit per 365 autors/autores: Labit, B.; Eich, T.; Harrer, G. F.; Wolfrum, E.; Bernert, M.; Dunne, M. 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This paper presents the engineering aspects of the design review of the Italian DTT (Divertor Tokamak Test facility), illustrating the rationale for the design choices and focusing on the main differences with respect to the original proposal.

The significant role of magnetic perturbations in mitigation and losses of runaway electrons (REs) was documented in dedicated experimental studies of RE at the COMPASS tokamak. REs in COMPASS are produced both in low density quiescent discharges and in disruptions triggered by massive gas injection (MGI). The role of the RE seed produced in the beginning of the discharge on the subsequent RE population proved significant. Modulation of the RE losses by MHD instabilities was observed at several characteristic frequencies, as well as by magnetic field oscillations related to power supplies. Magnetic islands seem to suppress the losses as the HXR signal is low and coherent with the island rotation frequency. Moreover, periods of increased losses of REs observed in the current quench (CQ) and early RE beam plateau phase of the MGI disruptions seem to be linked to the bursts of magnetic perturbation, and to the observation of filaments in the fast visible camera images.

An explanation is provided for the disruptive instability in diverted tokamaks when the safety factor$q$at the 95 % poloidal flux surface,$q_{95}$, is driven below 2.0. The instability is a resistive kink counterpart to the current-driven ideal mode that traditionally explained the corresponding disruption in limited cross-sections (Shafranov,Sov. Phys. Tech. Phys., vol. 15, 1970, p. 175) when$q_{edge}$, the safety factor at the outermost closed flux surface, lies just below a rational value$m/n$. Experimentally, external kink modes are observed in limiter configurations as the current in a tokamak is ramped up and$q_{edge}$decreases through successive rational surfaces. For$q_{edge}, the instability is always encountered and is highly disruptive. However, diverted plasmas, in which$q_{edge}$is formally infinite in the magnetohydrodynamic (MHD) model, have presented a longstanding difficulty since the theory would predict stability, yet, the disruptive limit occurs in practice when$q_{95}$, reaches 2. It is shown from numerical calculations that a resistive kink mode is linearly destabilized by the rapidly increasing resistivity at the plasma edge when$q_{95}, but$q_{edge}\gg 2$. The resistive kink behaves much like the ideal kink with predominantly kink or interchange parity and no real sign of a tearing component. However, the growth rates scale with a fractional power of the resistivity near the$q=2$surface. The results have a direct bearing on the conventional edge cutoff procedures used in most ideal MHD codes, as well as implications for ITER and for future reactor options.

This lecture was given at the 4th ITER International Summer School in May 2010 to describe the broad research program on feedback control of magnetohydrodynamic (MHD) stability performed by the rev...

The modified RFX is a very flexible device used for a variety of control schemes for MHD instabilities and for advanced reverse field pinch scenarios. Relative to the previous machine, RFX-mod has a thin Cu shell with vertical field penetration time τS, lowered from 450 to 50 ms and shell/plasma proximity from b/a ≤ 1.24-1.1. Toroidal equilibrium is feedback-controlled and new power supplies provide better Bφ control. Newly designed graphite tiles protect the vessel from localized power deposition. A mesh of 192 external saddle coils, supervised by a digital feedback system, controls radial fields due to field errors and MHD modes. The paper presents an overview of the very encouraging results obtained using both new and 'standard' advanced operational modes in the current range 0.3-1 MA. A dramatic improvement of plasma performance was obtained by using the saddle coils to cancel all the radial field components, an operation mode dubbed virtual shell (VS). The toroidal voltage was lowered by more than 25% and the pulse length was tripled, up to 7 times the τS. Steady-state RFP pulses are now limited only by the applied volt-seconds. The improved magnetic boundary also has an effect on the tearing modes underlying the sustainment of the RFP, whose core amplitude is more than halved. The VS combined with new schemes for the active rotation of the MHD dynamo modes has allowed us to obtain reliable and well-controlled long RFP pulses in the MA current range. This results in a 100% increase in the particle and energy confinement time relative to the previous experiment and opens the possibility of exploring the machine performance in the 2 MA design range. © 2007 IAEA.

Confinement in the reversed field pinch (RFP) has been shown to increase strongly with current profile control. The MST RFP can operate in two regimes: the standard regime with a naturally occurring current density profile, robust reconnection and dynamo activity; and the improved confinement regime with strong reduction in reconnection, dynamo and transport. New results in standard plasmas include the observation of a strong two-fluid Hall effect in reconnection and dynamo, the determination that the m = 0 edge resonant mode is nonlinearly driven, and the determination that tearing modes can lock to the wall via eddy currents in the shell. New results in improved confinement plasmas include observations that such plasmas are essentially dynamo-free, contain several isolated magnetic islands (as opposed to a stochastic field) and contain reduced high frequency turbulence. Auxiliary current drive and heating is now critical to RFP research. In MST, a programme to apply auxiliary systems to the RFP is underway and progress has accrued in several techniques, including lower hybrid and electron Bernstein wave injection, ac helicity injection current drive, pellet injection and neutral beam injection.

We present in this article the description of the upgraded soft x-ray tomographic diagnostic which has been installed in the Madison symmetric torus (MST) reversed field pinch. The previous diagnostic has been extended with two manipulators, for a total of three moveable probes and a fixed one. The line integrated emissivity (brightness) will be measured with arrays of photodiodes and the total number of channels has been increased from 24 to 74. The electronic layout has also been upgraded, and a type of custom made current-to-voltage amplifier has been developed. The diagnostic will be used to continue and extend the magnetohydrodynamics analysis of the MST plasma. The imaging of rotating coherent structures (one or, as recently showed, two) which appear during quasisingle helicity or pulsed poloidal current drive experiments will be performed in more detail. The photocameras can be easily extracted, modified, and inserted again, allowing applications and utilizations of the diagnostic. © 2004 American Institute of Physics.

Growth to large amplitude of a single core-resonant tearing mode in the Madison Symmetric Torus @R. N. Dexter et al., Fusion Technol. 19, 131 ~1991!# reversed-field pinch is accompanied by braking and eventual cessation of mode rotation. There is also a concurrent deceleration of bulk plasma rotation. The mode deceleration is shown to be well described by a time-dependent version of a magnetohydrodynamical model @R. Fitzpatrick et al., Phys. Plasmas 6, 3878 ~1999!# in which a braking torque originates from eddy currents induced by the rotating mode in the conducting shell surrounding the plasma. According to the model, the electromagnetic braking torque is localized to the plasma in the immediate vicinity of the mode's resonant surface, but viscosity transfers the torque to the rest of the plasma. Parametrizing the plasma viscous momentum diffusivity in terms of the global momentum confinement time, the model is used to predict both the momentum confinement time and the time evolution of the decelerating mode velocity. In both respects, the model is quite consistent with experimental data. © 2004 American Institute of Physics.

Evidence of a self-organized collapse towards a narrow spectrum of magnetic instabilities in the Madison Symmetric Torus (R. N. Dexter, D. W. Kerst, T. W. Lovell, S. C. Prager, and J. C. Sprott, Fusion Technol. 19, p.131, 1991) reversed field pinch device is presented. In this collapsed state, dubbed quasi-single helicity (QSH), the spectrum of magnetic modes condenses spontaneously to one dominant mode more completely than ever before observed. For the first time it is observed that the spectra are characterized not only by the increase of one m=1 mode, but also by the simultaneous decrease of the other m=1 modes. Identification of control parameters for the onset of the QSH spectra have also been performed and evidences of a hot core helical structure have been gathered. A higher probability to obtain QSH spectra is found at higher plasma current and at medium-low values of the electron density. The highest values of the SXR emission occur with QSH spectra, a result that is consistent with a hotter plasma core. Moreover, SXR measurements are consistent with the presence of a hotter m=1 helical structure in the plasma core, produced by the dominant MHD mode and associated with the formation of closed helical flux surfaces.

We describe a prototype linear vacuum-ultraviolet x-ray pinhole camera based on the luminescence of a phosphor screen working in reflection mode. The luminescence emitted by the phosphor, in response to radiation selected by a broadband filter, is detected using a 1024 pixel intensified linear array detector. By operating this detector with integration times from 20 to 50 ms, random plasma fluctuations can be smeared out, so that structures related to the magnetic topology can be searched for.

This paper reports the main recent results of the RFX-mod fusion science activity. The RFX-mod device is characterized by a unique flexibility in terms of accessible magnetic configurations. Axisymmetric and helically shaped reversed-field pinch equilibria have been studied, along with tokamak plasmas in a wide range of q(a) regimes (spanning from 4 down to 1.2 values). The full range of magnetic configurations in between the two, the so-called ultra-low q ones, has been explored, with the aim of studying specific physical issues common to all equilibria, such as, for example, the density limit phenomenon. The powerful RFX-mod feedback control system has been exploited for MHD control, which allowed us to extend the range of experimental parameters, as well as to induce specific magnetic perturbations for the study of 3D effects. In particular, transport, edge and isotope effects in 3D equilibria have been investigated, along with runaway mitigations through induced magnetic perturbations. The first transitions to an improved confinement scenario in circular and D-shaped tokamak plasmas have been obtained thanks to an active modification of the edge electric field through a polarized electrode. The experiments are supported by intense modeling with 3D MHD, gyrokinetic, guiding center and transport codes. Proposed modifications to the RFX-mod device, which will enable further contributions to the solution of key issues in the roadmap to ITER and DEMO, are also briefly presented. © 2017 IAEA, Vienna.

A credible roadmap to fusion maximizes performance while still allowing for a safe, efficient and reliable operation of the plant. In this effort control of plasma quantities and off-normal events plays a very important role: the challenge will be not only that of controlling individual quantities, but also that of integration in a harsh nuclear environment. Operation of a fusion reactor will need complete mastering of the plasma. Real time control of MHD stability is a paradigmatic example. Successful control of MHD stability is based in fact on integrated control of both magnetic and kinetic quantities, influences component integrity, plasma-wall interaction and D-T burn and is a requisite for handling off-normal events like disruptions. This final chapter aims at presenting a global view of the open issues and of the potential solutions that characterize the challenge of integrating stability control in an effective fusion scenario.

Magnetic feedback control of the resistive-wall mode has enabled the DIII-D tokamak to access stable operation at safety factor q(95) = 1.9 in divertor plasmas for 150 instability growth times. Magnetohydrodynamic stability sets a hard, disruptive limit on the minimum edge safety factor achievable in a tokamak, or on the maximum plasma current at a given toroidal magnetic field. In tokamaks with a divertor, the limit occurs at q(95) = 2, as confirmed in DIII-D. Since the energy confinement time scales linearly with current, this also bounds the performance of a fusion reactor. DIII-D has overcome this limit, opening a whole new high-current regime not accessible before. This result brings significant possible benefits in terms of fusion performance, but it also extends resistive-wall mode physics and its control to conditions never explored before. In present experiments, the q(95) < 2 operation is eventually halted by voltage limits reached in the feedback power supplies, not by intrinsic physics issues. Improvements to power supplies and to control algorithms have the potential to further extend this regime.

Recent DIII-D and RFX-mod experiments have demonstrated stable tokamak operation at very low values of the edge safety factor q(a) near and below 2. The onset of n = 1 resistive wall mode (RWM) kink instabilities leads to a disruptive stability limit, encountered at q(a) = 2 (limiter plasmas) and q 95 = 2 (divertor plasmas). However, passively stable operation can be attained for q(a) and q 95 values as low as 2.2. RWM damping in the q(a) = 2 regime was measured using active MHD spectroscopy. Although consistent with theoretical predictions, the amplitude of the damped response does not increase significantly as the q(a) = 2 limit is approached, in contrast with damping measurements made approaching the pressure-driven RWM limit. Applying proportional gain magnetic feedback control of the n = 1 modes has resulted in stabilized operation with q 95 values reaching as low as 1.9 in DIII-D and q(a) reaching 1.55 in RFX-mod. In addition to being consistent with the q(a) = 2 external kink mode stability limit, the unstable modes have growth rates on the order of the characteristic wall eddy-current decay timescale in both devices, and a dominant m = 2 poloidal structure that is consistent with ideal MHD predictions. The experiments contribute to validating MHD stability theory and demonstrate that a key tokamak stability limit can be overcome with feedback.

One of the main FAST (Fusion Advanced Studies Torus) goals is to have a flexible experiment capable to test tools and scenarios for safe and reliable tokamak operation, in order to support ITER and help the final DEMO design. In particular, in this paper, we focus on operation close to a possible border of stability related to low-q operation. To this purpose, a new FAST scenario has then been designed at I-p = 10 MA, B-T = 8.5 T, q(95) approximate to 2.3. Transport simulations, carried out by using the code JETTO and the first principle transport model GLF23, indicate that, under these conditions, FAST could achieve an equivalent Q approximate to 3.5. FAST will be equipped with a set of internal active coils for feedback control, which will produce magnetic perturbation with toroidal number n = 1 or n = 2. Magnetohydrodynamic (MHD) mode analysis and feedback control simulations performed with the codes MARS, MARS-F, CarMa (both assuming the presence of a perfect conductive wall and using the exact 3D resistive wall structure) show the possibility of the FAST conductive structures to stabilize n = 1 ideal modes. This leaves therefore room for active mitigation of the resistive mode (down to a characteristic time of 1 ms) for safety purposes, i.e., to avoid dangerous MHD-driven plasma disruption, when working close to the machine limits and magnetic and kinetic energy density not far from reactor values.

This paper describes the soft X-ray (SXR) cross-sectional distributions measured in the European Reversed Field Pinch (RFP) experiment RFX by means of a tomographic diagnostic. Due to the non-invasive character of this diagnostic, the plasma radiation emissivity pattern can be reconstructed. The emissivity depends on various crucial parameters and can be used to study several plasma properties. To reconstruct SXR emissivity in RFX, three different inversion schemes have been compared: the Cormack technique, based on the Bessel-Fourier series expansion, a maximum entropy algorithm and a hybrid method where the previous techniques have been merged together. In this paper the first survey of the SXR emissivity properties in the RFP configuration have been presented. Measurements have been taken in a variety of plasma conditions. The measured SXR data has been compared with the results of a one dimensional impurity diffusion model, which predicts the absolute SXR emissivity, with good agreement. In particular the effect of mode locking on plasma equilibrium in the standard Multiple Helicity (MH) states has been studied. This has given a significant contribute to the description of the plasma MHD structure, both in MH regime and in the Quasi Single Helicity (QSH) state, whose signature is the growth of a sizable m=1 structure in the emissivity distributions, and which has been experimentally discovered thanks to the tomography. A good correlation between SXR tomographic reconstruction and magnetic measurements has been found.

We describe the application of the singular spectrum analysis (SSA), an advanced adaptive statistical technique, to denoise experimental signals and to remove trends in order to isolate the relevant fluctuating components. We illustrate a fast denoising algorithm and show its performances relatively to synthetic and experimental signals of the soft x-ray (SXR) spectrometer and the polarimeter installed in the reversed field experiment (RFX) device. As a further application, we report a first estimate of the electron temperature fluctuations in the core of the RFX experiment. They have been performed with a multifilter SXR spectrometer and applying SSA for the first time in plasma physics the singular spectrum analysis. We find that temperature fluctuations are typically not larger than a few percent and are well correlated with magnetic fluctuations. (C) 2001 American Institute of Physics.

The reversed held pinch (RFP) is a configuration for magnetic confinement akin to the tokamak, but its toroidal field has an amplitude comparable to the poloidal one, and reverses in the outer region. Recent progress in experiments and theory open a path beyond the standard paradigm that a bath of magnetic turbulence is intrinsic to the RFP. In RFX, the largest present RFP experiment, the existence of plasma states with a hot helical core has been proven by soft x-ray tomography. Such states have been seen transiently in several RFPs, but also permanently in discharges in RFX. This last fact makes these states germane to the stationary single-helicity (SH) states displayed by three-dimensional visco-resistive MHD simulations. These simulations also display multiple-helicity (MW) states whose features, in particular magnetic chaos, are analogous to the traditional turbulent state of RFP plasmas. The numerical MH states bifurcate to SH ones when the product (viscosity x resistivity) is increased. The SH states correspond to an integrable magnetic field with good flux surfaces, a feature favourable to good confinement. SH states are not Taylor states, but they may be viewed as the nonlinear state of a resistive kink mode stabilized by the toroidal field reversal.

The reversed field pinch (RFP) is a configuration for plasma magnetic confinement. It has been traditionally viewed as dominated by a bath of MHD instabilities producing magnetic chaos and high energy transport. We report experimental results which go beyond this view. They show a decrease of magnetic chaos and the formation of a coherent helical structure in the plasma, whose imaging and temperature profile are provided for the first time. These quasi-single-helicity states are observed both transiently and in stationary conditions. The last case is consistent with a theoretically predicted bifurcation. Our results set a new frame for improving confinement in high current nonchaotic RFP's.

In RFX a monitor for the measurement of the on-axis electron temperature with high temporal resolution has been designed and operated during the last experimental campaigns. It consists of four soft x-ray (SXR) diodes which observe the plasma through four beryllium filters of different thickness. In this article the main design issues are presented. The measured values of the temperature, computed both with the standard double foil method and with a fit of all the available signals, are reported, correlated with the main plasma parameters and briefly compared with the results of other diagnostics. The main causes of systematic error, the effect of the noise and the reliability of the results in various plasma conditions have been investigated. Dynamic evaluations of the temperature up to a maximum bandwidth of 5 kHz are described, confirming the potentialities of the technique to follow fast phenomena. Possible improvements of the diagnostic and future developments are also pointed out.

In the framework of adding new information to the reversed field pinch (RFP) confinement database, the radiation emitted by RFX plasmas has been investigated using an eight chord bolometric camera, whose detectors have been calibrated absolutely. From the experimental data the emissivity profiles have been reconstructed by means of a generalized tomography reconstruction algorithm. This analysis confirms that the radiation emitted in RFX is systematically concentrated at the edge. The dependence of the emitted power on the plasma density shows that the radiation increases approaching the high density regime, but it rarely goes beyond 30% of the input power for stationary discharges. This behaviour is strongly dependent on the concentration of impurities but, in any case, in RFX there is no evidence of disruptions. A simple local energy balance allows a preliminary evaluation of the radial heat flux profile to be obtained. These measurements indicate that an active impurity screening mechanism is acting in the edge and that transport is the major energy loss mechanism in RFX.

RFX-mod is a toroidal device for fusion plasma magnetic confinement, which can be operated both as a reversed field pinch and as tokamak. In the former configuration it can reach plasma current up to 2 MA, while as a tokamak current is limited to 0.15 MA, due to the toroidal field available. This paper provides an overview of the 2013 RFX physics, both from an experimental and theoretical/numerical point of view. Highlights of the program concern MHD stability feedback control studies - both on physics and engineering -- applied to the RFP and tokamak configurations, edge physics, in particular as far the RFP density behavior is concerned, 3D physics, progress in understanding transport mechanisms, and results of advanced numerical simulations. Initial results on the exposure of various material samples to the RFX plasmas will also be presented. Results on the physics of low qedge tokamak (qedge less than 2) will be discussed.

The Fusion Advanced Study Torus (FAST) has been proposed as a high magnetic field, compact size tokamak providing a flexible integrated environment to study physics and technology issues in ITER and DEMO relevant conditions. FAST has a quite large natural toroidal field ripple (around 1.5%) due to its compactness and to the number of access ports: this ripple must be lowered to an acceptable level to allow safe operations and a good confinement quality. An Active Ripple Compensating System (ARCS) has been designed, based on a set of poloidal coils placed between the plasma chamber and the Toroidal Field Coils (TFCs). These ARCS coils will be fed with adjustable currents, opposite in direction respect to the TFC currents, and will allow lowering the ripple up to zero and beyond. The CAD model of FAST including the ARCS coils has been completed and preliminary electromagnetic and thermal analyses have been carried out. Moreover, a Feedback Active Control System (FACS) composed of two arrays of in-vessel saddle coils has been designed to allow safe high plasma current, low safety factor operation and to mitigate possibly large ELMs effects in FAST. These FACS coils will be fed by a feedback system to control MHD modes: a first engineering assessment of the current requirements has been carried out. © 2013 Euratom-ENEA Association sulla Fusione.

In the reversed field pinch RFX-mod at the highest plasma current of 2 MA, when error fields are not effectively feedback controlled, localized thermal loads up to tens of MW m-2 can be produced. The graphite tiles withstand such high power loads, but the high hydrogen retention makes density control extremely difficult. Several wall conditioning techniques have been optimized in the last campaigns, including helium glow discharge cleaning and wall boronization by diborane glow discharges. More recently, lithium conditioning has been applied for the first time in a reversed field pinch by the evaporation technique. The main results are discussed in this paper. Lithization leads to important operational advantages: a significant improvement of the density control is obtained. Densities up to n/nG ≈ 0.5 can be produced in a controlled way. At the same value of input power, plasmas at higher densities can be sustained. However, due to the short particle confinement time, such densities are reached with high rates of gas puffing and the resulting profiles at high density are edge peaked. A lithium multipellet injector, to be applied in order to obtain a more uniform deposition, has been tested. © 2013 IAEA, Vienna.

RFX-mod [Sonato et al., Fusion Eng. Des. 66, 161 (2003)] exploits its 192 active coils in both reversed-field pinch (RFP) and tokamak configurations with varying degrees of 3D shaping, providing also a test bed for validating stellarator codes and 3D nonlinear magnetohydrodynamic codes. This makes RFX-mod a unique and flexible facility for comparative studies on 3D shaping and control. The paper discusses how 3D fields allow access to RFP and tokamak advanced regimes. 3D fields are used to feedback control Single Helicity (SH) RFP equilibria with 1/7 helicity up to similar to 2 MA. They also allow accessing SH regimes with higher density (Greenwald fraction up to 0.5), presently inaccessible in spontaneous SH regimes. Feedback on the 2/1 resistive-wall mode in RFX-mod tokamak plasmas allows for safe operation at q(a) < 2, an almost unexplored promising regime. Forcing the 2/1 mode to saturate at finite but small level, a helical tokamak equilibrium with significant n=1 modulation is produced and a new way to tailor sawteeth is found. The effects of different levels of 3D shaping on momentum transport in both RFP and tokamak helical states are discussed.

Indirect digital imaging sensors employ tailored phosphors screens to convert incident x-ray or vacuum-ultraviolet (VUV) photons to visible light quanta A convenient method to prepare pixelated phosphor screens that can be easily tailored in thickness, type, and spatial resolution is presented. The characterization and evaluation of these screens in the laboratory is addressed and their application to high-resolution VUV and x-ray cameras is discussed.

Disruptions caused by tearing modes (TMs) are considered to be one of the most critical roadblocks to achieving reliable, steady-state operation of tokamak fusion reactors. Here we have demonstrated a promising scheme to avoid mode locking by utilizing the electromagnetic (EM) torque produced with 3D coils that are available in many tokamaks. In this scheme, the EM torque is delivered to the modes by a toroidal phase shift between the externally applied field and the excited TM fields, compensating for the mode momentum loss through the interaction with the resistive wall and uncorrected error fields. Fine control of torque balance is provided by a feedback scheme. We have explored this approach in two widely different devices and plasma conditions: DIII-D and RFX-mod operated in tokamak mode. In DIII-D, the plasma target was high beta N in a non-circular divertor tokamak. Here beta N is defined as beta N = beta/(I-p/aB(t)) (% Tm/MA), where beta, I-p, a, B-t are the total stored plasma pressure normalized by the magnetic pressure, plasma current, plasma minor radius and toroidal magnetic field at the plasma center, respectively. The RFX-mod plasma was ohmicallyheated with ultra-low safety factor in a circular limiter discharge with active feedback coils outside the thick resistive shell. The DIII-D and RFX-mod experiments showed remarkable consistency with theoretical predictions of torque balance. The application to ignition-oriented devices such as the International Thermonuclear Experimental Reactor (ITER) would expand the horizon of its operational regime. The internal 3D coil set currently under consideration for edge localized mode suppression in ITER would be well suited for this purpose.

Feedback stabilization of the m = 2, n = 1 resistive wall mode in q(a) < 2 plasmas is obtained in the RFX-mod device operated as a circular tokamak. The feedback scheme implements radial sensors within the clean mode control (CMC) algorithm, a technique based on the de-aliasing of the measurements from the high periodicity sidebands produced by the coils. A cylindrical zero-pressure MHD model, which includes a realistic discrete-time description of the feedback, is able to explain the main features of the experimental observation, including the necessity of the removal of sidebands when using radial sensors. The model is also used to show that the radial field CMC presents some advantages in comparison with poloidal field based control schemes, allowing more freedom in the arrangements of the sensors.

RFX-mod is a 2 MA reversed field pinch device (major radius R=2 m, minor radius a=0.457 m) equipped with a system of 192 feedback controlled active coils. This paper describes the recent results of the RFX fusion science program. The 2011 experimental campaign has been dedicated to exploration of RFP confinement in the current range above 1 MA, with particular attention to the optimization of MHD active control and electromagnetic boundary, and to the control of first wall properties and of density profiles. RFX-mod has been operated also as a tokamak, aiming at exploiting and optimizing operation at $q_{edge} \approx $2 with active control of MHD instabilities and resistive wall modes in particular. Results on properties of $q_{edge} \approx $2 tokamak discharges will be presented.

An interesting result of magnetic chaos reduction in RFX-mod high current discharges is the development of strong electron transport barriers. An internal heat and particle transport barrier is formed when a bifurcation process changes the magnetic configuration into a helical equilibrium and chaos reduction follows, together with the formation of a null in the q shear. Strong temperature gradients develop, corresponding to a decreased thermal and particle transport. Turbulence analysis shows that the large electron temperature gradients are limited by the onset of micro-tearing modes, in addition to residual magnetic chaos. A new type of electron transport barrier with strong temperature gradients develops more externally (r/a = 0.8) accompanied by a 30% improvement of the global confinement time. The mechanism responsible for the formation of such a barrier is still unknown but it is likely associated with a local reduction of magnetic chaos. These external barriers develop primarily in situations of well-conditioned walls so that they might be regarded as attempts towards an L–H transition. Both types of barriers occur in high-current low-collisionality regimes. Analogies with tokamak and stellarators are discussed.

RFX-mod self-organized single helical axis (SHAx) states provide a unique opportunity to advance 3D fusion physics and establish a common knowledge basis in a parameter region not covered by stellarators and tokamaks. The VMEC code has been adapted to the reversed-field pinch (RFP) to model SHAx equilibria in fixed boundary mode with experimental measurements as constraint. The averaged particle diffusivity over the helical volume, estimated with the Monte Carlo code ORBIT, has a neoclassical-like dependence on collisionality and does not show the 1/ trend of un-optimized stellarators. In particular, the helical region boundary, corresponding to an electron transport barrier with zero magnetic shear and improved confinement, has been investigated using numerical codes common to the stellarator community. In fact, the DKES/PENTA codes have been applied to RFP for local neoclassical transport computations, including radial electric field, to estimate thermal diffusion coefficients in the barrier region for typical RFX-mod temperature and density profiles. A comparison with power balance estimates shows that residual chaos due to secondary tearing modes and small-scale turbulence still contribute to drive anomalous transport in the barrier region. © 2011 American Institute of Physics.

The wall of any magnetic fusion device is characterized by the presence of several 3D structures, such as portholes for diagnostics and for heating and current drive systems, coil feeds and other features. Time-varying magnetic fields induce eddy currents in the wall, whose pattern is modified by these structures, giving rise to magnetic field errors that can be amplified or shielded by the plasma. Two examples will be given on how the dynamic response of a 3D wall to external magnetic fields can be identified and used to optimize magnetic feedback. In the RFX-mod reversed-field pinch, a dynamic decoupler algorithm has been developed, which allows for the production of pure radial magnetic field harmonics inside the wall, reducing the harmonic distortion due to the 3D wall structures. This is applied here to the problem of producing helical boundary conditions to control helical RFP equilibria. In the DIII-D tokamak, a frequency-dependent scheme for the compensation of the magnetic sensors from spurious n = 1 fields due to the coupling with the feedback and axisymmetric coils has been recently implemented in real time and tested with plasma. The possible relevance of these 3D effects for high performance scenarios is discussed. © 2011 IOP Publishing Ltd.