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The magnetic fault detection system of RFX (named RGM, from the italian acronym Rilevamento Guasti sistemi Magnetici), originally developed in early 1990s and refurbished in 2000s, has been designed and manufactured with the aim of detecting, within one millisecond, electromagnetic faults of the windings and their circuits occurring during the operation of RFX. RGM is designed to guarantee a high degree of reliability to ensure the integrity of the machine. The magnetic configuration of RFX is characterized by a high level of symmetry: different windings are connected in order to carry the same current (i.e., the four sectors of magnetizing winding are connected in series, the field shaping windings has up-down symmetry of the currents). RGM exploits these specific features to detect imbalances between symmetric circuits, ensuring a fast and reliable intervention in case of fault. Other types of protection are foreseen to detect abnormal currents or voltages in critical parts of the system. However, the RGM system needs to be updated for several reasons. Firstly, the modifications planned for the upgraded RFX-mod2 are incompatible with the layout of some of the original sensors. Moreover, RGM does not allow operations with non-symmetric plasma cross-section (i.e., upper/lower single null configurations), as these configurations would require breaking the symmetry of the windings, giving an erroneous fault identification (false positive). This paper describes the upgrade of the RGM system, which will be equipped with several additional magnetic sensors to overcome the aforementioned limitations, and refurbished with re-designed electronic boards based on state of the art technologies and components. A model-based approach has been followed: the modelling activity involved simulations with both lumped-parameter models of the plasma and passive structures, and more detailed FEM models describing the non-linear evolution of plasma in presence of the passive structures. These simulations, which extend and update the effort done to design the original RGM, have the aim of analyzing several fault conditions to properly characterize the evolution of current/ voltages during these abnormal events, as well as the electrodynamic forces acting on the copper shell, and tune the protections to guarantee a suitable fast detection of the faults.

A major refurbishment of the toroidal complex of the RFX-mod device is in progress and it will include the removal of the Inconel vacuum vessel and a modification of the stainless steel supporting structure to be made vacuum-tight. The axisymmetric equilibrium control in RFX-mod was responsible for the control of the plasma current, horizontal and vertical position along with its stability and the plasma shape. The new implementation framework, both hardware and software, is described. The system has been fully reviewed and modified, if needed, for its prospective use in RFX-mod2. In order to run tests in an operation-like context, the updated system has also been implemented in the MARTe2 framework, to be deployed for the real-time applications of RFX-mod2. The results of the previous versions were reproduced and the cycle-time requirements were met.

The RFX-mod Reversed Field Pinch device passive boundary is being improved: - Drastic reduction of resistivity of first shell surrounding the plasma; - Reduction of plasma-stabilizing conductor distance from b/a=1.11 to b/a=1.04. The RFX-mod core upgrades consist of: - Removal of Inconel vacuum vessel; - Modification of the stainless steel Support Structure to ensure Vacuum Tightness (VTSS); - Modification of the copper Passive Stabilizing Shell (PSS); - Installation of upgraded sensors inside the vacuum vessel. Initial main points of investigation in the new device are discussed.

In the RFX-mod2 experiment presently under construction, devoted to the study of magnetic confinement of fusion relevant plasmas, significant electric fields, in the kV/mm range, are expected to form in between in-vessel conductive plasma facing components during transient plasma current phases (start-up and termination). While such electric fields are of no concern for components in vacuum, the presence of a scrape-off plasma at the edge (electron density ne 1016 ÷1018 m-3, electron temperature Te of few eV) can create the conditions for potentially dangerous arc formation. For this reason part of the plasma facing components (in particular the graphite 'first wall' tiles covering the copper 'stabilizing shell' placed within the plasma chamber) require a proper conditioning technique capable of maintaining the insulation between conductive components even in presence of the scrape-off plasma. An experimental apparatus has been developed in order to test the conditions for the arc formation and prevention between two electrodes immersed in a plasma generated by a hot emitting filament. The results of an extensive experimental campaign will be presented, aimed at demonstrating the possibility of gaining a sufficient electrical conditioning by applying the standard conditioning technique usually employed for higher voltage ranges. It consists of a sequence of high voltage pulses applied to the pair of electrodes with current limitation, in the presence of a background cold plasma with low ionization degree. The experimental procedure is such that the voltage of the pulses is slightly increased when arcing ceases, until the final desired voltage level is achieved (2.5 kV). Different electrode materials have been tested in a variety of plasma conditions in terms of electron density and working gas pressure.

RFXMod2 is an upgrade of RFXMod that will use a modified shell and mechanical structure in order to enhance plasma proximity to the shell and therefore to improve plasma control. The Control and Data Acquisition System for most of the plant systems and diagnostics previously used in RFXMod will be refurbished, while others will be completely re-built. The most important component that will be completely renewed is the Electromagnetic Probes (EM) data acquisition system. In the former system data acquisition and real-time control were implemented using two different technologies, i.e. CompactPCI and VME, respectively. In RFXMod2 a new architecture based on XILINX Zynq FPGA will be used to carry out at the same time both high-speed data acquisition and resampled data streaming for active plasma control. Moreover, numeric integration of EM signals will be carried out by the same FPGA supervising data acquisition and streaming. The use of MDSplus will be retained in RFXMod2, while the MARTe framework used for real-time plasma control will be replaced by MARTe2, a new framework developed under strict software quality standards. Plant control in RFXMod2 will be supervised by WinCC-OA, replacing the previous FactoryLink SCADA systems. Older plant units such as vacuum control using outdated S5 PLCs will be replaced by new systems, using OPC communication with the supervisory system. Old CAMAC equipment in diagnostics will be replaced by CompactPCI technology, making it uniform with the rest of the diagnostic data acquisition systems. Finally, specialized devices implemented in CAMAC such as timing sequence generators will be replaced by custom devices implemented using the same Zynq FPGA technology of the EM data acquisition devices.

The design of the RFX-mod2 fusion experimental device requires a 3 mm copper shell close to the plasma to achieve improved plasma confinement properties. Such conductive structure, located around the plasma and aimed to passively stabilize the plasma, shall have an electrical discontinu-ity in both the poloidal and toroidal directions, in order to allow the penetration of electromagnetic fields into the plasma region. These gaps avoid the formation of net poloidal and toroidal eddy cur-rents during transient phases of the plasma confinement experiment. Moreover, the shell has been designed with an overlapping region at the poloidal gap in order to reduce the induced field errors. During operations in Reversed Field Pinch magnetic configuration, the loop voltage, that is the externally induced electromotive force sustaining the plasma current, can reach values up to 400 V. These values can rapidly step up to 1.5 kV during fast plasma current terminations. Therefore, intense electric fields can generate between the shell flaps, only a few millimetres apart, along the overlapping region. Furthermore, taking into account that the stabilizing shell, being located inside the vacuum chamber, is exposed to low temperature plasma, the formation probability of harmful electric arcs is high. In order to avoid arc formation, different kind of insulation coating on the copper, able to with-stand the applied electric fields in the presence of plasma, are under investigation. The electrical insulation performances of ceramic coating on a copper samples are investigated. Several deposition methods have been evaluated, including magnetron sputtering, atmospheric plasma spray and detonation gun spray, which differ in the adhesion to the substrate, compact-ness, porosity and mechanical strength of the deposited material. The last two methods have the advantage of being conducted in air at atmospheric pressure, without the aid of complex vacuum systems. In order to validate the process, an experimental apparatus was prepared in laboratory, aimed to reproduce the expected conditions at the shell gap. It consists of a vacuum chamber in which a helium plasma was generated by means of a hot tungsten filament and a DC power supply. A bias voltage was applied between a copper plate and a cylindrical electrode (Ø 4 mm). The plate side facing the electrode was covered with alumina. The two electrodes were floating and biased by a small capacitor bank (0.3÷2 ?F). The voltage on the electrodes was applied for 200ms, with a repetition rate of 1Hz. In this contribution, the experimental results, aiming to study the conditions for the arc formation in presence of weakly ionized plasma (ne ~ 1016 m-3), are presented. In particular, voltage pulses up to 2.7 kV were applied, with a background gas pressure between 10-3 and 10 mbar. Further-more, the electrodes were kept both in contact and spaced up to 5 mm. Furthermore, the results of similar experiments performed on a mini mock-up, simulating the RFX-mod2 overlapping shell, are presented.

RFXmod2 [1,2] (R=2.0 m, a=0.49 m), the upgraded version of the previous RFXmod fusion device, will be equipped with a new reflectometry system specifically designed for plasma position control purposes. The diagnostic system will consist of four bistatic ultrafast independent reflectometric units working in the K band (18-26 GHz) and installed in four different poloidal locations at the same toroidal angle: two on the equatorial plane (High Field Side / Low Field Side) and two at the vertical top/bottom ports. The main features of the system have been already introduced at the last ECPD conference in 2019 [3] and will be discussed in the light of the more recent modifications. In this contribution, we also present the advances in the antennae design and in the technical solutions investigated to integrate the subsystems with the RFXmod2 mechanical structure. Concerning the antennae, a COMSOL MultiPhysics model has been developed in order to optimize the design of the Hogg horn (to be installed in the RFXmod2 HFS) and the pyramidal horn (to be installed in LFS) antennae. Their radiative pattern has been studied and the relative performance evaluated with respect to the available physical space within the RFXmod2 structures. 3D additive manufacturing technique will be exploited for their production. Specific electrical insulation solutions, involving a ZrO paint coating, has been also proposed for the waveguides routing inside the vacuum chamber and the preliminary tests will be discussed. The first K band unit has been designed and will be operational for bench tests from mid 2021.

RFX-mod is an experimental fusion device, which contributes to the plasma physics studies both in Reversed Field Pinch (RFP) and tokamak configurations. Its high flexibility, due to an active magnetohydrodynamic (MHD) control system and the modular Coils Power Supply (CPS) system, allowed operating RFX-mod in a wide range of experimental conditions with a plasma current up to 2 MA. Experiments with such high plasma current allowed the study on new promising confinement regimes, dominated by a self-organization process with the generation of a helical structure in the plasma core. Presently RFX-mod is under a significant upgrade (RFX-mod2) to extend the operational scenarios increasing the proximity between the conductive shell and the plasma. This main modification of the magnetic front-end, together with other main improvements, are expected to increase the performance of the machine in both magnetic configurations. In the frame of the studies to exploit the new potential of RFX-mod2 achieving a higher plasma current and longer flat-top duration, a solution based on additional magnetic energy storage has already been proposed. The study of RFP physics at higher plasma current could be crucial to confirm positive trends, like electron temperature and persistence of Quasi Single Helicity states with the plasma current, and to explore and achieve improved confinement states. This paper presents an alternative reconfiguration of the poloidal power supply system of RFX-mod2 based on a combined resistor-capacitor energy transfer system. This system allows to store energy in capacitor banks during the first phase of the plasma current ramp-up and release it to the plasma when the magnetizing current change polarity, driving the plasma current over 2.5 MA. The proposed upgrade does not involve radical poloidal CPS modifications, maintaining the present converters number and ratings and remaining within the power limits of the main power transformers (300 MVA).

The RFX-mod Reversed Field Pinch device passive boundary is being improved: - Drastic reduction of resistivity of first shell surrounding the plasma; - Reduction of plasma-stabilizing conductor distance from b/a=1.11 to b/a=1.04. The RFX-mod core upgrades consist of: - Removal of Inconel vacuum vessel; - Modification of the stainless steel Support Structure to ensure Vacuum Tightness (VTSS); - Modification of the copper Passive Stabilizing Shell (PSS); - Installation of upgraded sensors inside the vacuum vessel. Initial main points of investigation in the new device are discussed.

Il progetto RFX-mod2 prevede una serie di modifiche sostanziali della macchina RFX-mod dedicata a sperimentazione nel campo della fusione nucleare controllata a confinamento magnetico. Parti funzionali della esistente macchina sperimentale, in particolare la camera a vuoto toroidale (diametro maggiore 5 m, diametro minore 1 m) e i componenti in essa inclusi, sono stati oggetto di una profonda riprogettazione e riconfigurazione [R1] [R2]. Scopo della presente specifica, sviluppata nell'ambito del progetto POR-FESR MIAIVO (Obiettivo M3.3 "Realizzazione di coatings di allumina su lotti significativi, atti a validare il processo e le caratteristiche del prodotto. Esecuzione delle relative prove di laboratorio"), è quello di fornire i requisiti e le prescrizioni di base per l'esecuzione di un deposito superficiale di allumina (Al2O3 nel seguito indicato come 'Alumina-coating') da applicare sui componenti in rame costituenti la "scocca stabilizzatrice" (schermo magnetico) che sarà posizionata all'interno della camera a vuoto della macchina RFX-mod2.

An upgrade of RFX-­mod toroidal machine assembly is under development, aimed at widening the explored operational scenarios both in RFP and Tokamak configuration. The main design driver for this machine upgrade, named RFX-mod2, is the enhancement of the ‘shell-plasma proximity’, which is expected to provide a significant improvement in the plasma magnetic confinement. In order to achieve this goal, the existing Passive Stabilizing Shell will be adapted in order to be enclosed in the new vacuum vessel. This reconfiguration of the internal components of the experiment requires the design of a new structure aimed at supporting both the shell and the plasma facing components. The paper describes the design aspects of the shell assembly, with particular focus on the thermo-mechanical FEM analyses performed to verify the design of the components at the expected operating conditions.

An upgrade of the RFX-mod experiment (RFX-mod2) is in progress, involving a major change and reconfiguration of the inner components of the machine assembly. One of the most critical components of the assembly is the Passive Stabilizing Shell (PSS), a 3 mm thick toroidal structure (Rmajor = 1995 mm and rminor = 513 mm) made of copper which must be kept as close as possible to the plasma, in order to provide the suitable stabilization of the RFP magnetic configuration. In the present machine configuration, the PSS is wrapped and fastened around the external surface of the vacuum vessel, whereas for RFX-mod2 the PSS will be enclosed within a new wider vacuum vessel to enhance the 'shell-plasma proximity'. For this reason a new supporting structure has been conceived, able to support the shell and the new first wall (an array of 72x28 graphite tiles covering the entire inner surface of the shell) with stringent electromagnetic, thermomechanical and vacuum operating conditions. The new supporting structure is composed of a frame of 72 rings made of a high-performance polymer (Torlon® PAI), which complies with the thermo-mechanical and electrical requirements, reinforced with suitable stainless steel stiffening ribs and fastening bands. Each ring is bolted to the PSS by means of 28 peculiar clamps which, on the internal surface of the shell, provide also the fastening of the first wall tiles. The paper will describe the following design aspects: numerical FEM analyses performed to verify the thermo-mechanical design of the composite system at the expected operating conditions; development of geometrical features for the optimization of the injection moulding process for the manufacture of the ring sectors; experimental tests, in particular dimensional inspections and mechanical strength tests, carried out on prototypes to assess the reliability of the design solution.

An upgrade of the RFX-mod experiment is in progress, involving a major change and reconfiguration of the inner components of the machine assembly. In particular, the present vacuum vessel will be replaced by the external stainless steel toroidal support structure, which will be significantly modified with the integration of approximately 150 vacuum-sealed ports, interfaced with existing machine sub-systems (diagnostics, pumping and fuelling), and the implementation of 2 vertical and 2 horizontal vacuum-tight electrically-insulated crossed joints, necessary to allow suitable penetration of electromagnetic fields within the plasma chamber. The design of such peculiar joints of the vessel required the development of several mock-ups in which different solutions and arrangement of materials have been tested. The final solution implies the use of high-performance polymers combined with acrylic based syntactic foam materials, which provide proper vacuum tightness (required leak rates < 1E-9 mbar·l/s) and electrical insulation (required dielectric strength > 1 kV/mm), with a suitable viscoelastic characteristic that guarantees the compensation of thermal deformations expected during operation. The paper will describe the numerical FEM analyses performed to verify the thermo-mechanical design of the system at the operating conditions and the experimental tests, in particular leak tests and outgassing tests, carried out on mock-ups to assess the reliability of the design solution.

RFX-mod is an experimental fusion device built to explore and study plasma physics, especially in Reversed Field Pinch (RFP) configuration. RFX-mod allowed performing experiments with a plasma current up to 2 MA and obtaining stationary self-organized helical plasmas. Considering the results achieved so far, a significant upgrade of the device (RFX-mod2) is presently being implemented. The main objective is to increase the shell-plasma proximity, by removing the vacuum vessel and transferring the vacuum barrier function to the toroidal support structure, with the aim to better explore the potentialities of the RFP configuration. This paper presents the feasibility study of a significant improvement of the plasma current and flat-top duration via a limited upgrade of the RFX Power Supply (PS) system, without overcoming the power limits of the main transformers (300 MVA) and maintaining the present thyristor converters number and ratings. The application of this concept to RFX-mod2 will be studied starting from the RFX-mod configuration, focusing on the needed modification of the present circuits and the desired improved performance. The increase of the maximum plasma current and flat-top duration, under the current and mechanical limits of the coils, can be reached using additional magnetic energy storage and a complete revision of the configuration and operating principle of the power supply system of the poloidal field coils. With this new operational concept, during the charging phase, the magnetizing winding is charged along with an additional inductor, which acts as an energy storage reservoir. The stored energy is transferred to the magnetizing winding, and thus to the plasma, via transfer resistors in the last phase of the plasma current ramp-up, to reach values of plasma current up to 2.5 MA.

An upgrade of the RFX-mod experiment is in progress, involving a major change and reconfiguration of the inner components of the machine assembly. In particular, the present vacuum vessel will be replaced by the external stainless steel toroidal support structure, which will be significantly modified with the integration of approximately 150 vacuum-sealed ports, interfaced with existing machine sub-systems (diagnostics, pumping and fuelling), and the implementation of 2 vertical and 2 horizontal vacuum-tight electrically-insulated crossed joints, necessary to allow suitable penetration of electromagnetic fields within the plasma chamber. The design of such peculiar joints of the vessel required the development of several mock-ups in which different solutions and arrangement of materials have been tested. The final solution implies the use of high-performance polymers combined with acrylic based syntactic foam materials, which provide proper vacuum tightness (required leak rates < 1E-9 mbar·l/s) and electrical insulation (required dielectric strength > 1 kV/mm), with suitable viscoelastic characteristic that guarantees the compensation of thermal deformations expected during operation. The paper will describe the numerical FEM analyses performed to verify thethermo-mechanical design of the system at the operating conditions and the experimental tests, in particular leak tests and outgassing tests, carried out on mock-ups to assess the reliability of the design solution.

The edge of magnetically confined plasmas in toroidal configurations is characterized by the presence of various magnetic perturbations (MPs), appearing spontaneously as tearing modes in the Reversed Field Pinch (RFP) [1] or as peeling ballooning modes (ELM) in the tokamak. In the RFX-mod device during high-current discharges (R=2m, a=0.46m, IP>1MA) an almost monochromatic tearing mode (TM) spectrum spontaneously develops: this is the so-called quasisingle helicity (QSH) [2, 3], characterized by the presence of a single mode with helicity m/n, with (m=1, n=7) the poloidal and toroidal mode numbers respectively. However, the presence of secondary modes (m=1,n>7), with amplitudes one order of magnitude smaller than the dominant one, results in a local pattern of constructive interference (phase locking) and in a radial displacement of the plasma edge surface [4]. The intensity of the deformation can be comparable to that of the dominant mode, appearing as a sharp decrease ("hole") of the connection length to the wall at the locking angle, as shown by simulations with the ORBIT code [5,6]. An upgrade of RFX-mod device, RFX-mod2 [7], will be assembled in the near future. It will be characterized by a copper shell as continuous conductor nearest to the plasma and by a shellplasma proximity reduction from b/a=1.11 to b/a=1.04, likely improving feedback coils action. 3D MHD non-linear visco-resistive simulations show that secondary TM amplitude and the edge deformation due to phase locking will decrease by a factor 2 [8, 9]. Simulations with ORBIT show that in RFX-mod2 the average parallel connection length to the wall is expected to increase by a factor 8 with respect to RFX-mod, with no "hole" at the locking angle [6]. Having virtually cancelled the effect of TMs at r=a with a front end which behaves like an ideal wall, plasma wall interaction in RFX-mod2 could arise only due to the residual error fields at the gaps [8]: these upgrades are expected to lead to an optimized edge transport, with a well-formed SOL and to an improvement of the global plasma performance.

The Reversed Field Pinch (RFP) is a low magnetic field configuration for the confinement of thermonuclear plasmas. It exploits the pinch effect due to a current flowing in a plasma embedded in the toroidal magnetic field, which is one order of magnitude smaller than in a tokamak. It thus represents a non-disruptive, ohmically heated approach to magnetic fusion, based on self-organization and technological simplicity. This talk presents recent advancements on the RFP physics obtained in the RFX-mod device, and the perspectives of the configuration as a viable alternative to magnetic fusion. RFX-mod allowed to perform experiments in regimes with plasma current levels up to 2 MA, thanks to its MHD active control system [1]. Experiments have shown that improved confinement performances are obtained when in the resonant part of the m=1 spectrum one dominant Tearing Mode is much higher than the other secondary ones (Quasi Single Helicity states) [2]. Tearing Modes play a crucial role in determining energy and particle transport. The interesting behavior of fast particles in the RFP will be shown to be connected to the spontaneous enhancement of neutron yields influenced by the MHD activity of the RFP plasmas. A positive isotope effect has been found in helical RFP plasmas, as well as the natural tendency of strong outward impurity convection [1]. Based on the present understanding of the interplay between Tearing modes and passive conductive boundaries in an RFP, an upgrade of RFX-mod machine has been designed, dubbed RFX-mod2 [3], now under implementation. The highly resistive Inconel vacuum vessel will be removed, first wall graphite tiles will be attached to the copper stabilizing shell and the stainless steel support structure will be modified in order to be vacuum tight. In RFX-mod2, the shell-plasma proximity will decrease from b/a=1.11 to b/a=1.04 and copper will be the continuous conducting structure nearest to the plasma, instead of Inconel. MHD non-linear simulations show that secondary Tearing Modes amplitude and the edge bulging due to their phase locking will be reduced; moreover, the plasma current threshold for Tearing Modes wall locking will also significantly increase with respect to that observed in RFX-mod [3].

We will review key results from the 3D nonlinear MHD numerical modeling of Reversed-Field Pinch helical self-organization processes. Magnetic transport-barrier formation and nearly periodic reconnection events are found to be at play [1-4], akin to the experimental observation of thermal transport barriers and residual "back-transition" cycles when approaching helical regimes in high current discharges [4-7]. Similarities with Tokamak visco-resistive MHD snake/sawtooth-like phenomena will be discussed, as well as aspects at odds with the original Taylor's relaxation theory for the Reversed Field Pinch [8]. We will describe the recent successful technique to "channel" the system towards chosen "stimulated" macroscopic helical shapes by applying suitable (either Resonant or Non-Resonant) Magnetic Perturbations at the edge of the plasma, as predicted by nonlinear MHD modeling and observed in recent RFX-mod experiments [2]. In so doing, we are able to modify the transport properties of the configuration, with the two-fold objective of developing "handles" for the understanding of transport barrier formation processes and exploring new routes for optimization of pinch configurations. We have found that the magnetic chaos healing effect by helical structure development [9] appears to be more robust in the case of Non-Resonant helical regimes [2]. This line of research will be further explored in the upgraded RFX-mod2-device in Padova-Italy, expected to start operation in 2021.

In RFX-mod2, an experiment for the magnetic confinement of fusion relevant plasmas presently under construction, the in-vessel conductive plasma facing components are expected to be subjected during transient plasma current phases (start-up and termination) to significant electric fields, in the kV/mm range. While such electric fields are of no concern for components in vacuum, the presence of the scrape-off plasma (electron density ne 1E16 ÷ 1E18 m-3, electron temperature Te of few eV) can create the appropriate conditions for dangerous arc formation. For this reason part of the plasma facing components require a proper conditioning technique capable of maintaining the insulation between conductive components even in presence of the scrape-off plasma. To this purpose an experimental apparatus has been developed where to test the conditions for the arc formation and prevention between two electrodes immersed in a plasma generated by a hot filament. The results of an extensive experimental campaign will be discussed, aimed at demonstrating the possibility of gaining a good electrical conditioning by applying the standard conditioning technique usually employed for higher voltage ranges. It consists of a repetitive sequence of pulses with voltage applied to the pair of electrodes with current limitation, in the presence of the background plasma. The voltage is then slightly increased when arcing ceases, until the final desired voltage level is achieved (2.5 kV). In order to validate the procedure, different electrode materials have been tested in a variety of plasma conditions in terms of electron density and working gas pressure.

The reversed-field pinch (RFP) is a configuration for the magnetic confinement of fusion plasmas, in which most of the toroidal field is generated by the plasma itself through a self-organized dynamo process, instead of being produced by external coils as in the tokamak. In the RFP, the nonlinear saturation of resistive-kink/tearing modes brings to the spontaneous emergence of helical states with improved confinement. This is observed both in nonlinear magnetohydrodynamics (MHD) modelling [1] and in RFP devices, especially at high current [2,3]. A major advance in the predictive capability of nonlinear MHD modelling for RFP plasmas was made possible by allowing helical perturbations of the radial magnetic field at the plasma boundary, which was suggested by analytical study of helical equilibrium equations [4]. A proper use of helical magnetic perturbations (MPs) in MHD modelling allowed to obtain experimental-like helical states [5] and to predict new helical states with chosen helical twist, successfully produced in RFX-mod [6]. The amplitude of the helical component and sawtooth frequency can be "tuned" as well, thus opening new routes to configuration optimization in several respects (ion heating, thermal and fast particle transport processes). Here, we discuss the effect of a more realistic magnetic boundary recently included in the modelling: a thin resistive shell and a vacuum layer between the plasma and the ideal shell are considered, similarly as in [7,8]. We present two main results: 1) The decrease of secondary modes by increased shell-plasma proximity (see Figure 1). This is of interest in view of the upgraded RFX-mod2 device (starting operation in 2021), in which the shell-plasma proximity will change from b/a=1.11 to b/a=1.04 [9]. 2) With a proper choice for the resistivity of the thin shell at the plasma boundary, helical states do emerge in a self-consistent way, as in the experiment, without the need to impose a fixed helical MP (see Figure 2). Finally, further extensions of the realistic boundary implementation, in order to take into account a double resistive shell and an active feedback control system, will be discussed.

RFX-mod device has been operated as circular and shaped tokamak. An upgrade of the machine assembly is now being implemented, dubbed RFX-mod2. The vessel will be removed, graphite tiles will be attached to the stabilizing shell and the support structure will be vacuum tight. Thus, shell-plasma radius ratio decreases and the shell will be the conducting structure nearest to the plasma. An improved diagnostic system would allow a better control of both plasma shape and H-mode transition. We present a feasibility study on new operational scenarios with the aim of increasing plasma performances by acting on the plasma shape. Thus, new D-shaped plasma equilibria with DEMO-like plasma shape parameters (i.e. ?\textgreater 0.3, ?\textgreater 1.5) are investigated with the IET (Inverse Equilibrium Tool) code, including the option of single/multiple X-points. In the same way, we investigated the possibility of reversing the triangularity of such plasmas. Thanks to its engineering flexibility, RFX-mod2 would allow to explore a wide window of shape parameters with low requirements on active coil currents.

Clear evidences that, due to a strong outward impurity convection, impurity core penetration is prevented have been found in the RFX-mod RFP device. A comparable convection of the main gas has not been observed [1] so that a favorable situation with peaked or flat density profiles and hollow impurity profiles is produced. Analysis of impurity transport relies on best reconstruction of impurity emission pattern with a 1-dim Collisional-Radiative code in which the radial impurity flux is schematized as a sum of a convective and a diffusive term [2,3]. The diffusion coefficient D and the velocity V, which are input to the simulation are varied until the experimental emission is reproduced. While the steady-state impurity profile is determined by the ratio V/D (peaking factor) , the discrimination between D and V requires transient perturbative experiments. The experimental evidence of impurity outward convection in RFX-mod helical regimes occurring at high plasma current (I>1.2 MA) has been found in Li and C solid room temperature pellets experiments [4], Ne doped D2 cryogenic pellet injection, Ne gas puffing and Ni LBO experiments [5](W LBO didn't show accumulation effects too). Similar D and V have been found for all the considered impurity species, without strong dependence on mass/charge. RFX-mod is now being upgraded to RFX-mod2, aiming at reducing secondary tearing mode amplitude which affects the duration of the improved confinement Single Helicity states [6]. In order to perform more detailed analysis of the impurity transport inside the outward convection barrier, the impurity source should be further inside the plasma. With this aim, Ni-tracer encapsulated solid pellet (Ni-TESPEL) experiments are foreseen in the new device [7]. The available 1-dimensional, time dependent Ni Collisional Radiative code, used to reconstruct experimental Ni emissions in RFX-mod [ 4] has been upgraded in preparation of such experiments in RFX-mod2 including the possibility of a Ni source (boundary condition) inside the plasma, placed in a time dependent position. The solid pellet injector already used in RFX-mod to inject C and Li solid pellets, will be adapted to inject TESPEL in RFX-mod2 (0.7/0.9 mm polystyrene ball with Ni powder inside, injection velocity up to 200 m/s can be reached). In this contribution, the solid pellet injector will be described, simulations of the pellet ablation [8] for different scenarios of RFX-mod2 plasma will be presented, Ni ion density, line and continuum emission profiles predicted by the code will be described and discussed.

The reversed-field pinch (RFP) is a configuration for the magnetic confinement of fusion plasmas, in which most of the toroidal field is generated by the plasma itself through a self-organized dynamo process, instead of being produced by external coils as in the tokamak. In the RFP, the nonlinear saturation of resistive-kink/tearing modes brings to the spontaneous emergence of helical states with improved confinement. This is observed both in nonlinear magnetohydrodynamics (MHD) modelling [1] and in RFP devices, especially at high current [2]. A major advance in the predictive capability of nonlinear MHD modelling for RFP plasmas was made possible by allowing helical perturbations of the radial magnetic field at the plasma boundary [3]. A proper use of helical magnetic perturbations (MPs) in MHD modelling allowed to obtain experimental-like helical states [4] and to predict new helical states with chosen helical twist, successfully produced in RFX-mod [5].Here, we describe a further refinement of the magnetic boundary modelling. Instead of applying fixed helical MPs, we study the helical self-organization in the presence of a thin resistive shell and a vacuum layer between the plasma and the ideal shell. Two main results are discussed. On the one hand, by varying the distance between the plasma and the ideal wall it is possible to provide a nonlinear estimate for the decrease of secondary modes by increased shell-plasma proximity. This is of interest in view of the upgraded RFX-mod2 device (starting operation in 2020), in which the shell-plasma proximity will change from b/a=1.11 to b/a=1.04 [6]. On the other hand, it is observed that with a proper choice for the resistivity of the conducting shell at the plasma boundary, experimental-like helical states do emerge in a self-consistent way, as in the experiment, without the need to impose a fixed helical MP. Finally, further extensions of the realistic boundary implementation, in order to take into account a double resistive shell and a feedback control system, will be discussed.

Collisional particle transport in the presence of field perturbations originating from various MHD activity is examined theoretically on tokamaks (ITER, ASDEX Upgrade, NSTX and DIIID) and the reversed-field pinch RFX-mod [1]. For ITER and ASDEX Upgrade, modes typically leading to a disruption [2] are considered. On NSTX and DIII-D unstable Alfvén modes are investigated. Finally on RFX-mode the effect of saturated tearing modes is studied. The existence of subdiffusive transport [3] for electrons is found to occur in some cases at very low mode amplitudes. Subdiffusion is also found for ions of high energy. In fact, orbit resonances can produce long time correlations and dynamical traps [4] for particle trajectories at perturbation amplitudes much too small for the orbits to be represented as uniformly chaotic. Besides this, in all devices orbits show a high degree of anisotropy, especially when comparing the angular (toroidal and poloidal) and radial directions. As a consequence, in the presence of field perturbations produced by MHD modes, the use of a traditional diffusive-convective scheme for transport, which is expressed by the Fick's law~? = -D~?n+~v n, leading to the well known transport scalings, is questionable. The existence and nature of subdiffusive transport is difficult to determine from first-principle theories, since it is found to depend on the nature of the mode spectrum and frequency, as well as on the mode amplitudes: this fact is mirrored in the different value of the Kubo number found in the devices analyzed in this paper. The connection between subdiffusive transport, Kubo and nonlocal models of transport [5] is also discussed.

The reversed-field pinch (RFP) is a configuration for the magnetic confinement of fusion plasmas, in which most of the toroidal field is generated by the plasma itself through a self-organized dynamo process, instead of being produced by external coils as in the tokamak. In the RFP, the nonlinear saturation of resistive-kink/tearing modes brings to the spontaneous emergence of helical states with improved confinement. This is observed both in nonlinear magnetohydrodynamics (MHD) modelling [1] and in RFP experiments, especially at high current [2,3]. A major advance in the predictive capability of nonlinear MHD modelling for RFP plasmas was made possible by allowing helical perturbations of the radial magnetic field at the plasma boundary, assuggested by analytical calculations based on helical equilibrium equations[4]. A proper use of helical magnetic perturbations (MPs) in MHD modelling allowed to obtain experimental-like helical states [5] and to predict new helical states with chosen helical twist, successfully produced in RFX-mod [6]. Here, we describe a further refinement of the magnetic boundary modelling. We study the helical self-organization in the presence of a thin resistive shell at the plasma boundary r=a, surrounded by a vacuum layer and an ideal shell at r=b. The new magnetic boundary is implemented in the SpeCyl code [7] in a similar way as in Refs. [8,9]. Two main results are discussed. On the one hand, by varying the distance between the plasma and the ideal shell it is possible to provide a nonlinear estimate for the decrease of secondary modes by increased shell proximity to the plasma. This is of interest in view of the upgraded RFX-mod2 device (starting operation in 2021), in which the shell proximity will change from b/a=1.11 to b/a=1.04 [10]. Based on nonlinear MHD modelling,a factor of 2 reduction of the edge radial magnetic field is expected going from RFX-mod to RFX-mod2, with the beneficial consequence of a milder plasma-wall interaction. On the other hand, it is observed that with a proper choice for the resistive diffusion time of the thin shell at r=a, helical states do emerge in a spontaneous and systematic way, as in the experiment, without the need to impose a fixed helical MP. Finally, further extensions of the realistic boundary implementation, in order to take into account a double resistive shell and a feedback control system, will be discussed.

The RFX-mod2 Nuclear Fusion experiment is an upgrade of RFX-mod. Among the other improvements in machine structure and diagnostics, a larger number of electromagnetic probes (EMs) is foreseen to provide more information about plasma instabilities and to allow an improved real-time plasma control. An Analog to Digital Converter (ADC) architecture able to provide both transient recording and real-time streaming is foreseen in RFX_mod2. Transient recording provides full speed data acquisition (up to 1 MSample/s) by recording data in local memory and reading memory content after the plasma discharge. Real-time streaming of subsampled data is required for active control. The chosen technology is based on the XILINX Zynq architecture that provides in the same chip a multicore ARM processor tightly couple to a FPGA. Time critical functions carried out by the FPGA in this context are: 1) The management of a circular data buffer and the DMA transfer in RAM of pre and post trigger samples after the trigger has been received; 2) Antialiasing filtering and subsampling of the samples to be streamed. The resulting samples are enqueued in a FIFO accessed by the processor. The functions carried out by the processor are: 1) The management of the configuration setting, received via TCP/IP or HTTP. The processor validates the configuration and write the appreciate registers in the FPGA; 2) Offline data readout of acquired samples in transient recording; 3) Network data streaming of subsampled data read from the FIFO and sent in UDP packets to the active plasma control system.

A number of modification s and enhancements are underway on the RFX - mod Reversed field Pinch (RFP) device . The main scientific goals motivating the modifications are the improvement of the confinement in the R eversed F i e l d P inch configuration and the investigation of a broad spectrum of plasma physics topics , through the exploitation of the multi configuration capability of the device, which can be operated as a RFP, ultra low q, circular and shaped tokamak . This paper describes the major challenges tackled to design technical solutions able to achieve this scientific mission.

In recent years the RFX-mod device has been operate d as a Reversed Field Pinch (RFP) and as a low-q Tokamak. In both configurations the capability and flexibility of its active control system of MHD instabilities have been exploited to control RWMs and tearing mode wall locking (RFP) and to stabilize the m/n=2/1 mode to operate at q(a)

A conceptual design of an upgrade for RFX-mod was described at the past IAEA-FEC conference and the device is presently being modified. Further analysis of the RFX-mod database, concerning m=1 Tearing Modes locking threshold and reconnection events characterization are presented. Estimates of nonlinear behavior of Tearing Modes in RFX-mod2 together with detailed electromagnetic modeling aiming at optimizing error fields for the upgraded device have been performed in order to consolidate the design.

The design of a major refurbishment of the toroidal complex of the RFX-mod experiment is going to be finalized before starting the realization phase. The Inconel vacuum vessel will be removed and the stainless steel supporting structure will be modified so as to become vacuum tight. The plasma facing graphite tiles will be mounted onto the inner surface of the copper shell, allowing an increase of the plasma proximity factor. This could allow different operation regimes expected to provide a significant reduction of the amplitude of RFP tearing modes. This reduction would lead to magnetic chaos mitigation with confinement improvement, better mode control capability and reduced plasma wall interaction. On the other hand, due to the shorter distance from the passive structures, the plasma will be even more sensitive to magnetic field errors produced by the induced currents at the cuts of the same structures. A careful assessment of different gap configuration is mandatory to optimize the final design of the machine modifications. Numerical analyses confirmed by the following experimental results showed that overlapping layers at the shell poloidal gap were an effective solution to minimize error fields in RFX-mod. This suggested analysing the possibility of a similar configuration at the shell inner equatorial gap, too. The direct reproduction is not straight-forward due to strict dimensional constraints and the required compatibility with a much more complex assembly procedure necessary to make the support structure vacuum tight. Parametric analyses of different constructive solutions have been made possible by a computational tool specialized to estimate the induced currents in thin conducting structures with complex geometry and the associated magnetic fields. The magnetic error fields due to other new features of the design such as the conducting cage introduced to assure the electrical equipotentiality of the graphite tiles have also been investigated.

Status of RFX-mod and RFX-mod2

RFX-mod is a Reversed Field Pinch device that allowed performing experiments in regimes with a plasma current up to 2 MA. Due to its low value of the safety factor (q<