Search

An accurate control of the magnetic boundary ofa thermonuclear plasma is an important issue in magnetic confinement research. The development of methods for the active control of magnetohydrodynamic instabilities and correction oferror fields is mandatory. A substantial modification of the layoutof the RFX-mod device is in progress: the aim of this paper isto assess the potential advantages of the proposed solution with advanced numerical techniques based on integral formulations.

Carbon-free energy sources are essential to avoid global warming, and nuclear fusion is expected to play a major role in achieving clean and sustainable energy. In the development of magnetic fusion, there exist a lot of issues related with plasma equilibrium to be studied extensively, typical representative of which is a problem of equilibrium reconstruction of plasma. In this paper we propose a new method of equilibrium reconstruction for fusion plasma based on the data assimilation. Aiming at dealing not only with axisymmetric toroidal plasmas but also with more general toroidal plasmas, we formulate the problems of equilibrium reconstruction in generalized forms and derive methods to solve them. We also propose a method for applying the equilibrium reconstruction method to the reversed field pinch (RFP), and it is applied to a real RFP experimental apparatus for its evaluation. It is shown through numerical experiments that the proposed equilibrium reconstruction method works well with reasonable accuracy. And it is also shown through real experiments applying the RFP apparatus that the equilibrium of RFP plasma is appropriately reconstructed.

We study the influence of the shape of the plasma container on the dynamics of the reversed-field pinch (RFP). The geometries we consider are periodic cylinders with elliptical and circular-shaped cross-sections. Numerical simulations of fully nonlinear viscoresistive magnetohydrodynamics are carried out to illustrate how the plasma dynamics is affected by shaping. It is shown that independent of the plasma shape, the quantity $\beta$ , comparing the hydrodynamic pressure to the magnetic pressure, decreases for increasing values of the Lundquist number, but the pressure gradient fluctuations remain roughly constant, when compared to the Lorentz force. Different elliptical shapes of the cross-section of the domain lead to the excitation of different toroidal (or axial) magnetic and dynamic modes. Furthermore, it is found that in a geometry with circular cross-section, a significant local poloidal angular momentum is observed, absent in the geometries with elliptical cross-section. Because the confinement is dominantly determined by plasma movement, and the dynamics of the velocity and magnetic field are modified by the modification of the geometry, shaping can thus affect the performance of RFP devices.

Electron density fluctuation has been measured based on the collect forward scattering method using a single-chord microwave interferometer system on the Keda Torus eXperiment (KTX) reversed field pinch. The interferometer system is a heterodyne one with two terahertz solid-state sources with frequency around 650 GHz, and the transmitted beam and forward scattering beam are received simultaneously by a Schottky planar diode mixer with sensitivity about 500 mV/mW. Constrained by the beamwidth and geometry limitation, the measurable wave vector range is $k_{w}\le 5 {\mathrm {cm}}^{-1}$ with a resolution $\Delta k\approx 2.07\,{\mathrm {cm}}^{-1}$ . MHD activities are detected via this forward scattering technique. The results are consistent with other diagnostics, such as the chord averaged electron density, the edge magnetic probe, and high-speed charge-coupled device (CCD) camera.

Analysis of Lagrangian Coherent Structures (LCSs) has been shown to be a valid mathematical approach to explain the formation of transport barriers in magnetized plasmas. Such LCSs, borrowed from fluid dynamics theory, can be considered as the hidden skeleton of the system and can be used for studying a wide spectrum of transport mechanisms, even in plasmas. An LCS can be considered as a generalization for a finite time interval and for a general dynamical system of what manifolds are for Hamiltonian systems. In this paper, we demonstrate that such structures can be particularly useful for underlying the hidden paths governing the motion of magnetic field lines in chaotic magnetic fields. To perform such an analysis, we developed a numerical tool able to detect LCSs for a general dynamical system. The tool is able to deal with general coordinate systems and it is shown to match with other techniques already used to analyse chaotic magnetic fields, e.g. connection length. After the description of the computational tool, we focus on the heat transport equation and the comparison between temperature profile and topology of the LCSs. We provide evidence that numerical simulations are able to reproduce the temperature profiles similar to those observed in reversed field pinch experiments and that our tool successfully predicts the location of temperature gradients. The results suggest that, inside the chaotic region, the field-lines motion is far from stochastic and that the presence of hidden patterns allows the development of high temperature gradients.

Significant advances have been achieved in recent years in Reversed Field Pinch (RFP) physics, both from the experimental and the theoretical point of view. The main theoretical advances are based on fully 3D nonlinear visco-resistive MHD simulations, which gained a deeper understanding of the RFP's magnetic self-organization. Improved confinement regimes have been explored with two different approaches in the RFP. The first is associated with the spontaneous topological bifurcation from a cylindrical to a helical magnetic equilibrium. Such equilibria are favored by high plasma current levels and are associated with the emergence of partially conserved magnetic flux surfaces, resulting in the onset of a transport barrier. A fundamental role in determining the improvement of the experimental results is played by advanced active feedback control systems for the control of MHD instabilities. The second approach is based on active control of the current profile, which can induce significant enhancements of the confinement properties. High confinement of fast particles has been also observed in RFP plasmas.

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. This paper presents the feasibility study of a significant increase of the plasma current and flat-top duration via a limited upgrade of the RFX power supply system, without overstressing the poloidal windings or overcoming the power limits of the main step-down transformers (300 MVA) and maintaining the present set of thyristor converters units. 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, in the order of 100 MJ, 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 2.6 MA.

The Greenwald density limit is a phenomenological limit that restricts the achievable density in a thermonuclear fusion device. This is a longstanding issue for research in this field, since density appears in the figure of merit defining the efficiency of the future fusion reactor, namely the triple product nrT . Despite this importance, the density limit defies a unified, convincing theoretical explanation. Recently a series of papers [1,2,3] point out the role of a combination of radiative collapse and magnetohydrodynamic (MHD) instabilities resonating in the edge plasma, in determining the limit.

An improvement of the boundary conditions scheme of the 3D nonlinear MHD numerical code SpeCyl is presented. Boundary conditions have been shown to play a key role in the helical selforganization both in Reversed Field Pinch and tokamak plasmas. Two different sets of boundary conditions have been extensively tested against ubiquitous relaxation phenomena induced by plasma current in toroidal devices: ideal kinks and tearing modes. The role of wall position and resistivity on linear perturbations profiles and their exponential growth rates was tested, motivating the need for a reformulation of fluid boundary conditions as well. Preliminary results of such new boundary conditions are also presented, along with a summary of the relevant theoretical framework underlying linear MHD instabilities.

Fusion–fission hybrid reactors are seen as a possible, mid-term, CO2-free energy source. Starting from the progress in understanding the reversed field pinch plasma confinement, a pilot neutron source with a configuration of R = 4 m and a = 0.8 m and a plasma current of up to 14 mega-ampere (MA) is proposed. A staged experimental approach, with increased complexity and investment, was identified to tackle the existing issues related to scientific and technological aspects and to test the D-T operation at reduced fusion power (P_fus $\approx ~30$ MW, Q $\approx ~0.4$ , continuous pulsed operation).

Fusion-Fission Hybrid Reactor (FFHR), neutron-producing fusion core surrounded by a fission blanket, benefits from a renovated interest because of (i) its potentiality to address energy production before the availability of pure fusion reactors, (ii) its capability of fission fuel fertilization and (iii) waste management. The core fusion reactor of a FFHR requires less stringent performances respect to DEMO or the fusion power plants: it has to produce less than a hundred of MW of fusion power with Q = 1-5 and comes with a reduced fast neutron wall load, sustainable by the available nuclear material. While the Reversed Field Pinch (RFP) has yet an energy confinement too low for a high Q reactor, it is an alternative configuration to tokamak and stellarator for this application. Its disruption-free configuration, the need of a much weaker toroidal magnetic field coils and the capability of reaching thermonuclear temperatures without additional heating, but only by ohmic heating (due to no intrinsic plasma current limit and the strong winding of current lines due to the low q values), are significant advantages. New studies are in progress in order to revisit the status and the potentiality of the RFP as a fusion core in FFHR, taking into account the recent progress and expected improvement in RFP physics brought by the results of the RFX-mod experiment (R=2, a=0.46, Ip=2MA) and its underway upgrade. Scaling from the obtained experimental results opens the possibility, with increased machine size and plasma current, to realize a more robust and simple fusion neutron source respect to the other considered solutions; starting from these result, the possibility to realize a preliminary pilot experiment with a RFP (R=4m, a=0.8m, Ip=12-14MA) as a simplified neutron fusion source is in progress.

The Multi-region Relaxed MHD (MRxMHD) has been successful in the construction of equilibria in three-dimensional (3D) configurations. In MRxMHD, the plasma is sliced into sub-volumes separated by ideal interfaces, each undergoing relaxation, allowing the formation of islands and chaos. The resulting equilibrium has a stepped pressure profile across sub-volumes. The Stepped Pressure Equilibrium Code (SPEC) [S.R. Hudson et al., Phys. Plasmas 19, 112502 (2012)] was developed to calculate MRxMHD equilibria numerically. In this work, we have extended the SPEC code to compute MRxMHD equilibria with field-aligned flow and rotation, following the theoretical development to incorporate cross-helicity and angular momentum constraints. The code has been verified for convergence and compared to a Grad-Shafranov solver in 2D. We apply our new tool to study the flow profile change before and after the sawtooth crash of a reversed-field pinch discharge, in which data of the parallel flow is available. We find the promising result that under the constraints of cross-helicity and angular momentum, the parallel flow profile in post-crash SPEC equilibrium is flat in the plasma core and the amplitude of the flow matches experimental observations. Finally, we provide an example equilibrium with a 3D helical field structure as the favoured lower energy state. This will be the first 3D numerical equilibrium in which the flow effects are self-consistently calculated.

The phenomenology of reconnection events, associated to relaxations in high-current (similar to 1.5 MA) plasmas of the reversed-field pinch device RFX-mod, is shown. Each relaxation event can be described as a series of stages starting in the core and propagating towards the plasma edge. In an initial stage (trigger), the evolution of the q profile brings resonant layers closer together, allowing for an interaction of the current sheets associated to tearing modes at the respective rational values of q. The phase locking of the resonant modes then initiates the reconnection process that, once started, changes the magnetic topology bringing the initially helical state to a more chaotic configuration.

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 (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.

We will briefly recall the features of the basic reconnection process within viscoresistive 2D nonlinear numerical MHD approximation for both the Reversed Field Pinch (RFP) and Tokamak pinch configurations. After the current sheet formation, plasmoid formation is observed [Fig.1(a,b)], when proper numerical resolution is used, which speeds up the process. Then, the full three dimensional reconnection processes will be revisited for the experimental-like RFP quasi helical regimes [1, 2]. The 3D RFP simulations show much larger and abrupt magnetic energy conversion into kinetic one, with respect to the 2D reconnection processes, overriding the fine structures observed in 2D approximation. Relaxation-reconnection events cyclically interrupt the formation of quasi helical regime, featuring 3D current sheets formation (Fig.2), mode-mode- phase locking, excitation of Alfvèn waves (Fig.3) [3, 4]. The process reflects in several respects the features highlighted in the past, including typical scaling laws, within the fully developed 3D MHD turbulence obtained in low collisional ideal boundary conditions [5, 6]. Such reconnection processes have been envisaged to provide direct ion heating in the RFP [7]. We will describe how we manage to suitably pace the relaxation cycle in simulation and experiments [8].

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.

The Keda Torus eXperiment (KTX) is a new medium-sized reversed-field pinch device designed and implemented jointly by the University of Science and Technology of China and the Institute of Plasma Physics at the Chinese Academy of Science. An important component of the KTX project is a central control system that is composed of six modules: 1) central console; 2) process control system; 3) central timing system; 4) safety and interlock system; 5) gas injection system; and 6) shot information server. The network and optical fiber infrastructures also provide great support for the stable operation of the experiment. This paper will describe the details about the system architecture and networks deployment.

KTX, which stands for “Keda Torus for Experiment”, is a reverse field pinch (RFP) device, with main parameters between RFX and MST. It is supported by China domestic ITER-related research project. The target of KTX is to do the research work of “self-organization”of magnetic confinement plasma. It will include the dynamo and single spiral phenomena. Its magnet system includes the Toroidal Field (TF) coils, Ohmic Heating (OH) coils and Equilibrium field (EF) coils. The toroidal field coils should produce the magnetic field of 0.7 T in center of plasma for the 1 MA operation mode, the current of each TF coil is designed as 210 kA. The power supplies and OH & PF coils are divided into three groups to reduce the voltages on each capacitor bank. Then flux swing is provided by the coils from OH 1 to OH 7, while the stray field in the plasma region is minimized by the coils from OH 8 to OH 13. The total ampere turns of OH coils is optimized as 11.6 MAT with 228 turns in total. The magnet system design is one of the major activities for KTX device design. The key engineering features and design including electromagnetic parameters and related R&D work are presented in this paper.

The knowledge of edge plasma transport parameters and plasma edge phenomena is a key element in the design of the first wall for a magnetically confined fusion experiment. In RFX-mod heat flux measurement and edge transport modelling have been done to improve the understanding of this aspect. Heat flux deposition profiles have been evaluated from infrared temperature measurements of insertable graphite limiters. They were inserted up to 12 mm into the reversed field pinch plasma of ohmically heated discharges with Ip= 0.6÷1.0 MA, ne= 0.5÷3·1019 m-3 (n/nG< 0.7) and total power of about 10÷15 MW. Strong asymmetries in heat flux deposition have been measured in poloidal direction at low density between the electron and the ion drift side and smaller ones in toroidal direction when q(a)=0. The poloidal asymmetry has been associated to the presence of superthermal electrons [1] while the toroidal one has been less clearly identified as due to the small toroidal extension of the limiters. To account for the 2D deposition nature of heat load on the surface of the employed limiters, a simple 3D code has been developed to evaluate heat flux from temperature data. In this way at the deeper limiter insertions a heat flux decay length of about 2 mm and 2.5 mm has been evaluated in electron and ion drift sides. Modelling of the evaluated heat fluxes has been done using the SOLEDGE2D-EIRENE edge code [2]. This fluid code is well suited for the RFX-mod wall limiter configuration because, thanks to the implemented penalization technique, the computational domain can be extended up to the entire first wall. Edge modelling has shown that measured decay lengths are compatible with energy diffusion coefficients in Scrape Off Layer (SOL) smaller than those commonly evaluated at plasma edge; the cause of the reduced diffusion in the SOL will be discussed in the paper.

A series of techniques is presented that has been developed to optimize the output magnetic field of the feedback control system on the RFX-mod reversed field pinch device. With the aim of minimizing the harmonic distortion and correcting localized error fields, these methods should be lightweight for real-Time application and effective in improving the performance of a system that is routinely used for active control of magnetohydrodynamic plasma instabilities. The implementation of simple, linear algebra based, real-Time optimization methods will be described along with proof of the beneficial effects sought. The focus of this paper is set on a spurious harmonics reduction technique based on the decoupling of sensors and actuators; a description of its derivation will be given together with the implementation in the control loop. A similar procedure for the compensation of faulted actuators will also be mentioned.

An equilibrium field (EQ) power supply (PS) circuit using a feedback H-bridge pulse-width-modulation controller has been developed to control the plasma horizontal position and to satisfy the requi...

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].

The Reversed Field Pinch (RFP) configuration is often characterized by a wide spectrum of unstable tearing modes (TMs) involved in the generation and sustainment of the magnetic field in the plasma. This dynamo process is heavily influenced also by other processes or parameters like intrinsic plasma flow (no external torque sources are normally present), wall resistivity and interaction with external non axi-symmetric magnetic fields provided by active coils. The transition between fast-slow rotation branches under the application of magnetic feedback boundary conditions was studied in [1] for the case of RFX-mod, a medium size (a = 0.459 m, R0 = 2m) flexible toroidal magnetic confinement device. In particular, RFX-mod is equipped with an advanced and sophisticated feedback system realized by a grid of independently controlled 48 (toroidal) x4 (poloidal) active saddle coils. In this contribution the interaction of the m=1, n>9 tearing modes with the bulk plasma close to the fast-slow transition threshold is investigated with the help of a systematic set of experiments where the active feedback control is selectively switched on and off on specific modes during the discharge dynamics. The peculiar result found is that also in fast rotating discharges, although the radial component at the wall of high n tearing modes is almost zero, in the absence of active feedback control a gradual modification (growth) of the total mode amplitude is measured, leading potentially to wall locking of that mode and to the back transition to the slow rotation branch. Data showing the mode-mode and mode-wall interaction will be presented for several combinations of non-controlled TMs. The process is discussed also in function of the dynamics of the plasma flow, following the lines presented in [2], and showing the clear correlation between active control, MHD dynamics and plasma flow.

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.

The operation of RFX-mod, the largest Reversed Field Pinch (RFP) experiment, with plasma currents up to 2 MA demonstrated a substantial increase of the electron and ion temperatures versus the plasma current. The observed trend outlines a path toward a significant D-T fusion process yield in larger devices with higher plasma current [1]. Such a device, in which the plasma is purely ohmically heated, could act as an efficient, robust and cheap fusion neutron source with a neutron rate to be used as the basis for a Fusion-Fission Hybrid Reactor (FFHR). Its peculiar features can be summarized as: ohmically heated plasma, toroidal field winding rated for a very low magnetic field (two orders of magnitude lower than in tokamaks), robust and cheap construction, simple access for Remote Handling and easy maintenance. The neutron production, as required in a hybrid reactor, is guaranteed by a continuously pulsed operation without the need of additional current drive systems. A realistic concept consist of an inductively operated RFP with major radius R=6 m, minor radius a=1 m and a current of 20 MA. The expected plasma temperature would be in the range 10-15 keV, leading to a neutron production rate in the range of 1019 n/s and a wall neutron load of 0.2 MW/m2.

The Reversed Field Pinch (RFP) configuration looks to be an attractive option for fusion-fission hybrid reactors: the toroidal magnetic systems would be made of copper coils instead of more expensive superconductive magnets; fusion conditions could be reached by ohmic heating only, therefore additional heating systems would not be required; the fission blanket could be located in the most external part of the torus thus facilitating maintenance operations. The paper aims at assessing the potentialities, such as fuel fertilization and/or nuclear waste transmutation and electricity production of a hybrid reactor with a RFP fusion core (R = 6 m, a = 1) whose conceptual design and plasma performances are based on RFX-mod, the largest RFP experiment currently in operation. Fusion conditions can be reached by heating a D-T plasma up to 9.6 keV by ohmic heating, generated by a 20 MA plasma current induced and sustained by flux swing only. The neutron flux (2.1 × 1013 fast neutron/cm2/s) is used to breed tritium in both the inner and outer blanket sections and induce fission reactions in dedicated areas in the external blanket section where Pu + MA (60%)-Zr (40%) rods are located. Both neutronic and safety analyses corroborate the viability of a FFH reactor with a RFP core.

We report on the first comprehensive experimental and numerical study of fast ion transport in the helical reversed-field pinch (RFP). Classical orbit effects dominate the macroscopic confinement properties. The strongest effect arises from growth in the dominant fast ion guiding-center island, but substantial influence from remnant subdominant tearing modes also plays a critical role. At the formation of the helical RFP, neutron flux measurements indicate a drastic loss of fast ions at sufficient subdominant mode amplitudes. Simulations corroborate these measurements and suggest that subdominant tearing modes strongly limit fast ion behavior. Previous work details a sharp thermal transport barrier and suggests the helical RFP as an Ohmically heated fusion reactor candidate; the enhanced transport of fast ions reported here identifies a key challenge for this scheme, but a workable scenario is conceivable with low subdominant tearing mode amplitudes.

The design of the RFX-mod device upgrade requires a 3 mm copper shell close to the plasma to achieve improved plasma confinement properties. Although a continuous conductive structure would be preferred as a passive plasma stabilizer, such a conductive structure shall have an electrical discontinuity in both the poloidal and toroidal directions, in order to allow the penetration of electromagnetic fields into the plasma region. These gaps prevent the formation of net poloidal and toroidal currents. Moreover, the shell has been designed with an overlapping region at the poloidal gap in order to reduce the induced field errors. During Reversed Field Pinch operations, the loop voltage, externally induced to sustain the plasma current, can reach values up to 400 V. These values can rapidly increase up to 2 kV during fast plasma current terminations. Therefore, intense electric fields can develop between the shell flaps, only a few millimetres apart, along the overlapping region. Furthermore, taking into consideration that the copper shell is located inside the vacuum chamber and is therefore exposed to the low temperature plasma of the scrape off layer, the formation probability of harmful electric arcs is high. In order to avoid arc formation, different types of insulation coating on the copper, able to withstand the applied electric fields in the presence of plasma, are under investigation. The electrical insulation performances of an alumina coating on a copper substrate 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, compactness, 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 set up in the 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 the use of a hot 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 are 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 results of the experiments, with the aim of studying the conditions for the arc formation in presence of a weakly ionized plasma (ne ~ 10e16 m-3), are presented. In particular, voltage pulses up to 2.5 kV were applied, with a background gas pressure between 10e-3 and 10 mbar. Furthermore, the electrodes were kept both in contact and spaced up to 5 mm. The analysis of alumina degradation, after dielectric cracking in the presence of arc, is also presented.

Reconnection is a basic phenomenon in magnetized plasmas by which the plasma relaxes to a state oflower magnetic energy. It plays a fundamental role in a variety of instabilities common to most fusiondevices, as sawtooth or tearing modes. Often it occurs in impulsive events (crash phase), following alonger period in which the magnetic fields evolve slowly. Energy conversion and topological change are two highly linked aspects of reconnection.Hot fusion plasmas may contain multiple reconnection sites and this is particularly true for the Reversed Field Pinch (RFP), due to the peculiar current density distribution, associated to a variety ofresonant tearing modes. In addition, large RFPs involve large plasma currents and therefore a largeamount of available magnetic energy. This makes the RFP a good device to observe and analyze reconnection events.In this contribution the phenomenology of the impulsive and periodic reconnections in the RFX-mod device is analyzed in detail, through an ensemble average of the data from many reconnection eventsin high plasma current discharges (1.5 MA). Spatial and temporal scales are invoked as fundamental observables. Roughly speaking, an almost stationary helical state is considered typical of phases before the crash. The crash destroys this symmetry, and the plasma topology undergoes a transition to a more chaotic state with broad MHD spectra. The crash evolution is described as a sequence of events. A trigger of the crash has been associated for the first time to a slight change in the helical equilibrium, which pushes closer the inner most important resonant flux surfaces in the core. Reconnection starts when the resonant surfaces - and therefore the intense related current sheets- are close enough and a precursor of the crash is found in the locking in phase of these resonant modes. After these slowly evolving phases, fast reconnection takes place, which lasts about 1 ms.Plasma parameters and magnetic measurements suggest that crashes are related with the interaction between tearing modes at different radii, and the reconnection process evolves according to the safety factor profile.