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With first-principles kinetic simulations, we show that a large-scale Alfv\'en wave (AW) propagating in an inhomogeneous background decays into kinetic Alfv\'en waves (KAWs), triggering ion and electron energization. We demonstrate that the two species can access unequal amounts of the initial AW energy, experiencing differential heating. During the decay process, the electric field carried by KAWs produces non-Maxwellian features in the particle VDFs, in accordance with space observations. The process we present solely requires the interaction of a large-scale AW with a magnetic shear and may be relevant for several astrophysical and laboratory plasmas., Comment: Accepted in PRL

While recent wireless micromachines have shown increasing potential for medical use, their potential safety risks concerning biocompatibility need to be mitigated. They are typically constructed from materials that are not intrinsically compatible with physiological environments. Here, we propose a personalized approach by using patient blood-derivable biomaterials as the main construction fabric of wireless medical micromachines to alleviate safety risks from biocompatibility. We demonstrate 3D printed multiresponsive microswimmers and microrollers made from magnetic nanocomposites of blood plasma, serum albumin protein, and platelet lysate. These micromachines respond to time-variant magnetic fields for torque-driven steerable motion and exhibit multiple cycles of pH-responsive two-way shape memory behavior for controlled cargo delivery and release applications. Their proteinaceous fabrics enable enzymatic degradability with proteinases, thereby lowering risks of long-term toxicity. The personalized micromachine fabrication strategy we conceptualize here can affect various future medical robots and devices made of autologous biomaterials to improve biocompatibility and smart functionality., Science Advances, 7 (36), ISSN:2375-2548

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.

Neutron emission spectroscopy and neutron yield measurements are important for high neutral beam injection (NBI) power heating at the Experimental Advanced Superconducting Tokamak (EAST). The neutron yields mainly depend on the deposition from NBI to the deuterium plasmas in the EAST. We have recently used TRANSP with time dependent diagnostic results to simulate the transport process of 30s long pulse deuterium plasma discharges in the EAST, obtaining the time dependent fast ion distribution, neutron emission spectrum, and total neutron emission rate. Combined with the time trace of the result measured by a standard 235U fission chamber, the effects of different configurations of NBI heating in EAST fusion plasmas have been evaluated.

This paper summarizes the physical principles behind the novel three-ion scenarios using radio frequency waves in the ion cyclotron range of frequencies (ICRF). We discuss how to transform mode conversion electron heating into a new flexible ICRF technique for ion cyclotron heating and fast-ion generation in multi-ion species plasmas. The theoretical section provides practical recipes for selecting the plasma composition to realize three-ion ICRF scenarios, including two equivalent possibilities for the choice of resonant absorbers that have been identified. The theoretical findings have been convincingly confirmed by the proof-of-principle experiments in mixed H?D plasmas on the Alcator C-Mod and JET tokamaks, using thermal 3He and fast D ions from neutral beam injection as resonant absorbers. Since 2018, significant progress has been made on the ASDEX Upgrade and JET tokamaks in H?4He and H?D plasmas, guided by the ITER needs. Furthermore, the scenario was also successfully applied in JET D?3He plasmas as a technique to generate fusion-born alpha particles and study effects of fast ions on plasma confinement under ITER-relevant plasma heating conditions. Tuned for the central deposition of ICRF power in a small region in the plasma core of large devices such as JET, three-ion ICRF scenarios are efficient in generating large populations of passing fast ions and modifying the q-profile. Recent experimental and modeling developments have expanded the use of three-ion scenarios from dedicated ICRF studies to a flexible tool with a broad range of different applications in fusion research. AB - This paper summarizes the physical principles behind the novel three-ion scenarios using radio frequency waves in the ion cyclotron range of frequencies (ICRF). We discuss how to transform mode conversion electron heating into a new flexible ICRF technique for ion cyclotron heating and fast-ion generation in multi-ion species plasmas. The theoretical section provides practical recipes for selecting the plasma composition to realize three-ion ICRF scenarios, including two equivalent possibilities for the choice of resonant absorbers that have been identified. The theoretical findings have been convincingly confirmed by the proof-of-principle experiments in mixed H?D plasmas on the Alcator C-Mod and JET tokamaks, using thermal 3He and fast D ions from neutral beam injection as resonant absorbers. Since 2018, significant progress has been made on the ASDEX Upgrade and JET tokamaks in H?4He and H?D plasmas, guided by the ITER needs. Furthermore, the scenario was also successfully applied in JET D?3He plasmas as a technique to generate fusion-born alpha particles and study effects of fast ions on plasma confinement under ITER-relevant plasma heating conditions. Tuned for the central deposition of ICRF power in a small region in the plasma core of large devices such as JET, three-ion ICRF scenarios are efficient in generating large populations of passing fast ions and modifying the q-profile. Recent experimental and modeling developments have expanded the use of three-ion scenarios from dedicated ICRF studies to a flexible tool with a broad range of different applications in fusion research. The authors are grateful to the anonymous reviewers for their constructive comments that allowed us to improve the paper, I. Voitsekhovitch for her valuable comments and suggestions during the paper preparation, and to H. Meyer and J. Faustin for fruitful discussions. We thank the ITPA Energetic Particle Physics Topical Group for its support. This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014–2018 and 2019–2020 under Grant Agreement No. 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission. Part of this work was also carried out in the framework of projects done for the ITER Scientist Fellow Network (ISFN). ITER is the Nuclear Facility INB No. 174. The views and opinions expressed herein do not necessarily reflect those of the ITER Organization. This publication is provided for scientific purposes only. Its contents should not be considered as commitments from the ITER Organization as a nuclear operator in the frame of the licensing process. Peer Reviewed "Article signat per 67 autors/es: Ye. O. Kazakov, J. Ongena, J. C. Wright, S. J. Wukitch, V. Bobkov, J. Garcia, V. G. Kiptily, M. J. Mantsinen, M. Nocente, M. Schneider, H. Weisen, Y. Baranov, M. Baruzzo, R. Bilato, A. Chomiczewska, R. Coelho, T. Craciunescu K. Crombé, M. Dreval, R. Dumont, P. Dumortier, F. Durodié, J. Eriksson, M. Fitzgerald, J. Galdon-Quiroga, D. Gallart, M. Garcia-Muñoz, L. Giacomelli, C. Giroud, J. Gonzalez-Martin, A. Hakola, P. Jacquet, T. Johnson, A. Kappatou, D. Keeling, D. King, K. K. Kirov, P. Lamalle, M. Lennholm, E. Lerche, M. Maslov, S. Mazzi, S. Menmuir, I. Monakhov5, F. Nabais, M. F. F. Nave, R. Ochoukov, A. R. Polevoi, S. D. Pinches, U. Plank, D. Rigamonti, M. Salewski, P. A. Schneider, S. E. Sharapov, Ž. Štancar, A. Thorman, D. Valcarcel, D. Van Eester, M. Van Schoor, J. Varje, M. Weiland, N. Wendler, JET Contributors, ASDEX Upgrade Team, EUROfusion MST1 Team, and Alcator C-Mod Team"

Experiments at ASDEX Upgrade have been conducted to study the impact of the alignment of external magnetic perturbations (MPs) with n = 2 toroidal mode symmetry on the power threshold of the transition from L- to H-mode ([Formula: see text]). This is interesting for ITER, since its operation will rely on an MP field configuration that securely suppresses edge localized modes (ELMs) while avoiding an increase in the L–H power threshold. [Formula: see text] can increase up to [Formula: see text] when the external MP field is aligned to maximize the associated plasma response at the edge. This alignment deviates from the equilibrium field pitch and is similar to the one that is required to suppress ELMs by MPs in H-mode. The relative radial magnetic field perturbation [Formula: see text] has to exceed a critical value, roughly [Formula: see text] (at the q = 5 surface), to cause an increase in [Formula: see text]. This value is above the one that is required to sustain ELM suppression at ASDEX Upgrade. Combining plasma response calculations with a critical value of [Formula: see text] captures well the measured dependencies of [Formula: see text] on the alignment of the applied MP field. The increase in [Formula: see text] is accompanied by a change in the [Formula: see text] flow velocity ([Formula: see text]) profile. The application of MPs in L-mode flattens and reverses the edge [Formula: see text] profile when at least [Formula: see text] is applied. This is the same value above which [Formula: see text] increases. More heating power is needed to roughly achieve the same shear in the [Formula: see text] profile at the transition to H-mode as without MPs but with an [Formula: see text] profile that is shifted into the ion diamagnetic direction.

The next fusion project DEMO, which will be the evolution of the experimental fusion reactor [International Tokamak Experimental Reactor (ITER)], would require a high efficient energy production. As in ITER, DEMO will use fast Neutral Beam (NB) injectors to increase the plasma temperature needed for the fusion reaction. A way to recover the electric energy production efficiency in DEMO could be the beam energy recovery in the NB production, which is produced by a D− beam, neutralized by a gas cell with 60% efficiency. A compact energy recovery device with an axisymmetric cylindrical ion collector that uses only decelerating electric fields combined with the beam space charge effect has been recently proposed. It can be used for a test on the beam of the NIO1 (Negative Ion Optimization 1) source, a compact ion source (scaled down from ITER size sources) that has been developed at INFN-LNL and Consorzio RFX (Padua). The detailed collector design to be used on one of the beamlets of the NIO1 source within typical space limitation is presented and discussed here. Furthermore, a preliminary trajectory simulation for a beam with a rectangular geometry similar to the beam used in ITER to verify the beam recovery for a nonaxial symmetric geometry is also shown.The next fusion project DEMO, which will be the evolution of the experimental fusion reactor [International Tokamak Experimental Reactor (ITER)], would require a high efficient energy production. As in ITER, DEMO will use fast Neutral Beam (NB) injectors to increase the plasma temperature needed for the fusion reaction. A way to recover the electric energy production efficiency in DEMO could be the beam energy recovery in the NB production, which is produced by a D− beam, neutralized by a gas cell with 60% efficiency. A compact energy recovery device with an axisymmetric cylindrical ion collector that uses only decelerating electric fields combined with the beam space charge effect has been recently proposed. It can be used for a test on the beam of the NIO1 (Negative Ion Optimization 1) source, a compact ion source (scaled down from ITER size sources) that has been developed at INFN-LNL and Consorzio RFX (Padua). The detailed collector design to be used on one of the beamlets of the NIO1 source within ty...

The exhaust gas in the future Tokamak demonstration reactor (DEMO) consists of more than 80 % of unburned fuel gas (D and T) with the balance being He, plasma enhancement gases (PEGs), and impurities. In DEMO we plan to recover the fuel gas (D and T) and PEGs. The paper focuses on the removal of He from the fuel gas and PEGs from the fusion reactors. Hence, the aim of the research is to test potential technologies to achieve this goal. In this paper, a carbon molecular sieve (CMS) membrane supplied by Tecnalia (Spain) is tested to remove the He along with D and T whilst retaining the PEGs. Agreement letter of collaboration Tecnalia-ENEA

An incoherent Thomson scattering diagnostic will be installed in the JT-60SA tokamak to measure electron temperature and electron density profiles. The target radial spatial resolution is 25 mm with 46 spatial channels. The accuracy in electron temperature and density is a few percent at ne = 7.5 × 1019 m-3, which is the expected value in the plasma core. This paper presents the designs of collection optics, fibers with their alignment system, and polychromators. The collection optics overcomes unique issues for superconducting fusion devices, i.e., limited design space, high-temperature measurements, and harsh radiation condition. When in several years the more performing plasma will generate intense nuclear radiation, the lens materials of the optics can be replaced by radiation resistant glasses without major changes in the lens holder. It will prevent transmission degradation and keep stable measurement accuracy.

We model meteoroids entering the Earth[U+05F3]s atmosphere as objects surrounded by non-magnetized plasma, and consider efficient numerical simulation of radar reflections from meteors in the time domain. Instead of the widely used finite difference time domain method (FDTD), we use more generalized finite differences by applying the discrete exterior calculus (DEC) and non-uniform leapfrog-style time discretization. The computational domain is presented by convex polyhedral elements. The convergence of the time integration is accelerated by the exact controllability method. The numerical experiments show that our code is efficiently parallelized. The DEC approach is compared to the volume integral equation (VIE) method by numerical experiments. The result is that both methods are competitive in modelling non-magnetized plasma scattering. For demonstrating the simulation capabilities of the DEC approach, we present numerical experiments of radar reflections and vary parameters in a wide range. © 2016 Elsevier Ltd.

This work presents the results of experimental and theoretical research of impact of tailored density profile and application of external axial magnetic field on initial spatial distribution of the plasma density in the plasma metal puff Z-pinch and on its implosion dynamics. It has been discovered that upon implosion of the plasma metal puff Z-pinch some stripes interpreted as the system of two coaxial shells appear on the optical images. With the help of numerical simulation, the formation of the plasma liner consisting of a mixture of carbon and bismuth ions and formed by the expansion of the plasma jet of the arc burning on the bismuth electrode has been considered in this work. It has been shown that the lightweight carbon ions facilitate formation of the density distribution smoothly decreasing with the increase in radius, that, in turn, leads to suppression of the Rayleigh-Taylor instability in the current sheath upon further implosion. It has also been demonstrated that availability of the two types of ions in plasma considerably different in mass leads to formation (in the compression phase) of a double shell with externally located heavy ions. It has also been revealed that the application of the external axial magnetic field leads to reduction in the plasma metal puff Z-pinch initial diameter. © 2020 Author(s). This work was funded by Russian Science Foundation. Project No. 19–19-00127.

12th International Congress on Artificial Materials for Novel Wave Phenomena, MetaMaterials 2018 -- 27 August 2018 through 1 September 2018 -- -- 142487, The possibility of negative refractive index in plasma and magnetic materials was considered with a new approach. It is shown that simplification of fundamental equations by neglecting losses leads to losing important features about refractive index and therefore, it is expedient to take into consideration imaginary parts of dielectric permittivity and magnetic permeability for an accurate statement of the problem. Also active materials were briefly discussed via new approach. © 2018 IEEE., FDK-2017-8948, The authors would like to acknowledge the Scientific and Technological Research Council of Turkey (TÜBİTAK) and the Scientific Project Unit of Cukurova University (FDK-2017-8948) for full financial support.

The edge of the RFX-mod ( R = 2 m, a = 0.46 m) Reversed Field Pinch is characterized by weak magnetic chaos affecting ion and electron diffusion. Edge particle transport is strongly influenced by a toroidal asymmetry caused by magnetic islands. An ambipolar radial electric field ensures local neutrality and possesses the same symmetry as the parent magnetic ripple: the result is the modulation of the perpendicular flow, with a slowing-down at the island X-point. In this paper we present a complete statistical analysis, over a large database of RFX-mod discharges, of the edge properties as they are modified by the magnetic topology: the plasma wall footprint follows the helical shape of the dominant central mode ( m / n = 1/7), with an increase of H α emission and electron density corresponding to the O-point of the inner magnetic island. Edge turbulence is modified by the magnetic topology, being generated in the O-point region and damped near the X-point.

The Upper port Wide Angle Viewing System (UWAVS) is diagnostic system that provides visible light and InfraRed images of the divertor area in the ITER tokamak. The main functions of the UWAVS are to provide divertor wall and plasma luminance and temperature distributions at high spatial resolution (3–5 mm) and frame rates (500–1000 Hz) for machine protection as well as to enable species influx monitoring (Be, C, O, Ne, Kr, Ar) to the divertor. In addition the UWAVS supports visual inspection as well as various other measurements. For this purpose an optical system is designed to image the divertor area onto four different cameras in the port cell area. Altogether five of these systems are foreseen to provide sufficient coverage of the divertor area (150m2). Each of these systems is equipped with active alignment to compensate for vessel motion, internal transmission monitoring sources to assess mirror reflectivity degradation, as well as calibration sources in the port cell to monitor the camera pixel response, all to guarantee the overall systems performance during operation and over lifetime. All these aspects are combined into an optical design targeting diffraction limited imaging, taking in account the complexities of the harsh ITER environment. This design and its performance are supported by manufacturability analysis, tolerancing analysis and alignment strategy and Structural Thermal Optical Performance (STOP) analysis. The latter is an TNO in-house developed method that provides a direct and reproducible link between the structural thermal analysis for a combined load case comprising nuclear heating, cooling water pressure and temperature, gravity and mounting stresses (in Ansys) and the associated changes of the optical performance, including imaging quality and alignment (in Zemax OpticStudio).

The Upper port Wide Angle Viewing System (UWAVS) is diagnostic system that provides visible light and InfraRed images of the divertor area in the ITER tokamak. The main functions of the UWAVS are to provide divertor wall and plasma luminance and temperature distributions at high spatial resolution (3–5 mm) and frame rates (500–1000 Hz) for machine protection as well as to enable species influx monitoring (Be, C, O, Ne, Kr, Ar) to the divertor. In addition the UWAVS supports visual inspection as well as various other measurements. For this purpose an optical system is designed to image the divertor area onto four different cameras in the port cell area. Altogether five of these systems are foreseen to provide sufficient coverage of the divertor area (150m2). Each of these systems is equipped with active alignment to compensate for vessel motion, internal transmission monitoring sources to assess mirror reflectivity degradation, as well as calibration sources in the port cell to monitor the camera pixel response, all to guarantee the overall systems performance during operation and over lifetime. All these aspects are combined into an optical design targeting diffraction limited imaging, taking in account the complexities of the harsh ITER environment. This design and its performance are supported by manufacturability analysis, tolerancing analysis and alignment strategy and Structural Thermal Optical Performance (STOP) analysis. The latter is an TNO in-house developed method that provides a direct and reproducible link between the structural thermal analysis for a combined load case comprising nuclear heating, cooling water pressure and temperature, gravity and mounting stresses (in Ansys) and the associated changes of the optical performance, including imaging quality and alignment (in Zemax OpticStudio).

We report the first experimental evidence of a controlled transition from the generation of periodic bursts of electromagnetic radiation into the continuous-wave regime of a cyclotron maser formed in magnetically confined nonequilibrium plasma. The kinetic cyclotron instability of the extraordinary wave of weakly inhomogeneous magnetized plasma is driven by the anisotropic electron population resulting from electron cyclotron plasma heating in a MHD-stable minimum-B open magnetic trap. peerReviewed

A large database of reciprocating probe data from the edge plasma of TCV (Tokamak a Configuration Variable) is used to test the radial velocity scalings of filaments from analytical theory [Myra et al., Phys. Plasmas 13, 112502 (2006)]. The measured velocities are mainly scattered between zero and a maximum velocity which varies as a function of size and collisionality in agreement with the analytical scalings. The scatter is consistent with mechanisms that tend to slow the velocity of individual filaments. While the radial velocities were mainly clustered between 0.5 and 2 km/s, a minority reached outward velocities as high as 5 km/s or inward velocities as high as −4 km/s. Inward moving filaments are only observed in regions of high poloidal velocity shear in discharges with B × ∇B away from the X-point, a new finding. The filaments have diameters clustered between 3 and 11 mm, and normalized sizes a clustered between 0.3 and 1.1, such that most filaments populate the resistive-ballooning regime; therefore, most of the filaments in TCV have radial velocities with little or no dependence on collisionality. Improvements in cross-correlation techniques and conditional averaging techniques are discussed which reduce the sizes determined for the largest filaments, including those larger than the scrape-off layer.A large database of reciprocating probe data from the edge plasma of TCV (Tokamak a Configuration Variable) is used to test the radial velocity scalings of filaments from analytical theory [Myra et al., Phys. Plasmas 13, 112502 (2006)]. The measured velocities are mainly scattered between zero and a maximum velocity which varies as a function of size and collisionality in agreement with the analytical scalings. The scatter is consistent with mechanisms that tend to slow the velocity of individual filaments. While the radial velocities were mainly clustered between 0.5 and 2 km/s, a minority reached outward velocities as high as 5 km/s or inward velocities as high as −4 km/s. Inward moving filaments are only observed in regions of high poloidal velocity shear in discharges with B × ∇B away from the X-point, a new finding. The filaments have diameters clustered between 3 and 11 mm, and normalized ...