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The gyrotron has already demonstrated the capability of generating high-power coherent electromagnetic (EM) radiation at high frequencies and finds application in fusion plasma heating and magnetic resonance spectroscopy. In this article, we propose a ${W}$ -band gyrotron, which uses a so-called multimode switching scheme to realize ultrabroadband frequency tuning capability, nearly covering about 70% of the ${W}$ -band (75–110 GHz) range. The tuning strategy used to suppress mode competition and stabilize gun parameters in such an open-cavity multimode switching gyrotron is presented. Theoretical study shows that the multimode switching gyrotron can generate a frequency tuning range much wider than a conventional step-tuning gyrotron or a single-mode tuning gyrotron. In addition, another technology that uses a slot-assisted circuit to select only axis-symmetrical TE operating modes is presented. Using such a circuit, a tuning range of about 25 GHz in the ${W}$ -band is obtainable. The proposed multimode switching gyrotron is a promising ultrabroadband source for millimeter-wave and terahertz-wave applications.

TJ-II stellarator results on modelling and validation of plasma flow asymmetries due to on-surface potential variations, plasma fuelling physics, Alfvén eigenmodes (AEs) control and stability, the interplay between turbulence and neoclassical (NC) mechanisms and liquid metals are reported. Regarding the validation of the neoclassically predicted potential asymmetries, its impact on the radial electric field along the flux surface has been successfully validated against Doppler reflectometry measurements. Research on the physics and modelling of plasma core fuelling with pellets and tracer encapsulated solid pellet injection has shown that, although post-injection particle radial redistributions can be understood qualitatively from NC mechanisms, turbulence and fluctuations are strongly affected during the ablation process. Advanced analysis tools based on transfer entropy have shown that radial electric fields do not only affect the radial turbulence correlation length but are also capable of reducing the propagation of turbulence from the edge into the scrape-off layer. Direct experimental observation of long range correlated structures show that zonal flow structures are ubiquitous in the whole plasma cross-section in the TJ-II stellarator. Alfvénic activity control strategies using ECRH and ECCD as well as the relation between zonal structures and AEs are reported. Finally, the behaviour of liquid metals exposed to hot and cold plasmas in a capillary porous system container was investigated.

Drift wave instabilities (DWI) associated with the two-fluid dynamics seems to be responsible for anomalous transport in modern day tokamaks. Ballooning instabilities tend to exchange flux tubes of different pressure, resulting in convective transport. The micro-level turbulence (drift wave) is coupled with the macro-level (ballooning mode) dynamics in fusion experiments. The co-existence of DWI and drift ballooning instabilities (DBI) is discussed in this work using a four-field plasma model. The formulation preserves both the microscopic and macroscopic dynamics of plasma. To demonstrate the coupling, a new dispersion relation is derived to analyze stability of the coupled modes in a non-uniform magnetized plasma. Linear stability of coupled drift-ballooning and drift-acoustic modes have been explored. The two-fluid effect (micro-level influence) through diamagnetic drift frequency for electrons and curvature drift frequency on unstable modes are demonstrated., Drift wave; Fusion plasma; Plasma instability; Coupled modes.

High-Z impurities released from plasma-material interactions have been shown to limit the performance of fusion plasmas, and understanding these impurity transport mechanisms throughout the...

Most if not all tokamak heating scenarios involve multiple ion populations being heated simultaneously. To allow the simulation of various aspects of physics dynamics determining the characteristics of operational scenarios in a flexible way, speedy yet sufficiently accurate models are needed, and they should be connected to each other via a ‘backbone’. Under the umbrella of EUROfusion's Integrated Modelling efforts, such a structure is provided. The present paper focuses on one physics aspect: auxiliary heating. After solving the wave equation or beam source equation, this requires solving a set of coupled Fokker–Planck equations for the various populations involved. The adopted modules – enabling accounting for the Coulomb collisional interaction of several non-Maxwellian (minority, majority and beam) populations – are discussed and a practical example of their use is provided: the JET ‘baseline’ scenario heating a minority of${}^3\textrm {He}$ions in a balanced D$+$T mix heated by D and T neutral beams.

High speed visible imaging allows to visualize the 2-dimensional dynamics of fast phenomena in the boundary layer of fusion plasmas. Here we describe the two high speed visible cameras currently operating in the JET tokamak, enabling the simultaneous observation of a large fraction of the JET tokamak vacuum vessel with the appropriate configuration for many different plasma phenomena. As an example we discuss recent observations on the spatial and temporal dynamics of visible emission at the outer wall during Edge Localized Modes (ELMs) and the penetration of Shattered Injected Pellet (SPI) into the plasma.

Potential loss of energetic ions including alphas and radio-frequency tail ions due to classical orbit effects and magnetohydrodynamic instabilities (MHD) are central physics issues in the design and experimental physics programme of the SPARC tokamak. The expected loss of fusion alpha power due to ripple-induced transport is computed for the SPARC tokamak design by the ASCOT and SPIRAL orbit-simulation codes, to assess the expected surface heating of plasma-facing components. We find good agreement between the ASCOT and SPIRAL simulation results not only in integrated quantities (fraction of alpha power loss) but also in the spatial, temporal and pitch-angle dependence of the losses. If the toroidal field (TF) coils are well-aligned, the SPARC edge ripple is small (0.15–0.30 %), the computed ripple-induced alpha power loss is small ( ${\sim } 0.25\,\%$ ) and the corresponding peak surface power density is acceptable ( $244\ \textrm{kW}\ \textrm {m}^{-2}$ ). However, the ripple and ripple-induced losses increase strongly if the TF coils are assumed to suffer increasing magnitudes of misalignment. Surface heat loads may become problematic if the TF coil misalignment approaches the centimetre level. Ripple-induced losses of the energetic ion tail driven by ion cyclotron range of frequency (ICRF) heating are not expected to generate significant wall or limiter heating in the nominal SPARC plasma scenario. Because the expected classical fast-ion losses are small, SPARC will be able to observe and study fast-ion redistribution due to MHD including sawteeth and Alfvén eigenmodes (AEs). SPARC's parameter space for AE physics even at moderate $Q$ is shown to reasonably overlap that of the demonstration power plant ARC (Sorbom et al., Fusion Engng Des., vol. 100, 2015, p. 378), and thus measurements of AE mode amplitude, spectrum and associated fast-ion transport in SPARC would provide relevant guidance about AE behaviour expected in ARC.

The isotope dependence of plasma transport will have a significant impact on the performance of future D-T experiments in JET and ITER and eventually on the fusion gain and economics of future reactors. In preparation for future D-T operation on JET, dedicated experiments and comprehensive transport analyses were performed in H, D and H-D mixed plasmas. The analysis of the data has demonstrated an unexpectedly strong and favourable dependence of the global confinement of energy, momentum and particles in ELMy H-mode plasmas on the atomic mass of the main ion species, the energy confinement time scaling as ${\tau _E}\sim {A^{0.5}}$ (Maggi et al., Plasma Phys. Control. Fusion, vol. 60, 2018, 014045; JET Team, Nucl. Fusion, vol. 39, 1999, pp. 1227–1244), i.e. opposite to the expectations based only on local gyro-Bohm (GB) scaling, ${\tau _E}\sim {A^{ - 0.5}}$ , and stronger than in the commonly used H-mode scaling for the energy confinement (Saibene et al., Nucl. Fusion, vol. 39, 1999, 1133; ITER Physics Basis, Nucl. Fusion, vol. 39, 1999, 2175). The scaling of momentum transport and particle confinement with isotope mass is very similar to that of energy transport. Nonlinear local GENE gyrokinetic analysis shows that the observed anti-GB heat flux is accounted for if collisions, E × B shear and plasma dilution with low-Z impurities (9Be) are included in the analysis (E and B are, respectively the electric and magnetic fields). For L-mode plasmas a weaker positive isotope scaling ${\tau _E}\sim {A^{0.14}}$ has been found in JET (Maggi et al., Plasma Phys. Control. Fusion, vol. 60, 2018, 014045), similar to ITER97-L scaling (Kaye et al., Nucl. Fusion, vol. 37, 1997, 1303). Flux-driven quasi-linear gyrofluid calculations using JETTO-TGLF in L-mode show that local GB scaling is not followed when stiff transport (as is generally the case for ion temperature gradient modes) is combined with an imposed boundary condition taken from the experiment, in this case predicting no isotope dependence. A dimensionless identity plasma pair in hydrogen and deuterium L-mode plasmas has demonstrated scale invariance, confirming that core transport physics is governed, as expected, by the 4 dimensionless parameters ρ*, ν*, β, q (normalised ion Larmor radius, collisionality, plasma pressure and safety factor) consistently with global quasi-linear gyrokinetic TGLF calculations (Maggi et al., Nucl. Fusion, vol. 59, 2019, 076028). We compare findings in JET with those in different devices and discuss the possible reasons for the different isotope scalings reported from different devices. The diversity of observations suggests that the differences may result not only from differences affecting the core, e.g. heating schemes, but are to a large part due to differences in device-specific edge and wall conditions, pointing to the importance of better understanding and controlling pedestal and edge processes.

The SPARC tokamak is a critical next step towards commercial fusion energy. SPARC is designed as a high-field ($B_0 = 12.2$T), compact ($R_0 = 1.85$m,$a = 0.57$m), superconducting, D-T tokamak with the goal of producing fusion gain$Q>2$from a magnetically confined fusion plasma for the first time. Currently under design, SPARC will continue the high-field path of the Alcator series of tokamaks, utilizing new magnets based on rare earth barium copper oxide high-temperature superconductors to achieve high performance in a compact device. The goal of$Q>2$is achievable with conservative physics assumptions ($H_{98,y2} = 0.7$) and, with the nominal assumption of$H_{98,y2} = 1$, SPARC is projected to attain$Q \approx 11$and$P_{\textrm {fusion}} \approx 140$MW. SPARC will therefore constitute a unique platform for burning plasma physics research with high density ($\langle n_{e} \rangle \approx 3 \times 10^{20}\ \textrm {m}^{-3}$), high temperature ($\langle T_e \rangle \approx 7$keV) and high power density ($P_{\textrm {fusion}}/V_{\textrm {plasma}} \approx 7\ \textrm {MW}\,\textrm {m}^{-3}$) relevant to fusion power plants. SPARC's place in the path to commercial fusion energy, its parameters and the current status of SPARC design work are presented. This work also describes the basis for global performance projections and summarizes some of the physics analysis that is presented in greater detail in the companion articles of this collection.

Ion cyclotron range of frequencies (ICRF) heating will be the sole auxiliary heating method on SPARC for both full-field (Bt0 ~ 12 T) D–T operation and reduced field (Bt0 ~ 8 T) D–D operation. Using the fast wave at ~120 MHz, good wave penetration and strong single-pass absorption is expected for D–T(3He), D(3He), D(H) and4He(H) heating scenarios. The dependences of wave absorption on${k_\parallel }$,3He concentration, resonance location, antenna poloidal location and losses on alpha particles and ash have been studied. The antenna loading has been assessed by comparison with the Alcator C-Mod antennae. An antenna spectrum of${k_\parallel }\sim 15\text{--}18\,{\textrm{m}^{ - 1}}$is shown to be good for both core absorption and edge coupling. For the control of impurity sources, the antenna straps are rotated ~10° to be perpendicular to the B field and the straps can run with different power levels in order to optimize the antenna spectrum and to minimize the image current on the antenna frame. Combining the physics constraints with the SPARC port design, maintenance requirement and contingency against antenna failure during D–T operation, we plan to mount on the inner wall of the vacuum vessel a total of 12 4-strap antennae in 6 ports while keeping 3-strap antennae that are insertable and removable on port plugs as the backup option.

SPARC is being designed to operate with a normalized beta of $\beta _N=1.0$ , a normalized density of $n_G=0.37$ and a safety factor of $q_{95}\approx 3.4$ , providing a comfortable margin to their respective disruption limits. Further, a low beta poloidal $\beta _p=0.19$ at the safety factor $q=2$ surface reduces the drive for neoclassical tearing modes, which together with a frozen-in classically stable current profile might allow access to a robustly tearing-free operating space. Although the inherent stability is expected to reduce the frequency of disruptions, the disruption loading is comparable to and in some cases higher than that of ITER. The machine is being designed to withstand the predicted unmitigated axisymmetric halo current forces up to 50 MN and similarly large loads from eddy currents forced to flow poloidally in the vacuum vessel. Runaway electron (RE) simulations using GO+CODE show high flattop-to-RE current conversions in the absence of seed losses, although NIMROD modelling predicts losses of ${\sim }80$ %; self-consistent modelling is ongoing. A passive RE mitigation coil designed to drive stochastic RE losses is being considered and COMSOL modelling predicts peak normalized fields at the plasma of order $10^{-2}$ that rises linearly with a change in the plasma current. Massive material injection is planned to reduce the disruption loading. A data-driven approach to predict an oncoming disruption and trigger mitigation is discussed.

The radial correlation length Lr of electron density fluctuations has been measured using a two-channel Omode reflectometer system in the helical-axis fusion plasma experimental device Heliotron J. The experimental results show that Lr is around 1.4 n 0.7mm in three different magnetic field bumpiness configurations (i.e. low, medium and high) for the low-density electron cyclotron heating (ECH) discharges. In high-density neutral beam injection (NBI) discharges with HIGP and pellet injection, Lr is found to be around 1 n 0.2 mm.

Finding coil sets with desirable physics and engineering properties is a crucial step in the design of modern stellarator devices. Existing stellarator coil optimization codes ultimately produce zero-thickness filament coils. However, stellarator coils have finite depth and thickness, which can make the single-filament model a poor approximation, particularly when coil build dimensions are relatively large compared to the coil–plasma distance. In this paper, we present a new method for designing coils with finite builds and present a mechanism to optimize the orientation of the winding pack. We approximate finite-build coils with a multi-filament model. A numerical implementation has been developed, and applications to the Helically Symmetric eXperiment stellarator and a new UW-Madison quasihelically symmetric configuration are shown.

A simplified analytical form of the on-axis magnetic well and Mercier's criterion for interchange instabilities for arbitrary three-dimensional magnetic field geometries is derived. For this purpose, a near-axis expansion based on a direct coordinate approach is used by expressing the toroidal magnetic flux in terms of powers of the radial distance to the magnetic axis. For the first time, the magnetic well and Mercier's criterion are then written as a one-dimensional integral with respect to the axis arclength. When compared with the original work of Mercier, the derivation here is presented using modern notation and in a more streamlined manner that highlights essential steps. Finally, these expressions are verified numerically using several quasisymmetric and non-quasisymmetric stellarator configurations including Wendelstein 7-X.

In this work, the FALCON code is adopted for illustrating the features of shear Alfvén and sound continuous spectra in toroidal fusion plasmas. The FALCON codes employ the local Floquet analysis discussed in (Phys. Plasmas, vol. 26, issue 8, 2019, 082502) for computing global structures of continuous spectra in general toroidal geometry. As particular applications, reference equilibria for the divertor tokamak test and ASDEX Upgrade plasmas are considered. In particular, we illustrate the importance of mode polarization for recognizing the physical relevance of the various branches of the continuous spectra in the ideal magnetohydrodynamics limit. We also analyse the effect of plasma compression and the validity of the slow sound approximation.

The paper presents a brief overview of energetic particle diagnostics in fusion plasmas. Differences between techniques used in nuclear and non-nuclear machines are in particular underlined and the challenges faced by fast ion diagnostics in a burning plasma are discussed.

The scintillator-based fast-ion loss detector (FILD) project in Japan, Korea, and China has been accelerated in an international collaboration framework to enhance comprehensive understanding of fast-ion behaviors in toroidal fusion plasmas. The FILDs in LHD heliotron, KSTAR, HL-2A, and EAST tokamaks are successfully working as a result of joint work. Physics experiments on fast ions, such as effects of Alfvenic mode, tearing mode, resonant magnetic perturbation field, and disruption on fast-ion behaviors are ongoing. This paper describes the FILD developed for each device and those effects on fast ions in LHD, KSTAR, HL-2A, and EAST.

The condition of omnigenity is investigated, and applied to the near-axis expansion of Garren & Boozer (Phys. Fluids B, vol. 3 (10), 1991a, pp. 2805–2821). Due in part to the particular analyticity requirements of the near-axis expansion, we find that, excluding quasi-symmetric solutions, only one type of omnigenity, namely quasi-isodynamicity, can be satisfied at first order in the distance from the magnetic axis. Our construction provides a parameterization of the space of such solutions, and the cylindrical reformulation and numerical method of Landreman & Sengupta (J. Plasma Phys., vol. 84 (6), 2018, 905840616); Landreman et al. (J. Plasma Phys., vol. 85 (1), 2019, 905850103), enables their efficient numerical construction.

Stellarator configurations with reactor relevant energetic particle losses are constructed by simultaneously optimizing for quasisymmetry and an analytically derived metric ( $\unicode[STIX]{x1D6E4}_{c}$ ), which attempts to align contours of the second adiabatic invariant, $J_{\Vert }$ with magnetic surfaces. Results show that with this optimization scheme it is possible to generate quasihelically symmetric equilibria on the scale of ARIES-CS which completely eliminate all collisionless alpha particle losses within normalized radius $r/a=0.3$ . We show that the best performance is obtained by reducing losses at the trapped–passing boundary. Energetic particle transport can be improved even when neoclassical transport, as calculated using the metric $\unicode[STIX]{x1D716}_{\text{eff}}$ , is degraded. Several quasihelically symmetric equilibria with different aspect ratios are presented, all with excellent energetic particle confinement.

We present initial results in the development of a gyrokinetic particle-in-cell code for the whole-volume modeling of stellarators. This is achieved through two modifications to the X-point Gyrokinetic Code (XGC), originally developed for tokamaks. One is an extension to three-dimensional geometries with an interface to Variational Moments Equilibrium Code (VMEC) data. The other is a connection between core and edge regions that have quite different field-line structures. The VMEC equilibrium is smoothly extended to the edge region by using a virtual casing method. Non-axisymmetric triangular meshes in which triangle nodes follow magnetic field lines in the toroidal direction are generated for field calculation using a finite-element method in the entire region of the extended VMEC equilibrium. These schemes are validated by basic benchmark tests relevant to each part of the calculation cycle, that is, particle push, particle-mesh interpolation, and field solver in a magnetic field equilibrium of Large Helical Device including the edge region. The developed code also demonstrates collisionless damping of geodesic acoustic modes and steady states with residual zonal flow in the core region.

The development of 140GHz and 170GHz MW gyrotron research in BVERI are given in this paper. There are three versions of the 140GHz MW gyrotrons which have been designed and fabricated in BVERI. All of the 140GHz gyrotron have been tested and the experimental results have been enhanced to output power of 500KW with Gaussian beam. Using the techniques achieved in the study on the 140GHz gyrotorns, the 170GHz gyrotron has been designed with the sing-depressed collector and quasi-optical mode converter.

To properly treat the collisional transport of alpha particles due to a weakly rippled tokamak magnetic field the tangential magnetic drift due to its gradient (the $\unicode[STIX]{x1D735}B$ drift) and pitch angle scatter must be retained. Their combination gives rise to a narrow boundary layer in which collisions are able to match the finite trapped response to the ripple to the vanishing passing response of the alphas. Away from this boundary layer collisions are ineffective. There the $\unicode[STIX]{x1D735}B$ drift of the alphas balances the small radial drift of the trapped alphas caused by the ripple. A narrow collisional boundary layer is necessary since this balance does not allow the perturbed trapped alpha distribution function to vanish at the trapped–passing boundary. The solution of this boundary layer problem allows the alpha transport fluxes to be evaluated in a self-consistent manner to obtain meaningful constraints on the ripple allowable in a tokamak fusion reactor. A key result of the analysis is that collisional alpha losses are insensitive to the ripple near the equatorial plane on the outboard side where the ripple is high. As the high field side ripple is normally very small, collisional $\sqrt{\unicode[STIX]{x1D708}}$ ripple transport is unlikely to be a serious issue.

This tutorial covers validation of gyrokinetic turbulent-transport models via comparison of measured turbulence with realistic simulations of fusion plasmas. It presents a brief history of validation of gyrokinetic simulations, the principal challenges encountered, a limited survey of common turbulence diagnostics used on tokamaks and stellarators, an overview of the fundamentals of synthetic diagnostic models and a discussion of frontiers in turbulent-transport model validation.

Avoiding impurity accumulation is a requirement for steady-state stellarator operation. The accumulation of impurities can be heavily affected by variations in their density on the flux-surface. Using recently derived semi-analytic expressions for the transport of a collisional impurity species with high-$Z$ and flux-surface density-variation in the presence of a low-collisionality bulk ion species, we numerically optimize the impurity density-variation on the flux-surface to minimize the radial peaking factor of the impurities. These optimized density-variations can reduce the core impurity density by $0.75^Z$ (with $Z$ the impurity charge number) in the Large Helical Device case considered here, and by $0.89^Z$ in a Wendelstein 7-X standard configuration case. On the other hand, when the same procedure is used to find density-variations that maximize the peaking factor, it is notably increased compared to the case with no density-variation. This highlights the potential importance of measuring and controlling these variations in experiments., 19 figures, 17 pages. Accepted into Nuclear Fusion

Runaway electrons present a potential threat to the safe operation of future nuclear fusion large facilities based on the tokamak principle (e.g., ITER). The article presents an implementation of runaway electron current estimations at COMPASS tokamak. The method uses a theoretical method developed by Fujita et al., with the difference in using experimental measurements from EFIT and Thomson scattering. The procedure was explained on the COMPASS discharge number 7298, which has a significant runaway electron population. Here, it was found that at least 4 kA of the plasma current is driven by the runaway electrons. Next, the method aws used on the set of plasma discharges with the variable electron plasma density. The difference in the plasma current was explained by runaway electrons, and their current was estimated using the aforementioned method. The experimental results are compared with the theory and simulation. The comparison presented some disagreements, showing the possible direction for the code development. Additional application on runaway electron energy limit is also addressed.

International audience; The Wendelstein 7-X (W7-X) optimized stellarator fusion experiment,which went into operation in 2015, has been operating since 2017 with anun-cooled modular graphite divertor. This allowed first divertor physicsstudies to be performed at pulse energies up to 80 MJ, as opposed to 4MJ in the first operation phase, where five inboard limiters wereinstalled instead of a divertor. This, and a number of other upgrades tothe device capabilities, allowed extension into regimes of higher plasmadensity, heating power, and performance overall, e.g. setting a newstellarator world record triple product. The paper focuses on the firstphysics studies of how the island divertor works. The plasma heat loadsarrive to a very high degree on the divertor plates, with only minorheat loads seen on other components, in particular baffle structuresbuilt in to aid neutral compression. The strike line shapes andlocations change significantly from one magnetic configuration toanother, in very much the same way that codes had predicted they would.Strike-line widths are as large as 10 cm, and the wetted areas alsolarge, up to about 1.5 m2, which bodes well for futureoperation phases. Peak local heat loads onto the divertor were ingeneral benign and project below the 10 MW m?2 limit of thefuture water-cooled divertor when operated with 10 MW of heating power,with the exception of low-density attached operation in the high-iotaconfiguration. The most notable result was the complete (in all 10divertor units) heat-flux detachment obtained at high-density operationin hydrogen.

In this work, we present advances in the numerical modeling of the non-linear magnetoacoustic cyclotron instability (MCI) in relation to ion cyclotron emission (ICE) in fusion plasmas. We present the results of first 2D hybrid simulations of the MCI showing that only when excited waves propagate obliquely to the background magnetic field, the characteristic fingerprint of ICE in fusion plasmas is recovered. Different dynamics in 1D and 2D hybrid simulations of the MCI are observed and described. The results in the present work shed some light on the self-consistent dynamics of the non-linear MCI in scenarios where no analytical theory is available.

Inertial confinement fusion, based on laser-heating a deuterium–tritium mixture, is one of the approaches towards energy production from fusion reactions. Now, record energy-yield experiments are reported—bringing us closer to ignition conditions.

A magnetic field configuration of an island divertor for a quasi-axisymmetric stellarator (CFQS) is proposed. The configuration incorporates large islands surrounding the core confinement region. The interface between the core region and the peripheral region of the island divertor is a clear magnetic separatrix similar to a tokamak divertor. The structure of divertor magnetic field lines is very regular without stochasticity and the connection length is sufficiently long for good divertor performance. Such a divertor configuration is produced in the magnetic field configuration for the CFQS device, which is now under construction in China.

Since impurities determine the plasma performance in magnetically-confined fusion plasmas, analyzing their behavior and controlling their confinement is significantly important for a stable and steady-state operation of fusion devices. A comparison of the impurity transport in the core region of large stellarators, such as Wendelstein 7-X and Large Helical Device (LHD) is quite useful to gain a better understanding of the impurity transport in reactor-relevant stellarator plasmas. Therefore, a Tracer-Encapsulated Solid Pellet (TESPEL) injection system has been newly installed for the operational phase OP1.2b of W7-X. The injector was designed and manufactured by NIFS to achieve a similar TESPEL deposition location as in LHD. Technical guidelines and functional specifications, addressing the specific conditions at W7-X had been implemented in close collaboration between NIFS and IPP. This resulted in a compact 3-stage differential pumping system considering spatial restrictions due to existing and future diagnostics, material requirements and limitations due to the magnetic stray field in the vicinity of the turbo pumps. This contribution reports on the design and the integration of the new TESPEL injection system to W7-X and first achievements during the commissioning of the system.

The confinement of the guiding-centre trajectories in a stellarator is determined by the variation of the magnetic field strength $B$ in Boozer coordinates $(r,\unicode[STIX]{x1D703},\unicode[STIX]{x1D711})$, but $B(r,\unicode[STIX]{x1D703},\unicode[STIX]{x1D711})$ depends on the flux surface shape in a complicated way. Here we derive equations relating $B(r,\unicode[STIX]{x1D703},\unicode[STIX]{x1D711})$ in Boozer coordinates and the rotational transform to the shape of flux surfaces in cylindrical coordinates, using an expansion in distance from the magnetic axis. A related expansion was done by Garren and Boozer (Phys. Fluids B, vol. 3, 1991a, 2805) based on the Frenet–Serret frame, which can be discontinuous anywhere the magnetic axis is straight, a situation that occurs in the interesting case of omnigenity with poloidally closed $B$ contours. Our calculation in contrast does not use the Frenet–Serret frame. The transformation between the Garren–Boozer approach and cylindrical coordinates is derived, and the two approaches are shown to be equivalent if the axis curvature does not vanish. The expressions derived here help enable optimized plasma shapes to be constructed that can be provided as input to VMEC and other stellarator codes, or to generate initial configurations for conventional stellarator optimization.

There has been a growing interest, over the past few years, on understanding the effect on radio frequency waves due to turbulence in the scrape-off layer of tokamak plasmas. While the far scrape-off layer density width is of the order of centimetres in contemporary tokamaks, in ITER (International Thermonuclear Experimental Reactor), and in future fusion reactors, the corresponding width will be of the order of tens of centimetres. As such, this could impact the spectral properties of the waves and, consequently, the transport of wave energy and momentum to the core plasma. The turbulence in the scrape-off layer spans a broad range of spatial scales and includes blobs and filaments that are elongated along the magnetic field lines. The propagation of radio frequency waves through this tenuous plasma is given by Maxwell’s equations. The characteristic properties of the plasma appear as a permittivity tensor in the expression for the current in Ampere’s equation. This paper develops a formalism for expressing the permittivity of a turbulent plasma using the homogenization technique. This technique has been extensively used to express the dielectric properties of composite materials that are spatially inhomogeneous, for example, due to the presence of micro-structures. In a similar vein, the turbulent plasma in the scrape-off layer is spatially inhomogeneous and can be considered as a composite material in which the micro-structures are filaments and blobs. The classical homogenization technique is not appropriate for the magnetized plasma in the scrape-off layer, as the radio frequency waves span a broad range of wavelengths and frequencies – from tens of megahertz to hundreds of gigahertz. The formalism in this paper makes use of the Fourier space components of the electric and magnetic fields of the radio frequency waves for the scattered fields and fields inside the filaments and blobs. These are the eigenvectors of the dispersion matrix which, using the Green’s function approach, lead to a homogenized dielectric tensor.

This document is the product of a stellarator community workshop, organized by the National Stellarator Coordinating Committee and referred to as Stellcon, that was held in Cambridge, Massachusetts in February 2016, hosted by MIT. The workshop was widely advertised, and was attended by 40 scientists from 12 different institutions including national labs, universities and private industry, as well as a representative from the Department of Energy. The final section of this document describes areas of community wide consensus that were developed as a result of the discussions held at that workshop. Areas where further study would be helpful to generate a consensus path forward for the US stellarator program are also discussed. The program outlined in this document is directly responsive to many of the strategic priorities of FES as articulated in “Fusion Energy Sciences: A Ten-Year Perspective (2015–2025)” [1]. The natural disruption immunity of the stellarator directly addresses “Elimination of transient events that can be deleterious to toroidal fusion plasma confinement devices” an area of critical importance for the US fusion energy sciences enterprise over the next decade. Another critical area of research “Strengthening our partnerships with international research facilities,” is being significantly advanced on the W7-X stellarator in Germany and serves as a test-bed for development of successful international collaboration on ITER. This report also outlines how materials science as it relates to plasma and fusion sciences, another critical research area, can be carried out effectively in a stellarator. Additionally, significant advances along two of the Research Directions outlined in the report; “Burning Plasma Science: Foundations—Next-generation research capabilities”, and “Burning Plasma Science: Long pulse—Sustainment of Long-Pulse Plasma Equilibria” are proposed.

Controlling nonlinear phenomena is essential in many real-world applications. This paper explores the possibility to design novel nonlinear control strategies for nuclear fusion devices. In particular, a nonlinear phenomena relevant for nuclear fusion research is analyzed and two nonlinear control approaches, namely thermostat control and pellet injection, are opportunely designed.

Starting from expressions in Connor et al. (1988) [1], we derive a one-dimensional tearing equation similar to the approximate equation obtained by Hegna and Callen (1984) [2] and by Nishimura et al (1998) [3], but for more realistic toroidal equilibria. The intention is to use this approximation to explore the role of steep profiles, bootstrap currents and strong shaping in the vicinity of a separatrix, on the stability of tearing modes which are resonant in the H-mode pedestal region of finite aspect ratio, shaped cross-section tokamaks, e.g. JET. We discuss how this one-dimensional model for tearing modes, which assumes a single poloidal harmonic for the perturbed poloidal flux, compares with a model that includes poloidal coupling by Fitzpatrick et al (1993) [4]., 10 pages

We start our brief journey across the world of complex behaviours in plasmas by focusing on those confined by magnetic fields in order to produce fusion energy on Earth. The field is often referred to as magnetic confinement fusion (MCF). For those readers unfamiliar with it, we will start by providing a broad overview of MCF. Those already acquainted with the field are more than welcomed to move directly to Sect. 6.4, where the basic physics that appear to be responsible for one of the more important complex behaviours observed in MCF plasmas will be discussed.

Breit and Wheeler suggested that it should be possible to turn light into matter by smashing together only two particles of light (photons), to create an electron and a positron – The simplest method of turning light into matter ever predicted.

The basic physics of the processes playing the most important role in divertor plasma detachment is reviewed. The models used in the two-dimensional edge plasma transport codes that are widely used to address different issues of the edge plasma physics and to simulate the experimental data, as well as the numerical schemes and convergence issues, are described. The processes leading to ultimate divertor plasma detachment, the transition to and the stability of the detached regime, as well as the impact of the magnetic configuration and divertor geometry on detachment, are considered. A consistent, integral physical picture of ultimate detachment of a tokamak divertor plasma is developed.

The collisionless axisymmetric zonal flow residual calculation for a tokamak plasma is generalized to include electromagnetic perturbations. We formulate and solve the complete initial value zonal flow problem by retaining the fully self-consistent axisymmetric spatial perturbations in the electric and magnetic fields. Simple expressions for the electrostatic, shear and compressional magnetic residual responses are derived that provide a fully electromagnetic test of the zonal flow residual in gyrokinetic codes. Unlike the electrostatic potential, the parallel vector potential and the parallel magnetic field perturbations need not relax to flux functions for all possible initial conditions.

A dual-frequency gyrotron allowing operation at either 104 GHz or 140 GHz has been developed for use in plasma heating experiments at the W7-X stellarator at IPP Greifswald. The gyrotron has been fabricated, tested, and shipped to the installation site, where long-pulse conditioning is currently in progress (as of January, 2017).

The first results of particle-in-cell simulations of the electrostatic sheath and magnetic pre-sheath of thermionically emitting planar tungsten surfaces in fusion plasmas are presented. Plasma con ...

Korean activities on fusion licensing and safety studies have been going through three phases. The first phase started with the licensing of the Korean Superconductor Tokamak Advanced Research (KSTAR) classified as a radiation-generating device on the basis of the existing nuclear regulations. To accommodate the specific characteristics of a tokamak device that is to use deuterium in fusion plasma physics studies, a set of instructions to examine the safety of KSTAR and to approve an operating permit were developed and issued by the regulatory body, Korean Institute of Nuclear Safety. The second phase is being investigated for the International Thermonuclear Experimental Reactor (ITER). To verify the design of the test blanket module that will be installed and tested in the ITER, safety studies were carried out using the experimental results and computational codes. To minimize occupational irradiation damages, a study on the tritium behaviors is being carried out as Korea will provide a part of the tritium systems of ITER. The last phase has initiated to recognize significant safety issues and to support future licensing process for a fusion DEMO plant. Relevant studies are being carried out in collaboration with the academies funded by the National Fusion Research Institute of Korea. These front-end studies include the conceptual analysis on the safety and radiological consequences, preliminary studies on the layout of research, and on the future regulatory requirements of the fusion DEMO plant. This paper summarizes these activities on fusion safety with the findings of the front-end studies to date. It also proposes a pathway for the novel licensing process of a fusion DEMO plant with the tokamak concept.

A stellarator built by an international team of researchers demonstrate reduced plasma energy losses and optimized fusion confinement performance.

Fusion plasmas are complex systems which involve a variety of physical processes interacting with each other across wide ranges of spatiotemporal scales. In the National Institute for Fusion Science (NIFS), we are utilizing the full capability of the supercomputer (Plasma Simulator) and propelling domestic and international collaborations in order to conduct the Numerical Simulation Reactor Research Project (NSRP). Understanding physical mechanisms of complex plasma phenomena for the systematization of fusion science, NSRP aims at realization of the Numerical Helical Test Reactor, which is an integrated system of simulation codes to predict behaviors of fusion plasmas over the whole machine range. In NSRP, eight task groups are organized to cover a wide range of fusion simulation subjects: plasma fluid equilibrium stability, energetic-particle physics, integrated transport simulation, neoclassical and turbulent transport simulation, peripheral plasma transport, plasma-wall interaction, multi-hierarchy physics, and simulation science basis. Verification and validation researches are in progress in these task groups collaborating with each other as well as with experimental and engineering groups. Successful examples of validations of large-scale simulations of energetic particle driven instabilities and neoclassical and turbulent transport against experimental results from tokamaks and helical systems are highlighted. In addition, recent achievements in advanced simulation studies on ion heating processes and plasma-wall interactions, as well as those in the application of Virtual-Reality (VR) technology to fusion engineering, are presented.

Validation of physics models using the newly uncovered physics with a 2-D electron cyclotron emission imaging (ECEi) system for magnetic fusion plasmas has either enhanced the confidence or substantially improved the modeling capability. The discarded “full reconnection model” in sawtooth instability is vindicated and established that symmetry and magnetic shear of the 1/1 kink mode are critical parameters in sawtooth instability. For the 2/1 instability, it is demonstrated that the 2-D data can determine critical physics parameters with a high confidence and the measured anisotropic distribution of the turbulence and its flow in presence of the 2/1 island is validated by the modelled potential and gyro-kinetic calculation. The validation process of the measured reversed-shear Alfvéneigenmode (RSAE) structures has improved deficiencies of prior models. The 2-D images of internal structure of the ELMs and turbulence induced by the resonant magnetic perturbation (RMP) have provided an opportunity to establish firm physics basis of the ELM instability and role of RMPs. The importance of symmetry in determining the reconnection time scale and role of magnetic shear of the 1/1 kink mode in sawtooth instability may be relevant to the underlying physics of the violent kink instability of the filament ropes in a solar flare.

The operation of ITER requires high-quality estimates of the plasma electron density over multiple regions in the plasma for plasma evaluation, plasma control and machine protection purposes. Although the density regimes of ITER are not very different from those of existing tokamaks (1018–1021 m−3), the severe conditions of the fusion plasma environment present particular challenges to implementing these density diagnostics. In this paper we present an overview of the array of ITER electron density diagnostics designed to measure over the entire ITER domain: plasma core, pedestal, edge, scrape-off layer and divertor. It will focus on the challenges faced in making these measurements, and the technical solutions of the current designs.

Large edge-localized mode (ELM) control techniques must be developed to help ensure the success of burning and ignited fusion plasma devices such as tokamaks and stellarators. In full performance ITER tokamak discharges, with QDT = 10, the energy released by a single ELM could reach ∼30 MJ which is expected to result in an energy density of 10–15 MJ/m2on the divertor targets. This will exceed the estimated divertor ablation limit by a factor of 20–30. A worldwide research program is underway to develop various types of ELM control techniques in preparation for ITER H-mode plasma operations. An overview of the ELM control techniques currently being developed is discussed along with the requirements for applying these techniques to plasmas in ITER. Particular emphasis is given to the primary approaches, pellet pacing and resonant magnetic perturbation fields, currently being considered for ITER.

The presence of dust particles in a fusion plasma is recognized as a serious issue for safe and efficient operation of the ITER tokamak. The paper presents an in situ laser assisted method for characterization of dust from thermal emission of the particles. The method was developed in the TEXTOR tokamak with the use of Thomson scattering (TS). The diagnostic is capable to detect single particles and measure the dust density profile along the laser probing axis, velocity distribution of dust particles along this axis as well as surface temperature and size of the detected particles.