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We report a DIII-D database study of the H-mode power threshold over a wide range of plasma conditions and in the presence of resonant magnetic perturbations (RMPs). This database consisting of global (i.e. averaged) quantities is first compared to the 2008 ITPA database and the resulting L-H power scaling [38]. Since ELM control is critical for ITER, and applied 3D fields will likely be present prior to the transition into the H-mode, this study is important for assessing the impact of RMP ELM suppression on the L-H power threshold. The L-H transition has been studied extensively and is dependent on the physical and magnetic divertor geometry, shear flows, and drifts, among other parameters, some of which are altered by RMP fields. In order to understand the effects of RMPs on the L-H threshold, we attempt to make a robust empirical model, using only DIII-D data, that includes magnitudes and the toroidal modes of various resonant and non-resonant 3D fields. In addition, we assess the validity of previous assumptions about fast ion losses, as well as the usefulness of 0-D database regressions for extrapolation to ITER. Results from this database study show the standard 0-D parameters to be insufficient for capturing the complex L-H transition physics at a level high enough to provide an extrapolation to ITER with reasonable certainty.

Stickiness refers to the destructed orbits in phase space that remain in a particular region around intact surfaces for a long time before entering the chaotic sea. In fusion plasmas, stickiness regions trap magnetic field lines for many toroidal turns and have considerable effect on plasma transport. We locate stickiness regions in the chaotic phase space of resonant magnetic perturbations in tokamak using three dynamical quantities which are field lines diffusion coefficient, maximal Lyapunov exponent and the growing rate of an infinitesimal line segment. We compare the values of these quantities in regular, stickiness and chaotic regions to find a criterion for distinguishing stickiness from chaotic and regular areas in phase space.

Monte Carlo orbit-following simulations based on a phase-space coordinate transform method are presented in this paper. Guiding-center orbits are computed by using this method, that is, using the phase-space coordinate transform and solving the gyrokinetic equations of motion. This method has been adopted in a Monte Carlo orbit-following code GYCAVA for studying the behavior of fast ions in a tokamak. The finite Larmor radius effect can be included in the guiding-center orbit computation. The cylindrical coordinates are used for avoiding singularity of the safety factor at X points of the divertor configuration. An n -point discrete sum method is used to compute the gyro-average, which is related to the finite Larmor radius effect. The GYCAVA code has been integrated with other equilibrium, magnetohydrodynamic and neutral beam injection codes for studying losses and wall loads of fast ions in the presence of perturbation fields. The GYCAVA code has been verified by tests of the orbit computation and the Monte Carlo Coulomb collision. Simulations of fast ion losses in the presence of resonant magnetic perturbations in the EAST tokamak have been performed.

Charged particle motion in the regions of magnetic islands and near-separatrix ergodicity of magnetic field lines in the Т-15 tokamak is studied numerically. The trajectories are calculated by integration of exact three-dimensional equations of motion for trapped and passing charged particles starting at different pitch angles. It is shown that the presence of resonant magnetic perturbations qualitatively affects the trajectories of passing particles. In the region of magnetic islands, the orbits of passing particles acquire an island structure, while in the region of the near-separatrix ergodicity of magnetic field lines, the orbits are stochastized. It is shown that magnetic perturbations insignificantly affect trapped particle motion. Even in the absence of magnetic surfaces, the trajectories of trapped particles are regular and their cross sections have the shape of a standard banana orbit. The possibility of charged particles to pass between regions with different magnetic topologies is analyzed.

It is well known that heat exchange between the hot plasma and solid wall of a magnetic fusion reactor (tokamak or stellarator) depends, to a great extent, on local disturbances of its magnetic configuration, which occur under the action of resonant magnetic perturbations. One possible reason for their development in tokamaks is splitting of an originally axisymmetric current channel into separate filaments. Among negative consequences of such filamentation, there are spatial modulations of the plasma density and temperature, which can considerably increase local heat loads on the tokamak intrachamber elements. This study draws attention to the fact that one of the most dangerous manifestations of current filamentation in tokamaks is the formation of so-called “positive” magnetic islands, which are clearly observed during the development of internal disruptions in the form of “hot spots”—local hot dense helical plasma structures formed in the central region of the plasma column just before the major disruption. An increase in the heating efficiency and improvement of plasma confinement within magnetic islands (which are, in fact, closed magnetic configurations) could explain energy balance in such structures. In this article, stability of such structures and possible sources of initial current disturbances initiating their formation are discussed as illustratively as possible on the basis the initial physical principles, which, in the author’s opinion, can be useful for the experimenters who start solving problems of plasma–wall heat exchange under real tokamak conditions. A number of the well-known experimental observations that can be explained in the framework of this concept are discussed. It is argued that the concept of positive magnetic islands makes it possible to interrelate the sequence of MHD events resulting in the tokamak major disruption.

Plasma response to resonant magnetic perturbations (RMPs) near the pedestal top is crucial for accessing edge localized modes (ELMs) suppression in tokamaks. Since radial location of rotation zero-crossing plays a key role in determining the threshold for field penetration of RMP, plasma response may be different in low input torque plasmas. In this work, the linear MHD code MARS-F is applied to reveal the dependence of plasma response to RMP on rotation zero-crossing by a scan of rotation profiles based on an EAST equilibrium. It is shown that the plasma response is enhanced when zero-crossing occurs near rational surfaces. The dependence of plasma response on the location of rotation zero-crossing is well fitted by a double Gaussian, indicating two effects in this enhancement. One is induced by rotation screening effect shown as a wide base (with a width around 10-20 krad/s), and the other is related to resistive singular layer effect characterized by a localized peak (with a width around 3-4 krad/s). The width of the peak scales with the resistive singular layer width. The plasma displacement suggests the response is tearing like when zero-crossing is within the singular layer, while it is kink like when zero-crossing is far from the layer. The enhancement of magnetic islands width at the peak is only around a factor of two, when the absolute value of local rotation is not larger than 10-20 krad/s. It is further confirmed in a modeling of plasma response in an EAST ELM suppression discharge. Though there is a zero-crossing in $E\times B$ rotation but not in electron perpendicular rotation, no significant difference in plasma response is obtained using these two rotation profiles. This suggests that the rotation near pedestal top should not be far away from zero but may not be necessary to have zero-crossing for accessing ELM suppression., 9 pages, 9 figures, submitted to Physics of Plasmas

The confinement of neutral beam injection (NBI) particles in the presence of n = 3 resonant magnetic perturbations (RMPs) in 15 MA ITER DT plasmas has been studied using full orbit ASCOT simulations. Realistic NBI distribution functions, and 3D wall and equilibria, including the plasma response to the externally applied 3D fields calculated with MARS-F, have been employed. The observed total fast-ion losses depend on the poloidal spectra of the applied n = 3 RMP as well as on the absolute toroidal phase of the applied perturbation with respect to the NBI birth distribution. The absolute toroidal phase of the RMP perturbation does not affect the ELM control capabilities, which makes it a key parameter in the confinement optimization. The physics mechanisms underlying the observed fast-ion losses induced by the applied 3D fields have been studied in terms of the variation of the particle canonical angular momentum (δP ϕ ) induced by the applied 3D fields. The presented simulations indicate that the transport is located in an edge resonant transport layer as observed previously in ASDEX upgrade studies. Similarly, our results indicate that an overlapping of several linear and nonlinear resonances at the edge of the plasma might be responsible for the observed fast-ion losses. The results presented here may help to optimize the RMP configuration with respect to the NBI confinement in future ITER discharges.

The studies on tungsten (W) sputtering and transport characteristics on EAST tokamak under resonant magnetic perturbation (RMP) fields have been performed with the three-dimensional edge transport code EMC3-EIRENE. The estimation of the W sputtering flux has been carried out based on EMC3-EIRENE modeling for RMP and no RMP applications. The W sputtering flux shows a small difference between RMP and no RMP cases due to weak sensitivity of W sputtering yield on deuterium impact energy for low edge plasma density. However, for high edge plasma density, the strong dependence of W sputtering yield on deuterium impact energy comes into play, which results in a remarkable difference in the W sputtering flux between RMP and no RMP cases. Impacts of the W impurity perpendicular transport on W ions transport have been investigated. With low W perpendicular transport coefficient for RMP case, the edge plasma has a better W ions screening effect for high edge plasma density. The increased W perpendicular transport coefficient results in a good W ions screening effect for both high and low edge plasma densities.

The effect of externally applied resonant magnetic perturbations (RMPs) on the local equilibrium plasma current density profile is studied numerically based on two-fluid equations in simplified cylindrical geometry. It is found that a moderate RMP below its penetration threshold, via non-linear mode coupling, induces a parallel electric field around its rational surface that can significantly change the local flux-surface-averaged current density gradient. At a given RMP amplitude, the modification of the current density profile increases with increasing electron temperature, and it significantly depends on the bi-normal electron fluid velocity at the resonant surface. The effect of this modification on the magnetic island growth is demonstrated by the example of small m/n = 2/1 islands (m/n being the poloidal/toroidal mode numbers), driven by an unfavorable plasma current density profile and bootstrap current perturbation. The 2/1 mode growth is stabilized by moderate static 4/2 or 6/3 RMPs if the local electron fluid velocity is in the ion drift direction or sufficiently large in the electron drift direction. These results reveal that a weakly three-dimensional equilibrium, containing a moderate 4/2 RMP and the associated shielding current, can be more stable against the 2/1 mode, which often causes tokamak plasma major disruptions.

Type I Edge Localized Modes (ELMs) occur naturally in H-mode plasmas, the operational regime envisaged for ITER, and lead to high temporal heat load peaks on plasma-facing components that might induce surface melting and deteriorated material properties due to recrystallization. ELM mitigation techniques are being developed that either attempt to fully suppress the ELM, e.g. using resonant magnetic perturbations (RMP), or to trigger the ELM at higher frequencies, intending to reduce its energy content and hence the heat load peak. In this paper, the effect of the increased ELM frequency achieved by ELM mitigation techniques on the ITER tungsten divertor monoblock thermal and recrystallization behavior is numerically analyzed by solving the unsteady heat conduction equation when repeatedly exposing the monoblock to ELM-like heat loads under ITER baseline burning plasma operating conditions. The impinging heat load is based on a set of empirical relations readily found in literature and the recrystallization fraction is computed using an empirical data fit which is based on the non-isothermal JMAK equation as function of temperature and time. Uncertainties such as the statistical spread on empirical relations and the unknown inter-ELM heat load, as well as the effect of the ELM frequency, are quantified by a parameter scan. Adopting the recently developed concept of a ‘recrystallization budget’ to determine the allowed monoblock heat loads [G. De Temmerman et al., PPCF 60 (2018) 0044018], it was found that ELM frequencies between 25 – 50 Hz are allowed for inter-ELM heat loads between 9 – 6 MW m − 2 , respectively, when assuming that the scaling for the peak ELM target energy density in [T. Eich et al., NME 12 (2017) 84-90] holds. This provides a new limit for ELM mitigation strategies in ITER.

An upgrade to the 3D plasma boundary model EMC3-EIRENE for detached divertor plasmas is introduced and evaluated. Stabilization of the iterative approximation of a self-consistent plasma–neutral gas solution is required at low divertor plasma temperatures, and this has been achieved by linearization of the energy loss term associated with the electron–neutral gas interaction. An implicit method for volumetric recombination has been implemented, which improves agreement with SOLPS-ITER results for the pre-fusion power operation phase in ITER. Resonant magnetic perturbations for control of edge localized modes are found to result in a partially detached divertor plasma with a significantly broader reduction of particle loads in the traditional strike zone, which is attributed to volumetric recombination. The additional non-axisymmetric far scrape-off layer strike points, on the other hand, remain attached at relatively high temperature so that volumetric recombination is not effective there.

JOREK is a massively parallel fully implicit non-linear extended MHD code for realistic tokamak X-point plasmas. It has become a widely used versatile code for studying large-scale plasma instabilities and their control developed in an international community. This article gives a comprehensive overview of the physics models implemented, numerical methods applied for solving the equations and physics studies performed with the code. A dedicated section highlights some of the verification work done for the code. A hierarchy of different physics models is available including a free boundary and resistive wall extension and hybrid kinetic-fluid models. The code allows for flux-surface aligned iso-parametric finite element grids in single and double X-point plasmas which can be extended to the true physical walls and uses a robust fully implicit time stepping. Particular focus is laid on plasma edge and scrape-off layer (SOL) physics as well as disruption related phenomena. Among the key results obtained with JOREK regarding plasma edge and SOL, are deep insights into the dynamics of edge localized modes (ELMs), ELM cycles, and ELM control by resonant magnetic perturbations, pellet injection, as well as by vertical magnetic kicks. Also ELM free regimes, detachment physics, the generation and transport of impurities during an ELM, and electrostatic turbulence in the pedestal region are investigated. Regarding disruptions, the focus is on the dynamics of the thermal quench and current quench triggered by massive gas injection (MGI) and shattered pellet injection (SPI), runaway electron (RE) dynamics as well as the RE interaction with MHD modes, and vertical displacement events (VDEs). Also the seeding and suppression of tearing modes (TMs), the dynamics of naturally occurring thermal quenches triggered by locked modes, and radiative collapses are being studied., Comment: Comprehensive review published as special topic article in Nuclear Fusion

Numerical simulation of tearing mode suppression by applying non-resonant magnetic perturbations (non-RMPs) has been carried out in the J-TEXT tokamak based on the NIMROD (Non-Ideal Magnetohydrodynamics with Rotation, Open Discussion) code. In the present work, we investigate the growth of n = 1 and n = 2 (n is the toroidal mode number) magnetohydrodynamics (MHD) instabilities, mainly as tearing modes, in response to different modes and amplitudes of non-RMPs generated by the external helical coils. It is found that the tearing mode can be stabilized by non-RMPs, especially for the non-RMPs with mode m/n = -1/1 (m is the poloidal mode number) and coil current I MP = 4kA.

Experiments on DIII-D show that resonant magnetic perturbations (RMPs) reduce the impurity confinement time in the H-mode pedestal below that of naturally ELMy plasmas. STRAHL modeling of discharges with aluminum laser blow off (LBO) injection show increased diffusion in the pedestal of H-mode discharges with RMPs. The increased diffusion from the RMPs provides improved impurity control compared to the naturally ELMy discharges and prevents the buildup of impurities near the edge of the plasma. Tungsten LBO injection into discharges with and without RMPs led to striking differences in the evolution of the discharge. The naturally ELMy discharge became ELM free and showed an increase in edge radiation, leading to the eventual radiative collapse of the plasma. Injecting a similar number of W particles into discharges with RMPs, either with or without ELMs, led to a measurable accumulation of core W but did not significantly impact the evolution of the discharge. These results indicate that RMPs are beneficial for controlling the buildup of impurities in the pedestal region of the plasma.

A class of topological magnetic island bifurcations that has not previously been observed in toroidal plasmas is described. Increasing an externally applied three-dimensional magnetic field in resistive magnetohydrodynamic simulations results in the asymmetric elongation of resonant island flux surfaces followed by a sequence of heteroclinic bifurcations. These bifurcations produce new sets of hyperbolic-elliptic fixed points as predicted by the Poincare-Birkoff fixed point theorem. Field line calculations verify that the new fixed points do not connect to those of the prebifurcated islands as required for heteroclinic bifurcations on a torus with winding numbers composed of common integer factors.

Detailed 2D turbulence measurements from the DIII-D tokamak provide an explanation for how resonant magnetic perturbations (RMPs) raise the L-H power threshold $P_\textrm{LH}$ [P. Gohil et al., Nucl. Fusion 51, 103020 (2011)] in ITER-relevant, low rotation, ITER-similar-shape plasmas with favorable ion $\nabla B$ direction. RMPs simultaneously raise the turbulence decorrelation rate $\Delta \omega_D$ and reduce the flow shear rate $\omega_\textrm{shear}$ in the stationary L-mode state preceding the L-H transition, thereby disrupting the turbulence shear suppression mechanism. RMPs also reduce the Reynolds stress drive for poloidal flow, contributing to the reduction of $\omega_\textrm{shear}$ On the ~100 {\mu}s timescale of the L-H transition, RMPs reduce Reynolds-stress-driven energy transfer from turbulence to flows by an order of magnitude, challenging the energy depletion theory for the L-H trigger mechanism. In contrast, non-resonant magnetic perturbations, which do not significantly affect $P_\textrm{LH}$, do not affect $\Delta \omega_D$ and only slightly reduce $\omega_\textrm{shear}$ and Reynolds-stress-driven energy transfer., Comment: The following article has been submitted to Physics of Plasmas

Self-consistent simulations of neoclassical and electrostatic turbulent transport in a DIII-D H-mode edge plasma under resonant magnetic perturbations (RMPs) have been performed using the global total-f gyrokinetic particle-in-cell code x-point gyrokinetic code (XGC), in order to study density pump-out and electron heat confinement. The RMP field is imported from the extended magneto-hydrodynamics code M3D-C1, taking into account the linear two-fluid plasma response. With both neoclassical and turbulence physics considered together, the XGC simulation reproduces two key features of experimentally observed edge transport under RMPs: increased radial particle transport in the pedestal region that is sufficient to account for the experimental pump-out rate and suppression of the electron heat flux in the steepest part of the edge pedestal. In the simulation, the density fluctuation amplitude of modes moving in the electron diamagnetic direction increases due to interaction with RMPs in the pedestal shoulder and outward, while the electron temperature fluctuation amplitude decreases.

The electron cyclotron resonance heating (ECRH) system of the ASDEX Upgrade tokomak has been upgraded over the last 15 years from a 2 MW, 2 s, 140 GHz system to an 8 MW, 10 s, dual frequency system (105/140 GHz). The power exceeds the L/H power threshold by at least a factor of two, even for high densities, and roughly equals the installed ion cyclotron range of frequencies power. The power of both wave heating systems together (>10 MW in the plasma) is about half of the available neutral beam injection (NBI) power, allowing significant variations of torque input, of the shape of the heating profile and of Qe/Qi, even at high heating power. For applications at a low magnetic field an X3-heating scheme is routinely in use. Such a scenario is now also forseen for ITER to study the first H-modes at one third of the full field. This versatile system allows one to address important issues fundamental to a fusion reactor: H-mode operation with dominant electron heating, accessing low collisionalities in full metal devices (also related to suppression of edge localized modes with resonant magnetic perturbations), influence of Te/Ti and rotational shear on transport, and dependence of impurity accumulation on heating profiles. Experiments on all these subjects have been carried out over the last few years and will be presented in this contribution. The adjustable localized current drive capability of ECRH allows dedicated variations of the shape of the q-profile and the study of their influence on non-inductive tokamak operation (so far at q 95 > 5.3). The ultimate goal of these experiments is to use the experimental findings to refine theoretical models such that they allow a reliable design of operational schemes for reactor size devices. In this respect, recent studies comparing a quasi-linear approach (TGLF) with fully non-linear modeling (GENE) of non-inductive high-beta plasmas will be reported.

Microtearing turbulence in an idealized pedestal scenario is found to saturate via zonal fields, while also exciting strong zonal flows; a concurrent upshift of the non-linear critical gradient is observed. The zonal flows cause electron-temperature-gradient-driven turbulence to be ameliorated. When applying resonant magnetic perturbations, the prompt charge loss off the flux surface erodes the zonal flow, leading to higher electron-scale fluxes, while leaving microtearing saturation physics unaffected.

The effects of resonant magnetic perturbations on the turbulent transport of fast ions in tokamak devices are investigated using a theoretical transport model of test-particle type. The direct numerical simulation method is used to compute, via the transport model, the diffusion coefficients. The numerical results are in good agreement with other, analytically derived, estimations. It is found that finite Larmor radius effects algebraically decrease the transport, while the amplitude of magnetic perturbations has an opposite effect. In the presence of stochastic dynamics, the asymmetric toroidal magnetic field induces a small, radial, outward pinch. A synergistic mechanism of non-linear coupling between the effects of turbulence and those of magnetic perturbations enhances the radial diffusion. General scaling laws are proposed for the transport coefficients.

It is demonstrated that tokamak plasma can be fuelled by pellets while simultaneously maintaining ELM suppression by external resonant magnetic perturbations (RMPs). Pellets are injected vertically from high field site and deposited at outer part of plasma cross section. Each pellet triggers a benign MHD event followed by a short lived ELM-free phase. The ELM suppression phase with pellet fuelling lasts 11 pellet cycles and is terminated by intentionally increasing the pellet rate to cause a transition to the ELMy phase.

The modelling of edge carbon transport and emission on EAST tokamak under resonant magnetic perturbation (RMP) fields has been conducted with the three-dimensional edge transport code EMC3-EIRENE. The measured vertical distribution of CVI emission by the extreme ultraviolet spectrometer system for the perturbed case shows a reduction in the CVI emission by 20 % compared to the equilibrium case. The chord-integrated CVI emission can be reconstructed by EMC3-EIRENE modelling, which presents an increase in the CVI emission with RMP fields. The discrepancy between experiments and simulations has been investigated by parameter study to examine the sensitivity of the simulation results on the edge plasma conditions and the impurity perpendicular transport. It is found that the variation of edge plasma conditions for the equilibrium case cannot resolve the discrepancy in the CVI emission between simulations and measurements. The simulations with enhanced impurity perpendicular transport coefficient allows a reasonable agreement with the measured reduction of CVI emission.

Tearing modes (TMs) degrade the performance of tokamak plasma, and can even lead to disruption. Using externally exerted resonant magnetic perturbations (RMPs) to suppress TM is a promising and effective way. In order to suppress the 2/1 TM, 2/1 RMP must be applied in given phase regions to stabilize magnetic island and accelerate island rotation. The RMP feedback control system acquires the multiple-channel poloidalMirnov signal, processes the acquired data and calculates the magnetic island phase in real time, and switches on the RMP power supply when the phase is in the designated region. The feedback control system is based on NI cRIO and developed mainly using LabVIEW field-programmable gate array. The typical 2/1 mode magnetic island on J-TEXT rotates at a frequency from 2 to 10 kHz. To ensure the control precision within 2°, the control period must be within 500 ns. Due to the noisy characteristic of the Mirnov signal, the feedback control system uses a series of error correction methods in real time to obtain accurate phase. Various control strategies have been adopted to exploit the RMP coil and power supply. Observable result is presented in this paper.

In certain circumstances, plasma response suppresses magnetic islands expected at perturbed resonant magnetic surfaces. We investigate the plasma response to the resonant magnetic perturbations in a large aspect ratio tokamak perturbed by external resonant helical windings, considering polar toroidal coordinates for which analytical toroidal equilibrium solutions and perturbing fields are available. We apply an empirical approach to mimic the plasma screening effects by introducing presumed plasma current sheets on the resonance surfaces to cancel the RMP effects. Numerical examples show the effect of plasma response reducing magnetic islands at the plasma edge and also regularizing field lines around the resonant surface. The distribution of connection lengths along the plasma cross section indicates that the plasma response increases the connection lengths since more toroidal turns are performed until a field line reaches the tokamak wall.

A two-dimensional tomographic reconstruction method based on the Phillips-Tikhonov regularization is proposed in this study, for a tangentially viewing visible light imaging diagnostic on EAST. Phantom tests are demonstrated to evaluate the performance of this method. The purpose of this method is to reconstruct the local emission from experimental images in various parts of the visible spectrum, which can be employed to reflect spatial distributions of certain particles (Li-II in this study) in plasmas. As a potential application of the Li-II distribution, the measurement of plasma boundary displacements in the presence of resonant magnetic perturbations is tried. Though the result of this attempt is limited by the computational precision and conditions of diagnostic system at present, it provides valuable references for further works.

A space-resolved extreme ultraviolet (EUV) spectrometer has been used to observe the spatial distribution and temporal behavior of core impurity emissions at the experimental advanced superconducting tokamak (EAST), evaluating the effects of resonant magnetic perturbations (RMPs) on impurity concentration. The vertical profiles of impurity emissions from the 0 ≤ Z ≤ 450 mm range results in spectra in the wavelength range of 30 A $\le \lambda \le 500\,{\mathring {\mathrm {A}}}$ . Line emissions of impurity carbon, lithium, and iron were observed for C VI (33.7 A, 2p–1s), Li III (135 A, 2p–1s), and Fe XXIII (132.9 A, 1s22s2p–1s22s2), respectively, whereas the unresolved transition array (UTA) of tungsten in the wavelength range of $\lambda = 48$ to 70 A is monitored to study tungsten behavior with RMP. It is clearly observed that the intensities of C VI, Li III, Fe XXIII, and W-UTA emissions are reduced by the application of RMP fields, during stages of edge-localized modes (ELMs). In addition, the reduction in Fe and W emissions with $n = 1$ and 2 (n: toroidal mode number) RMP is more significant than that of carbon and lithium emissions, i.e., ~50% to 70% for Fe XXIII and W-UTA, and ~30% for C VI and Li III. Considering that the changes in $n_{e}$ and $T_{e}$ cannot account for the impurity emissions mitigation, the EUV measurements may indicate that RMP can significantly and selectively decrease the impurity accumulation in the core region with slight effects on the main plasma. Spatial distributions of impurity emissions are also measured for different impurity species using the EUV spectrometer, confirming the space-resolved diagnostic’s feasibility to study RMP effects on core impurity concentration.

Edge plasma properties have been modified in the Saskatchewan Torus-Modified tokamak by means of resonant magnetic perturbations (RMP). It has been found that the radial profiles of ion saturation ...

The structure of the non-axisymmetric heat load distribution at the divertor plates is determined not only by the toroidal but also from the poloidal spectrum of non-axisymmetric field perturbations. Whether they are intrinsic, like error fields, or they are applied through 3D coils, the non-axisymmetric fields produce complex 3D edge magnetic topologies (footprints) that alter the properties of the heat and particle flux distributions on the divertor target plates. In this manuscript, a study of the impact of applied 3D field poloidal spectrum on the footprint size and structure is done for the DIII-D tokamak using the resistive MHD code M3D-C1 coupled with the field line tracing code TRIP3D. To resolve the impact of the poloidal spectrum of the magnetic perturbation, the relative phase of the two rows of in-vessel 3D coils used to produce both a n = 2 and a n = 3 perturbation is varied, where n is the toroidal harmonic of the magnetic perturbation. This shows that the largest footprint is predicted when the relative phase of the two rows is close to zero, which is also where the resonant coupling with the plasma is maximized. These results suggest that it will be challenging to decouple the footprint size from the requisite resonant coupling for RMP–ELM control. The correlation between the measured heat load and particle flux distributions at the outer divertor plates in DIII-D and the magnetic measurements is in good agreement with the predicted dependence of the magnetic footprint size on the amplitude of the resonant component of the plasma response.

Systematic toroidal modeling of the plasma response to the n = 1–4 (n is the toroidal mode number) resonant magnetic perturbation (RMP) field is carried out in order to understand the plasma-shaping effect on controlling the type-I edge-localized modes (ELMs) in tokamak experiments. Considered are large variations of the plasma elongation and triangularity at a fixed edge safety factor q a, for limiter plasmas with both single-null (SN) and double-null (DN) divertor-like boundary shapes. Numerical results assuming conformal 3D RMP coils show that (i) the optimum coil phasing between the upper and lower rows for ELM control becomes increasingly sensitive to the plasma elongation with higher-n toroidal spectra, (ii) the optimum coil phasing is however essentially independent of the plasma triangularity for all n = 1–4 RMP fields, (iii) with the same coil current and the optimum coil phasing, high elongation generally favors ELM control but it may be more challenging for plasmas with intermediate elongation and with lower-n (n = 1–2) RMPs, and (iv) higher triangularity is generally always better for ELM control with all n = 1–4 fields for both DN and SN divertor-like plasma boundary shapes.

openaire: EC/H2020/633053/EU//EUROfusion The physics of the tokamak pedestal is still not fully understood, for example there is no fully predictive model for the pedestal height and width. However, the pedestal is key in determining the fusion power for a given scenario. If we can improve our understanding of reactor relevant pedestals we will improve our confidence in designing potential fusion power plants. Work has been carried out as part of a collaboration on reactor relevant pedestal physics. We report some of the results in detail here and review some of the wider work which will be reported in full elsewhere. First, we attempt to use a gyrokinetic-based calculation to eliminate the pedestal top density as a model input for Europed/EPED pedestal predictions. We assume power balance at the top of the pedestal, that is, the heat flux crossing the separatrix must be equal to the heat source at the top of the pedestal and investigate the consequences of this assumption. Unfortunately, the transport assumptions of the EPED model mean that this method does not discriminate between different pairs of density and temperature profiles for a given pressure profile. Second, we investigate the effects of non flux surface density on the bootstrap current. Third, type I ELMs will not be tolerable for a reactor relevant regime due to the damage that they are expected to cause to plasma facing components. In recent years various methods of running tokamak plasmas without large ELMs have been developed. These include small and no ELM regimes, the use of resonant magnetic perturbations and the use of vertical kicks. We discuss the quiescent H-mode here. Finally we give a summary and directions for future work.

For disruptions without mitigation, a large amount of thermal energy and poloidal magnetic energy will be dissipated inside the vacuum vessel (VV). A slow current quench may be accompanied by a large halo current, while a fast current quench often causes large eddy current, which will result in electromagnetic force. At the same time, fast current quench will induce strong toroidal electric field, which will result in a large fraction of runaway current and the hitting of runaway beam on first wall. The disruption mitigation is essential for large scale tokamak. The existing methods to mitigate disruptions, such as massive gas injection and resonant magnetic perturbations, are aimed at increasing the runaway generation threshold or the lose rate of runaway electrons. It may not work for ITER with Ip = 15 MA operation. The root of runaway generation is the large toroidal electric field induced by fast current quench and the large avalanche factor with high plasma current. The reduction of toroidal electric field is favor for the runaway suppression. The magnetic energy transfer (MET) based on electromagnetic coupling for disruption mitigation has been proposed on J-TEXT. It has the advantage of transferring the magnetic energy to outside of vessel by the electromagnetic coupling. It accelerates the current quench (CQ) rate and reduces the toroidal electric field at the same time. The runaway current has been suppressed by the MET system on J-TEXT. The experimental results show that the MET can reduce the energy dissipated in the VV by 20 % through transferring of energy to outside of VV. The MET can increase the CQ rate about 50.7 % and decrease the loop voltage about 35.3 %. The MET provides a new idea to transfer the magnetic energy and to suppress runaway current for disruption mitigation in future devices.