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The heat flux mitigation during the thermal quench (TQ) by the shattered pellet injection (SPI) is one of the major elements of disruption mitigation strategy for ITER. It's efficiency greatly depends on the SPI and the target plasma parameters, and is ultimately characterised by the heat deposition on to the plasma facing components. To investigate such heat deposition, JOREK simulations of neon-mixed dual-SPIs into ITER baseline H-mode and a 'degraded H-mode' with and without good injector synchronization are performed with focus on the first wall heat flux and its energy impact. It is found that low neon fraction SPIs into the baseline H-mode plasmas exhibit strong major radial plasmoid drift as the fragments arrive at the pedestal, accompanied by edge stochasticity. Significant density expulsion and outgoing heat flux occurs as a result, reducing the mitigation efficiency. Such drift motion could be mitigated by injecting higher neon fraction pellets, or by considering the pre-disruption confinement degradation, thus improving the radiation fraction. The radiation heat flux is found to peak in the vicinity of the fragment injection location in the early injection phase, while it relaxes later on due to parallel impurity transport. The overall radiation asymmetry could be significantly mitigated by good synchronization. Time integration of the local heat flux is carried out to provide its energy impact for wall heat damage assessment. For the baseline H-mode case with full pellet injection, melting of the stainless steel armour of the diagnostic port could occur near the injection port, which is acceptable, without any melting of the first wall tungsten tiles. For the degraded H-mode cases with quarter-pellet SPIs, which have 1 / 4 total volume of a full pellet, the maximum energy impact approaches the tolerable limit of the stainless steel with un-synchronized SPIs, and stays well below such limit for the perfectly synchronized ones. [ABSTRACT FROM AUTHOR]

The linear and nonlinear effects of hydrogen isotope species on ion temperature gradient (ITG) instability in the Large Helical Device (LHD) stellarator are studied using radially global gyrokinetic simulation. We found that the coupling range of linear toroidal harmonics depends on the ion mass of the hydrogen isotope. The growth rates of ITG mode are almost the same for H, D, and T plasmas, indicating a gyro-Bohm scaling of ion-mass dependence. The nonlinear electrostatic simulations show that the zonal flow breaks the radially elongated eigenmode structures and reduces the size of the turbulence eddies, which suppresses the turbulence and the ion heat transport in the LHD. The turbulence amplitude without the zonal flow is almost the same for H, D, and T plasmas, while it decreases with increasing ion mass of the hydrogen isotope when the zonal flow is present. The reduction of the turbulent transport with larger ion mass is mostly due to the enhancement of zonal flows by larger ion mass. The ion heat conductivity deviates from the gyro-Bohm scaling for both cases with and without the zonal flow. [ABSTRACT FROM AUTHOR]

The effect of different off-axis energetic particle (EP) slowing down distribution on beta-induced Alfvén eigenmode (BAE), driven by the on-axis EP distribution, is systematically studied using kinetic-magnetohydrodynamic code M3D-K. The aim is to analyze the optimal parameter region for controlling AEs via varying EP distribution parameters. The simulation results reveal that by modifying the gradients of the EP distribution, the off-axis EP can further destabilize or mitigate the on-axis EP driven BAE, depending on the off-axis EP distribution's parameters: deposition profile, EP beta, pitch angle, injection velocity and direction. When the off-axis EP is deposited outside the mode center, and its injection velocity is sufficiently large to satisfy the resonance with BAE, the stabilization of BAE is achieved. This stabilizing effect is directly proportional to the off-axis EP beta, while excessive off-axis EP beta can trigger a new EP-driven instability located outside the BAE. Furthermore, to achieve a stronger stabilizing effect, the pitch angle distribution and velocity direction of the off-axis EP should be close to those of the on-axis EP. For instance, compared to the off-axis counter-passing EP, the off-axis co-passing EP can lead to a more effective mitigation of the BAE driven by the on-axis co-passing EP. [ABSTRACT FROM AUTHOR]

We perform a systematic simulation study of energetic passing particle-driven instabilities in KSTAR using the kinetic-MHD hybrid code M3D-K. Linear simulation results show that the observed n = 1 mode in the early phase of the discharge is the low-frequency fishbone driven by energetic passing beam ions. The mode frequency computed is in a good agreement with the experimental measurement. Nonlinear simulations show that the frequency of the n = 1 mode jumps up to a higher value corresponding to the β -induced Alfvén eigenmode (BAE). In the later phase of the discharge, the simulated n = 5 mode is identified as a BAE in its linear phase. In the nonlinear phase, the n = 5 mode exhibits a similar frequency jump to a higher value of an energetic particle (EP) mode after mode saturation. Analysis of perturbed beam ion distributions in phase space shows that these new modes in nonlinear stages are driven by new resonances due to nonlinearly evolved beam ion distributions. Further simulations of a beam beta scan for the n = 5 mode show that the frequency jump disappears for a sufficiently small beam beta or beam ion drive. This result may explain the non-existence of frequency jump in the experiment. Finally, the impact of toroidal rotation on mode characteristics is investigated, showing that it has a marginal influence on EP driven modes. [ABSTRACT FROM AUTHOR]

The centrifugal force effects from toroidal rotation in improved confinement plasmas are analyzed on high- Z impurities in tokamaks. Tungsten (W) transport simulations are performed using the impurity transport code developed in the integrated code TASK. The geometric factors PA and PB are introduced into the neoclassical transport coefficients to include the effects of the toroidal rotation, which come from poloidal asymmetry in the high- Z impurity profiles. Inward neoclassical particle pinch driven by the main ion density gradient is enhanced by the poloidal asymmetry to be the dominant mechanism for W accumulation in the plasma central region. Simulations with experimental plasma profiles show good agreement with the experimental results and first-principle simulation results in the H-mode. In the hybrid mode and advanced mode, the impurity accumulation is enhanced in the internal transport barrier (ITB) regions. The condition to suppress impurity accumulation is investigated by calculating dependencies on the toroidal rotation velocity and ITB position. The neoclassical transport is sufficiently small with the prospected ITER condition of the Mach number of main ions Mi ∼ 0.1. The impurity transport inside the ITB is strongly influenced by competition between the density peaking effect and the temperature screening effect, and the present simulations show suppression of the impurity accumulation with the outer ITB position to improve the plasma performance, due to the relatively larger temperature gradient of the main ion. [ABSTRACT FROM AUTHOR]

We investigated the degradation of neutral beam (NB) fast-ion confinement depending on the NB power without magnetohydrodynamics instabilities in the Large Helical Device (LHD). In the LHD deuterium experiment, the neutron emission rate per NB power decreased by up to 20% with increasing injected NBs during a single discharge. Because there were no significant variations in the electron temperature and density, the NB shine-through rate, or the magnetic fluctuation due to the change in NB power, the reduction in the neutron emission rate indicates the degradation of the fast-ion confinement. In this paper, we formulated this degradation depending on the NB power and quantitatively estimated the degraded effective confinement time. In addition, we performed neutron emission rate simulations using the obtained effective confinement time. The simulation and experimental results were in good agreement, suggesting that the degraded effective confinement time is valid. [ABSTRACT FROM AUTHOR]

The three-dimensional variation inherent to stellarator geometries and fusion sources motivates three-dimensional modeling to obtain accurate results from computational modeling in support of design and analysis of first wall, blanket, and shield (FWBS) systems. Manually constructing stellarator fusion power plant geometries in computer-aided design (CAD) and defining the corresponding fusion source can be cumbersome and challenging. The open-source parametric modeling toolset ParaStell has been developed to automate construction of such geometries in low-fidelity. Low-fidelity modeling is useful during the conceptual phase of engineering design as a means of rapidly exploring the design space of a given device. The modeling capability of ParaStell includes in-vessel components and magnets, for any given stellarator configuration, using a parametric definition and plasma equilibrium data. Furthermore, the toolset automates the generation of detailed, tetrahedral neutron source definitions and DAGMC geometries for use in neutronics modeling. ParaStell assists rapid design iteration, parametric study, and design optimization of stellarator fusion cores. As a demonstration of the design iteration capability, the effect of the three-dimensional parameter space on tritium breeding and magnet shielding is investigated, using the WISTELL-D configuration as a design basis. Blanket and shield thicknesses are varied in three dimensions, using the space available between the plasma edge and magnet coils as a constraint. The corresponding effects on tritium breeding ratio and magnet heating are tallied using the open-source Monte Carlo particle transport code OpenMC. The inclusion of additional and higher-fidelity modeling capabilities is planned for ParaStell's future, as well as its implementation in machine-driven optimization. [ABSTRACT FROM AUTHOR]

This study is motivated by experiments on Tore Supra and WEST tokamaks where a deepening of the E × B velocity—governed by the radial electric field Er —near the edge is observed when the safety factor decreases in L-mode plasmas. Flux-driven global simulations of ion temperature gradient turbulence recover qualitatively the trend observed in the experiments, i.e. the E × B velocity increases when decreasing the safety factor. From these simulations, multiple clues point out the role of turbulence in the establishment of the radial electric field even though the turbulent intensity increases with the safety factor. The proposed mechanism to elucidate this phenomenon, backed up by a reduced model, is that the damping of the poloidal flow, governed by the neoclassical friction, increases more strongly with the safety factor than the turbulent drive for Er, due to the (r , θ) component of the Reynolds stress. [ABSTRACT FROM AUTHOR]

The confinement of energetic particles (EPs) generated by fusion reactions and external heating methods is crucial for the performance of future fusion devices. However, EP transport can occur due to their interaction with electromagnetic perturbations, affecting heating efficiency and overall performance. Robust reduced models are needed to analyze stability and transport. This paper presents an automated IMAS-based workflow for analyzing the time-dependent stability of EP-driven modes, focusing on the linear properties of Toroidal Alfvén Eigenmodes (TAEs) in general tokamak geometry. The workflow utilizes efficient computational methods and reduced models to deliver fast and reproducible results. A demonstration of the workflow's effectiveness was performed, identifying key linear properties of TAEs in various simulated ITER scenarios. This approach represents a critical step toward developing tools for analyzing EP transport and optimizing the performance of future fusion reactors. [ABSTRACT FROM AUTHOR]

A significant impact of pressure anisotropy on the plasma displacement associated with ideal Magnetohydrodynamic (MHD) stability is found. A heliotron plasma, such as a large helical device plasma, is analyzed. Simulations are performed using the equilibrium solver Anisotropic Neumann Inverse Moments Equilibrium Code and the ideal MHD stability code TERPSICHORE. Both codes provide a treatment of the pressure anisotropy by the bi-Maxwellian model. The ratio of hot particle pressure to total pressure < β h > / < β > has been scanned over. Other simulation parameters have been chosen such that the simulations represent an experimentally relevant condition with an external, off-axis heating scheme. The radial location of the peak of the plasma displacement of the n = 1, m = 2 mode number has been compared to the radial location of the ι = 0.5 resonant surface. This comparison shows that this difference in location increases monotonically for increasing < β h > / < β >. These results provide insight on the effect of external heating schemes on MHD stability. [ABSTRACT FROM AUTHOR]

The dimensionless isotope mass scaling experiment between pure Deuterium and pure Tritium plasmas with matched ρ ∗ , ν ∗ , β n , q and T e / T i has been achieved in JET L-mode with dominant electron heating (NBI+ohmic) conditions. 28% higher scaled energy confinement time B t τ E , t h / A is found in favour of the Tritium plasma. This can be cast in the form of the dimensionless energy confinement scaling law as Ω i τ E , t h ∼ A 0.48 ± 0.16 . This significant isotope mass scaling is consequently seen in the scaled one-fluid heat diffusion coefficient A χ e f f / B t which is around 50% lower in the Tritium plasma throughout the whole plasma radius. The isotope mass dependence in the particle transport channel is negligible, supported also by the perturbative particle transport analysis with gas puff modulation. The comparison of the edge particle fuelling or ionisation profiles from the EDGE2D-EIRENE simulations show that the absolute density differences that are necessary for the dimensionless match in the confined plasma dominate over any isotope mass dependencies of particle fuelling and ionization profiles at the plasma edge. Local GENE simulation results indicate a mild anti-gyroBohm effect at ρ t o r = 0.6 and thereby a small isotope mass dependence in favour of Tritium on heat transport and a negligible effect on particle transport. A significant fraction of the isotope scaling and reduced heat transport observed in the Tritium plasma is not captured in the GENE and ASTRA-TGLF-SAT2 simulations by simply changing the isotope mass for the same input profiles. [ABSTRACT FROM AUTHOR]

For more than a decade, an unprecedented predict-first activity has been carried in order to predict the fusion power and provide guidance to the second Deuterium–Tritium (D–T) campaign performed at JET in 2021 (DTE2). Such an activity has provided a framework for a broad model validation and development towards the D–T operation. It is shown that it is necessary to go beyond projections using scaling laws in order to obtain detailed physics based predictions. Furthermore, mixing different modelling complexity and promoting an extended interplay between modelling and experiment are essential towards reliable predictions of D–T plasmas. The fusion power obtained in this predict-first activity is in broad agreement with the one finally measured in DTE2. Implications for the prediction of fusion power in future devices, such as ITER, are discussed. [ABSTRACT FROM AUTHOR]

A toroidal Alfvén eigenmode (TAE) has been observed to be driven by alpha particles in a JET deuterium-tritium internal transport barrier plasma. The observation occurred 50 ms after the removal of neutral beam heating (NBI). The mode is observed on magnetics, soft-xray, interferometry and reflectometry measurements. We present detailed stability calculations using a similar tool set validated during deuterium only discharges. These calculations strongly support the conclusion that the observed mode is a TAE, and that this mode was destabilized by alpha particles. Non-ideal effects from the bulk plasma are interpreted as responsible for suppressing the majority of TAEs which were also driven by alpha particles, but the modes that match the observations are predicted to be particularly weak for these non-ideal effects. This mode located far from the core on the outboard midplane is found to be driven by both trapped and passing particles despite alpha particles originating in the core. [ABSTRACT FROM AUTHOR]

In JET deuterium-tritium (D-T) plasmas, the fusion power is produced through thermonuclear reactions and reactions between thermal ions and fast particles generated by neutral beam injection (NBI) heating or accelerated by electromagnetic wave heating in the ion cyclotron range of frequencies (ICRFs). To complement the experiments with 50/50 D/T mixtures maximizing thermonuclear reactivity, a scenario with dominant non-thermal reactivity has been developed and successfully demonstrated during the second JET deuterium-tritium campaign DTE2, as it was predicted to generate the highest fusion power in JET with a Be/W wall. It was performed in a 15/85 D/T mixture with pure D-NBI heating combined with ICRF heating at the fundamental deuterium resonance. In steady plasma conditions, a record 59 MJ of fusion energy has been achieved in a single pulse, of which 50.5 MJ were produced in a 5 s time window (P fus = 10.1 MW) with average Q = 0.33, confirming predictive modelling in preparation of the experiment. The highest fusion power in these experiments, P fus = 12.5 MW with average Q = 0.38, was achieved over a shorter 2 s time window, with the period of sustainment limited by high-Z impurity accumulation. This scenario provides unique data for the validation of physics-based models used to predict D-T fusion power. [ABSTRACT FROM AUTHOR]

The problem of inferring the internal inductance ℓ i from external magnetic measurements in tokamaks is considered. It was practically resolved in JET (Barana O et al 2002 Plasma Phys. Control. Fusion 44 2271), but several questions remain to be addressed before extrapolation of that method on other tokamaks. These naturally arise because of the integral nature of ℓ i determined by unknown distribution of the magnetic field inside the plasma. Without universal solutions, the quality of approximations should be examined along with potential experimental enforcements of the existing procedures. This is the main goal of the study that is fully analytical. Several relevant quantities are calculated for the large-aspect-ratio plasma with elliptical shifted magnetic surfaces. Usually the plasma elongation K is mentioned as a key factor in the task. Here it is explicitly shown that not only K , but its radial derivative too plays a role. The latter parameter did not appear in the JET scaling and in the accompanying discussions. It is also shown that the K -dependent term representing there the first iteration actually brings a systematic error. Therefore, several improvements to that scaling can be proposed. The study is based on the use of virial relations for the toroidal axisymmetric configurations. [ABSTRACT FROM AUTHOR]

Recently, stationary plasma with a world-record pulse length of 1056 s was achieved on the Experimental Advanced Superconducting Tokamak (EAST). In this work, the core magnetohydrodynamics (MHD) events as well as mode coupling processes have been investigated in EAST long pulse operation with electron heating dominant and pure radio frequency wave heating low collisionality, by using several diagnostics and the nonlinear numerical code M3D together. A saturated m / n = 1/1 kink mode was observed in the core region, where a stable internal transport barrier was found in the electron temperature channel. The frequencies and 2D structures of these modes were studied by a combination of soft X-ray (SXR) imaging and electron cyclotron emission diagnostics. The frequency of the m / n = 1/1 mode exhibited a feature of chirping down with time, and this chirping rate corresponds to the rate of electron diamagnetic drift frequency change. A twisted pattern ('Taichi structure in shape') was reconstructed by SXR tomography of the m / n = 1/1 mode. The perturbations of electron temperature and density caused by m / n = 1/1 are different in size, the latter one being much smaller. The destabilization of m / n = 1/1 was due to the strong central heating combined with electron cyclotron resonance heating and lower hybrid current drive. In the presence of the m / n = 1/1 mode, a negative current was generated on the magnetic axis, which anomalously broadened the core current profile. A m / n = 3/2 tearing mode triggered by m / n = 1/1 mode was also observed. The m / n = 3/2 mode has a smaller frequency than the m / n = 1/1 mode, and carries an m / n = 3/2 island with detectable size. A novel 3D MHD model that separately evolves the plasma density and temperature is applied to the m / n = 3/2 triggered by m / n = 1/1. It is found that a toroidal current density at the q = 1.5 surface, caused by non-axisymmetric density perturbation during the m / n = 1/1 nonlinear growth phase, was generated and hence the destabilization of the m / n = 3/2 tearing mode. Both the modeled electron temperature and density perturbations agree well with experimental observations. In the last section of this paper, the interaction between the m / n = 1/1 mode and fast electrons as well as the active control of this mode are also presented. [ABSTRACT FROM AUTHOR]

In this work, a novel practical strategy to emulate a reactor scenario on present tokamak experiments is presented. A recipe how to scale several relevant parameters from a hypothetical reactor scenario down to present devices is discussed. Equivalence between the energy flux channels is detailed, and the practical actuation scheme is presented. The application of the proposed protocol on the ASDEX Upgrade tokamak is shown foremost using the virtual flight simulator Fenix, with practical experiments planned for future campaigns. [ABSTRACT FROM AUTHOR]

Plasma β dependence of ion temperature gradient (ITG) driven turbulence is investigated using gyrokinetic simulations, where β is the normalized pressure. In our β scan, self-consistent magnetohydrodynamic (MHD) equilibrium state is numerically calculated for each value of β. It is found that the influence of the Shafranov shift cancels out the electromagnetic stabilizing effect on the ITG mode, and the growth rate of the ITG mode is accordingly unchanged as β increases. As a result, the turbulent energy transport does not decrease with β as suggested by the s − α model (Ishizawa et al 2019 Phys. Rev. Lett. 123 025003). A significant difference from the s − α model is the increase of the energy transport with β. It is also found that the critical onset β value for the kinetic ballooning mode is significantly increased by the influence of the Shafranov shift. The cancellation of the electromagnetic stabilization by the Shafranov shift is explained by the decrease of magnetic drift frequency in the dispersion relation of electromagnetic ITG modes obtained by using a fluid approximation. [ABSTRACT FROM AUTHOR]

NBI-assisted plasma heating with one or two injectors of fast neutral atoms was studied at the Globus-M2 spherical tokamak at the toroidal magnetic fields of 0.8–0.9 T and plasma currents of 0.35–0.4 MA. Measurements of the spatial temperature and electron density distributions, performed using the Thomson scattering diagnostics, showed a twofold increase in heating of plasma electrons during the injection of neutral particles with energies of up to 45 keV at the beam power of 0.75 MW, as compared to the ohmic heating regime. Switching on the second additional beam with the particle energy of up to 30 keV and power of up to 0.5 MW resulted in obtaining the hot ion mode in the range of mean plasma densities of (1.6–10) × 1019 m−3. According to the data of active spectroscopy and neutral particle analyzer diagnostics, in the hot zone, the ion temperature reached 4 keV at the plasma density of 8 × 1019 m−3, which is more than 2.5 times higher than the electron temperature. [ABSTRACT FROM AUTHOR]

The description of the zero-dimensional engineering-physical code GLOBSYS (Globus spherical tokamak system code), designed for parametric analysis of the next step of the program Globus-M, Globus-M2, is given. Within the framework of the zero-dimensional approximation, the definitions of the main scaling parameters of the plasma (poloidal beta, the fraction of bootstrap current, the energy lifetime of the plasma), as well as the specifics of calculating the inductance and resistance of the plasma in spherical tokamaks, are refined. The results of calculations of the plasma parameters by the code were compared with the experimental data of one of the Globus-M2 discharges (no. 38800) with neutral beam heating and showed good agreement. It is proposed to perform a comparison of calculations based on the code with the achieved and predicted parameters of the spherical tokamaks NSTX, NSTX-U, MAST, MAST-U, and ST40 in a separate paper. The goals of the next step (Globus-3) are formulated, the main ones of which are long pulse, high toroidal field, and powerful heating, which allow us to consider Globus-3 as a hydrogen prototype of a neutron source. The infrastructural restrictions on the Globus-3 parameters are given, which require further analysis of various versions of the electromagnetic system. Using the example of Globus-M2 discharge no. 38800, the effect of restrictions on the flow balance and heating of the elements of the electromagnetic system is shown. [ABSTRACT FROM AUTHOR]

The engineering part of the GLOBSYS code is presented, and the parameters of the Globus-3 facility, which is a development of the Globus program, are analyzed. The facility is primarily designed to provide a long pulse, a large toroidal magnetic field and strong heating. The concepts of searching for Globus-3 parameters under physical and engineering limitations are described. Obviously that reliable confinement and a large part of noninductive current are necessary to ensure existence of a plasma for a long time. Engineering constraints are involved in the choice of parameters in a more complex way: in some cases, it is overheating of the coils, in other cases, it is the total power supply, or the limit on the flux provided by the ohmic solenoid, or the strength of the constructions. The parameters of the Globus-3 spherical tokamak were preliminarily selected for the cases of a "warm" copper EMS (Electromagnetic system) and the EMS precooled to liquid nitrogen temperature. The exceeding of the duration of the plasma current plateau Δtplateau over the characteristic settling time of the plasma profiles τL/R was chosen as the key condition. At values of the toroidal magnetic field Bt0 = 3 T, the condition Δtplateau > τL/R cannot be attained even for precooled EMS. At Bt0 = 2 T, only options with precooled EMS can be considered acceptable, but the facility dimensions are fairly large. For the field Bt0 = 1.5 T, the options with "warm" EMS correspond to the duration of the plasma current plateau ~3 s (Δtplateau/τL/R ~ 1–1.5). In the case of precooled EMS, the duration of the plateau can increase to 12–13 s (Δtplateau/τL/R ~ 5). In the latter case, as a basis for further development of the Globus-3 facility, options with the following geometric dimensions are reasonable: R0 ~ 0.6–0.7 m, a ~ 0.35–0.4 m, А ≤ 1.7–1.8, k95 ~ 1.7–1.8. The minimum allowable value of the plasma current under the condition of effective absorption of the input power of neutral injection has been calculated. In the Globus-3 facility, Ip ≈ 0.8 MA was chosen as the base value. [ABSTRACT FROM AUTHOR]

Studies of energetic particle transport due to energetic-particle-driven Alfvénic instability have progressed using neutron and energetic particle diagnostics in Large Helical Device deuterium plasmas. Alfvénic instability excited by injecting an intensive neutral beam was observed by a magnetic probe and a far-infrared laser interferometer. The interferometer showed Alfvénic instability composed of three modes that existed from the core to the edge of the plasma. A comparison between the observed frequency and shear Alfvén spectra suggested that the mode activity was most likely classified as an Alfvénic avalanche. A neutron fluctuation detector and a fast ion loss detector indicated that Alfvénic instability induced transport and loss of co-going transit energetic ions. The dependence of the drop rate of the neutron signal on the Alfvénic instability amplitude showed that significant transport occurred. Significant transport might be induced by the large amplitude and radially extended multiple modes, as well as a large deviation of the energetic ion orbit from the flux surface. [ABSTRACT FROM AUTHOR]

Ion orbit loss is considered important for generating the radially inward electric field E r in a tokamak edge plasma. In particular, this effect is emphasized in diverted tokamaks with a magnetic X point. In neoclassical equilibria, Coulomb collisions can scatter ions onto loss orbits and generate a radially outward current, which in steady state is balanced by the radially inward current from viscosity. To quantitatively measure this loss-orbit current in an edge pedestal, an ion-orbit-flux diagnostic has been implemented in the axisymmetric version of the gyrokinetic particle-in-cell code XGC. As the first application of this diagnostic, a neoclassical DIII-D H-mode plasma is studied using gyrokinetic ions and adiabatic electrons. The validity of the diagnostic is demonstrated by studying the collisional relaxation of E r in the core. After this demonstration, the loss-orbit current is numerically measured in the edge pedestal in quasisteady state. In this plasma, it is found that the radial electric force on ions from E r approximately balances the ion radial pressure gradient in the edge pedestal, with the radial force from the plasma flow term being a minor component. The effect of orbit loss on E r is found to be only mild. [ABSTRACT FROM AUTHOR]

The properties of L-mode confinement have been investigated with a set of dedicated experiments in ASDEX Upgrade and with a related modelling activity with the transport code ASTRA and the quasi-linear turbulent transport model TGLFâ€"SAT2, with boundary conditions at the separatrix. The values at the boundary have been set by the two-point model for the electron temperature, with the ion temperature proportional to the electron temperature by a constant factor, and the electron density set by a constant fraction of the volume averaged density. The influx of neutrals has been set through a feedback procedure which ensures that in the simulation the same particle content as in the experiment is obtained. The sensitivity of the results under considerable variations in the choice of the boundary conditions has been investigated and found to be limited. The predictions of this full-radius modelling set-up have been compared to experimental results covering a scan in electron cyclotron resonance heating (ECRH) power in both hydrogen and deuterium plasmas, a plasma current scan with fixed magnetic field, under both ECRH and neutral beam injection heating, an increase in plasma density with constant ECRH power in hydrogen plasmas, as well as variations of the fraction of electron and ion heating at approximately constant total heating power, as well as a change of main ion from deuterium to hydrogen. The ASTRA-TGLF predictions have been found to reproduce all of the experimentally explored dependences with relatively good accuracy, providing evidence, for the first time to our knowledge, that the main properties of L-mode confinement can be reproduced by conventional full-radius transport modelling with a quasi-linear turbulent transport model. Evidences of largest disagreement, although usually not exceeding the 20%, have been found at high electron heating power, where TGLF underpredicts the electron and particularly the ion thermal stored energies, and in the current dependence of confinement, which, in electron heated conditions, is predicted to be weaker than in the experiment. [ABSTRACT FROM AUTHOR]

Density peaking has been studied between an ICRH and NBI identity plasma in JET. The comparison shows that 8 MW of NBI heating/fueling increases the density peaking by a factor of two, being R / L n = 0.45 for the ICRH pulse and R / L n = 0.93 for the NBI one averaged radially over ρ tor = 0.4, 0.8. The dimensionless profiles of q, ρ *, Ď... *, β n and T i/ T e ≈ 1 were matched within 5% difference except in the central part of the plasma (ρ tor < 0.3). The difference in the curvature pinch (same q -profile) and thermo-pinch (T i = T e) between the ICRH and NBI discharges is virtually zero. Both the gyro-kinetic simulations and integrated modelling strongly support the experimental result where the NBI fuelling is the main contributor to the density peaking for this identity pair. It is to be noted here that the integrated modeling does not reproduce the measured electron density profiles, but approximately reproduces the difference in the density profiles between the ICRH and NBI discharge. Based on these modelling results and the analyses, the differences between the two pulses in impurities, fast ions (FIs), toroidal rotation and radiation do not cause any such changes in the background transport that would invalidate the experimental result where the NBI fuelling is the main contributor to the density peaking. This result of R / L n increasing by a factor of 2 per 8 MW of NBI power is valid for the ion temperature gradient dominated low power H-mode plasmas. However, some of the physics processes influencing particle transport, like rotation, turbulence and FI content scale with power, and therefore, the simple scaling on the role of the NBI fuelling in JET is not necessarily the same under higher power conditions or in larger devices. [ABSTRACT FROM AUTHOR]

The evidence of a large diffusive turbulent contribution to the radial impurity transport in Wendelstein 7-X (W7-X) plasmas has been experimentally inferred during the first campaigns and numerically confirmed by means of gyrokinetic simulations with the code stella. In general, the absence of strong impurity accumulation during the initial W7-X campaigns is attributed to this diffusive term. Given the large variety of possible stellarator configurations, in the present work the diffusive contribution is also calculated in other stellarator plasmas. In particular, a numerical cross-device comparison is presented, where the diffusion (D) and convection (V) coefficients of carbon and iron impurities produced by ion-temperature-gradient (ITG) turbulence are obtained. The simulations have been performed for the helias W7-X, the heliotron LHD, the heliac TJ-II and the quasi-axisymmetric stellarator NCSX at the radial position r / a = 0.75. The results show that, although the size of D and V can differ across the four devices, inward convection is found for all of them. For W7-X, TJ-II and NCSX the two coefficients are comparable and the turbulent peaking factor is surprisingly similar. In LHD, appreciably weaker diffusive and convective impurity transport and significantly larger turbulent peaking factor, in comparison with the other three stellarators, are predicted. All this suggests that ITG turbulence, although not strongly, would lead to negative impurity density gradients in stellarators. Then, considering mixed ITG/trapped electron mode (TEM) turbulence for the specific case of W7-X, it has been quantitatively assessed to what degree pellet fueled reduced turbulence scenarios feature reduced turbulent transport of impurities as well. The results for trace iron impurities show that, although their turbulent transport is not entirely suppressed, a significant reduction of the convection and a stronger decrease of the diffusion term are found. Although the diffusion is still above neoclassical levels, the neoclassical convection would gain under such conditions a greater specific weight on the dynamics of impurities in comparison with standard ECRH scenarios without pellet fueling. [ABSTRACT FROM AUTHOR]

Nuclear fusion is one of the most attractive alternatives to carbon-dependent energy sources 1 . Harnessing energy from nuclear fusion in a large reactor scale, however, still presents many scientific challenges despite the many years of research and steady advances in magnetic confinement approaches. State-of-the-art magnetic fusion devices cannot yet achieve a sustainable fusion performance, which requires a high temperature above 100 million kelvin and sufficient control of instabilities to ensure steady-state operation on the order of tens of seconds 2,3 . Here we report experiments at the Korea Superconducting Tokamak Advanced Research 4 device producing a plasma fusion regime that satisfies most of the above requirements: thanks to abundant fast ions stabilizing the core plasma turbulence, we generate plasmas at a temperature of 100 million kelvin lasting up to 20 seconds without plasma edge instabilities or impurity accumulation. A low plasma density combined with a moderate input power for operation is key to establishing this regime by preserving a high fraction of fast ions. This regime is rarely subject to disruption and can be sustained reliably even without a sophisticated control, and thus represents a promising path towards commercial fusion reactors., (© 2022. The Author(s), under exclusive licence to Springer Nature Limited.)