Table of contents

Of all electrons, runaway electrons have long been recognized in the fusion community as a distinctive population. They now attract special attention as a part of ITER mission considerations. This review covers basic physics ingredients of the runaway phenomenon and the ongoing efforts (experimental and theoretical) aimed at runaway electron (RE) taming in the next generation tokamaks. We emphasize the prevailing physics themes of the last 20 years: the hot-tail mechanism of runaway production, RE interaction with impurity ions, the role of synchrotron radiation in runaway kinetics, RE transport in presence of magnetic fluctuations, micro-instabilities driven by REs in magnetized plasmas, and vertical stability of the plasma with REs. The review also discusses implications of the runaway phenomenon for ITER and the current strategy of RE mitigation.

The neutral beam heating system for the future international fusion experiment ITER will be based on radiofrequency driven ion sources delivering a large (≈1  ×  2 m) and homogeneous negative hydrogen or deuterium ion beam of severals tens of amperes for up to 1 h. Such beams have never been produced before and a dedicated R&D process has been ongoing for more than two decades. An important intermediate step is the size scaling test facility ELISE (Extraction from a Large Ion Source Experiment) with its half-ITER size ion source. Recently, ELISE has fulfilled its first main aim, demonstrating hydrogen ITER-relevant accelerated negative ion current densities over 1000 s, at the required filling pressure of 0.3 Pa, with an electron–ion ratio below one and a beam homogeneity better than 90%. The measures identified as essential for achieving such pulses are the introduction of external permanent magnets and internal potential rods as well as a dedicated caesium conditioning technique.

Global gyrokinetic simulations with self-consistent coupling of neoclassical and turbulent dynamics show that turbulence can significantly affect plasma self-driven mean current generation in tokamaks. The current amplitude, profile and associated phase space structures can all be modified. Turbulence can significantly reduce the current generation in the collisionless regime, generate current profile corrugation near the rational magnetic surface and nonlocally drive current in the linearly stable region—all these are expected to have a radical impact on broad tokamak physics. Both electron parallel acceleration and residual stress from turbulence play crucial roles in turbulence-induced current generation.

Injection of solid pellets is a key element in several aspects of the operation of a magnetic confinement plasma reactor for fusion applications, including plasma fueling, control and diagnosis. This letter reports observations demonstrating that pellet ablation can begin outside the plasma boundary, by effect of supra-thermal ions, whose orbits, under appropriate conditions, can extend well into the vacuum region. The phenomenon was recorded during plasma discharges in the DIII-D tokamak, combining pulsed modulation of heating neutral beams, with high-frequency injection of sub-millimeter lithium pellets, and it is ascribed to the large fraction of trapped beam ions associated with counter current neutral beam injection. The effect was quantitatively evaluated by means of Monte-Carlo simulations of supra-thermal ion orbits, finding that the heat-deposition was of the order of 50–100 W mm⁻², in the region traversed by the lithium pellets before reaching the plasma boundary, which is consistent with the severe pellet deterioration observed for low velocity pellets.

In high-current tokamak devices such as ITER, a runaway avalanche can cause a large amplification of a seed electron population. We show that disruption mitigation by impurity injection may significantly increase the runaway avalanche growth rate in such devices. This effect originates from the increased number of target electrons available for the avalanche process in weakly ionized plasmas, which is only partially compensated by the increased friction force on fast electrons. We derive an expression for the avalanche growth rate in partially ionized plasmas and investigate the effects of impurity injection on the avalanche multiplication factor and on the final runaway current for ITER-like parameters. For impurity densities relevant for disruption mitigation, the maximum amplification of a runaway seed can be increased by tens of orders of magnitude compared to previous predictions. This motivates careful studies to determine the required densities and impurity species to obtain tolerable current quench parameters, as well as more detailed modeling of the runaway dynamics including transport effects.

An internal transport barrier (ITB) and double tearing modes (DTM) have been observed during the off-axis sawteeth in EAST. The ITB of electron temperature T_e is modulated by the sawteeth oscillation, and the formation of an ITB can be divided into three stages: (i) the transport produced by a sawteeth final crash is suppressed at the first stage with a steep gradient of T_e; (ii) the micro-instability is developed at the second stage for further increasing the gradient of T_e; (iii) the ITB is formed eventually after the transition from beta-induced Alfvén eigenmodes (BAEs) to reversed shear Alfvén eigenmodes (RSAEs), where the BAEs-RSAEs pair enables the tracking of directly from the experiment. The off-axis sawteeth final crash is triggered by DTM: (i) the DTM can be excited by the redistribution of the profile of thermal particles, where the downward transport of energetic ions is detected indirectly by the soft x-ray arrays for the first time; (ii) the DTM can be excited by the transformation from the kink instability.

The characteristics of a negative hydrogen ion (H-) source and its neutralization efficiency determine the performance of a negative ion based neutral beam injector (NNBI). Therefore, for the safe operation of an NNBI system, it is necessary to monitor the performance of the ion source and its beam through a systematic characterization process. A judicious selection of different diagnostics based on electrical, optical and thermal types and including calorimetric techniques are required. In this regard, a number of diagnostics are being developed under the NNBI R&D program in the Indian Test Facility (INTF). These diagnostics are versatile in nature in terms of their working principles and independent prototype experimental efforts have been carried out to establish them and prepare them for operational use. Electrical probes (EP), optical emission spectroscopy and cavity ring down spectroscopy (CRDS) are mainly envisaged for ion source plasma characterization. Additionally, standard electrical measurements in RF and DC power supply circuits are already in regular use in the operational experimental setups, ROBIN and HELEN-I, for monitoring the power supplies. Doppler shift spectroscopy (DSS) and optical emission tomography (TOMO) are developed for beam characterization in terms of divergence, stripping and beam profile. Some of these are characterized on separate prototype experiments and are already integrated and have been tested in the available operational plasma experimental setups: ROBIN and HELEN-I. The DSS system, with multiple lines of sight (LOS) (blue-shifted and red-shifted), is integrated in the ROBIN setup and CRDS is arranged in HELEN-I. The TOMO technique is used to find the beam power density profile from the hydrogen beam emitted Balmer line intensity. The optical brightness profile of a neutral beam due to beam emission radiation is proportional to the beam power density. In this regard, a tomography code based on maximum entropy is developed to reconstruct the 2D optical emissivity profile of the INTF beam by inverting the LOS integral of the brightness of the beam. The code has been validated with the simulated INTF beam power density profile, in terms of the mathematical functions representing it. In the present manuscript, the performance evaluations of these diagnostics are presented.

Experimental observations and theoretical interpretations are presented in detail for the high frequency axisymmetric mode driven by nonlinear mode coupling. In the neutral beam injection heating plasma, toroidal Alfvén eigenmodes (TAEs) are found to nonlinearly interact with the tearing mode and result in the generation of other Alfvénic sidebands (Chen et al 2014 Europhys. Lett. 107 25001). Then two TAEs couple together and contribute to the excitation of an axisymmetric mode with frequency in the ellipticity-induced Alfvén eigenmode (EAE) frequency region. Direct evidence suggests the axisymmetric modes play an energy transfer channel role during the nonlinear procedure and cause the growth of magnetohydrodynamic instability with finite toroidal mode number via beating with tearing mode. Finally, the nonlinear process leading to high frequency n_E  =  0 mode generation is demonstrated using nonlinear gyrokinetic theory. Here n_E is toroidal mode number and the subscript E represents the mode frequencies lying in the EAE frequency range.

In plasmas heated with deuterium beams a deficit of the expected fusion neutron rate is an indicator of the deterioration of the fast-ion confinement, caused, for instance, by magnetohydrodynamic instabilities. The capability of predicting this deficit during the discharge relies on the availability of real-time estimates of the neutron rate from NBI codes which must be fast and accurate at the same time. Therefore, the recently developed real-time RABBIT code for neutral beam injection (NBI) simulations has been extended to output the distribution function and calculate the neutron emission. After the description of this newly installed diagnostics in RABBIT, benchmarks with NUBEAM, a massively used and validated Monte Carlo NBI solver, are discussed on ASDEX-Upgrade and JET cases. A first application for control-room intershot analysis on DIII-D is presented, and the results are compared on a large database with a slower NUBEAM analysis. Further application possibilities, e.g. for real-time control of Alfvén eigenmodes, are outlined.

Tungsten has established itself as the most suitable plasma-facing material for long-term operation in future magnetic-confinement fusion devices, but its properties make it a poor structural material and complicate the manufacturing of complex components. Recent advances in additive-manufacturing (AM) technology have begun to make the production of tungsten components with complex geometry more feasible. The design freedom afforded by AM could be leveraged to produce more resilient plasma-facing components (PFCs). A methodology to optimize the material distribution of composite PFCs was developed to reduce the maximum thermal stress caused by high heat fluxes. Its use was demonstrated for copper-infiltrated AM tungsten (W_AM/Cu) structures. Stress reductions of 50%–85% are predicted under nominal load conditions. Optimized designs also reduce stress over a wide range of off-nominal conditions. The resulting optimized structures are composed of a spatially heterogeneous distribution of W and Cu comprising a broad range of composite mixtures. A sample manufacturable component was modelled based on optimization results.

Highlights

• A methodology to optimize the material distribution of composite PFCs was developed to reduce the maximum thermal stress caused by high heat fluxes.

• Stress reductions of up to 85% compared to a monolithic W block may be feasible with topology optimization techniques.

• Optimized component designs are effective at reducing stress even over a wide range of off-nominal conditions.

• Manufacturable components can be designed based on optimization results.

Motivated by the superior confinement observed in the relaxed three dimensional (3D) states in the reversed field pinch, 3D plasma equilibria in coordinate systems based on space curves with a constant curvature as the axial coordinates are studied by using the method of metric perturbation. Constancy of the curvature allows the development of magnetohydrodynamic equilibrium with asymptotic good 2D flux surfaces near the coordinate axis. The perturbation parameter is the product of the torsion variation along the coordinate axis and the distance from it. The lowest order equilibrium with good 2D flux surfaces is symmetric with respect to translation along the space curve. It embodies the approximate toroidal-helical symmetry and is determined by a generalized Grad–Shafranov equation which includes information of the constant curvature and the average torsion of the space curve. Based on this fundamental equilibrium, a formal scheme is developed that allows us to find the ideal MHD equilibrium taking into account the full metric variation of the torsion along the spatial axis. In this limit, the flux surfaces are shown to exist for the full plasma. Numerical examples are given for the lowest order equilibria. It is suggested that equilibria based on this type of coordinates can allow easier evaluation of plasma shapes and magnetic boundary conditions that are more compatible with the relaxed central core. This could be the needed requirement for strongly self-organized equilibria with a large good confinement region.

In the ADITYA Upgrade tokamak, glow discharge wall conditioning (GDC) is performed regularly during the high-temperature plasma operation cycle using hydrogen (H) and helium (He) gases. H GDC is carried out after long durations (few hours) of plasma operations on every plasma operation day in automatic mode to control oxygen (O) and carbon (C) containing impurities. This leads to high retention of H gas on graphite limiter plates and stainless steel (SS) vessel walls. Subsequently, the high outgassing of H requires a prolonged pumping time and high H recycling during plasma discharges affects the plasma performance in respect to H fueling control of the plasma. To overcome the above-mentioned issues with continuous H GDC for longer durations, a new approach involving pulsed glow discharge wall conditioning (P-GDC) has been introduced in the ADITYA-U tokamak to reduce the residual H and He concentration in SS vessel walls and graphite limiter plates. To facilitate the fast initiation of a discharge in the case of P-GDC, a source of free electrons from a hot filament has been introduced in the vessel. A fast feedback controlled pulsed-gas-fueling system has been developed to initiate a glow discharge in each gas-feed pulse at various operating pressures from 1  ×  10⁻⁴ Torr to 10⁻³ Torr in the presence of an applied DC voltage. The different P-GDC experiments have been carried out with H, He and argon as the working gases and the results are compared with traditional continuous GDC. The P-GDC experiments have been optimized to provide beneficial wall conditioning for plasma operations. In this paper, the design, development and operation of P-GDC has been described along with the preliminary studies of its effect on the measured impurity line radiation during a plasma discharge.

NSTX-Upgrade will operate with toroidal magnetic fields () up to 1 T, nearly twice the value used in the experiments on NSTX, and the available neutral beam injection (NBI) power will be doubled. The doubling of while retaining the 30 MHz RF source frequency has moved the heating regime from the high harmonic fast wave (HHFW) regime used in NSTX to the mid harmonic fast wave regime. By making use of the full wave code AORSA (assuming a Maxwellian plasma), this work explores different HHFW scenarios for two possible antenna frequencies (30 and 60 MHz) and with and without NBI. Both frequencies have large electron absorption for large wave toroidal number particularly without NBI. With the presence of NBI, the fast ions absorption can be dominant in some scenarios. Therefore, a competition between electron and fast ion absorption is clearly apparent partially explaining why in previous NSTX HHFW experiments, a less efficient electron heating was observed. Moreover at the toroidal field of 1 T, a direct thermal ion damping might be possible under the condition when the ion temperature is larger than electron temperature. In general, the electron and ion absorption are found very sensitive to the ratio of electron and ion temperature. The impact of the hydrogen species is also studied showing that, for hydrogen concentration below , the hydrogen absorption is not significant. However, a larger hydrogen concentration could open up new HHFW heating scenarios in NSTX-U. Launching at high toroidal wave number appears to be one way to significantly reduce the ion damping and in turn to obtain large electron damping in the core which can play an important role in the transport studies for NSTX-U. Finally, an higher magnetic field could also playing a role in increasing the electron temperature and consequently the electron absorption. Indeed a magnetic field scan is also shown and discussed.

The effects of sawtooth on fast ion transport have been studied in reproducible, 2 s long sawtoothing L-mode discharges during the 2016 experimental campaign on National Spherical Torus Experiment Upgrade (NSTX-U) (Menard et al 2012 Nucl. Fusion52 083015). Analysis of the discharges demonstrated that standard sawtooth models (full/partial reconnection models) in the TRANSP code were not capable to fully reproduce the fast ion redistribution induced by sawtooth crashes. Some global parameters such as neutron rate can be recovered while detailed features, e.g. distribution functions, estimated using the models were different from the experimental observation. The standard sawtooth models in TRANSP do not take into account the different effect of sawtooth crashes depending on fast ion energy and orbit type and that may cause the disagreement between experiments and simulations. In this work, the newly developed kick model has been applied to replace the standard sawtooth models for the fast ion transport. TRANSP simulation results using the kick model, taking into account the characteristics of fast ion such as energy and pitch angle, can reproduce experimental neutron rates within 10% difference. The qualitative comparison of the measurements and synthetic diagnostics of fast ion D-alpha (FIDA) and solid state neutral particle analyser (SSNPA) using the TRANSP simulation results with kick model shows good agreements.

The purpose of this paper is to analyse higher toroidal harmonics m  =  1 with 7  <  n  <  23 (secondary modes) in the quasi single helicity, reversed field pinch plasma in the RFX-mod device. The quasi single helicity is an improved confinement state characterized by a dominant mode m/n  =  1/7, rotating along the toroidal direction. The spectroscopic measurements in the boundary plasma show footprints of a localized plasma deformation called 'phase locking', which can be described as an interferential pattern of toroidal Fourier harmonics. In the locking region the magnetic field lines are deformed in large poloidal lobes hitting the plasma facing components, a mechanism similar to the 'homoclinic lobes' observed in the Tokamak divertor with RMP application. Correspondingly, the magnetic connection length to the wall presents a strong decrease ('hole') with increased plasma-wall interaction, which fortunately is not stationary, but jumps in the toroidal direction thanks to the effective action of the RFX-mod feedback system. The global plasma wall interaction is the superposition of a rotating helix and of the localized toroidal deformation due to the secondary modes, which mirrors the local emission and particle influx from the wall. The local edge perturbation impacts also on the global plasma performance: a threshold to obtain an electron transport barrier has been identified. An improvement in plasma performance is expected in the upgraded device RFX-mod2, where the magnetic boundary will be modified to decrease the edge field deformation by a factor about 2.

Data on erosion and melting of beryllium upper limiter tiles, so-called dump plates (DP), are presented for all three campaigns in the JET tokamak with the ITER-like wall. High-resolution images of the upper wall of JET show clear signs of flash melting on the ridge of the roof-shaped tiles. The melt layers move in the poloidal direction from the inboard to the outboard tile, ending on the last DP tile with an upward going waterfall-like melt structure. Melting was caused mainly by unmitigated plasma disruptions. During three ILW campaigns, around 15% of all 12376 plasma pulses were catalogued as disruptions. Thermocouple data from the upper dump plates tiles showed a reduction in energy delivered by disruptions with fewer extreme events in the third campaign, ILW-3, in comparison to ILW-1 and ILW-2. The total Be erosion assessed via precision weighing of tiles retrieved from JET during shutdowns indicated the increasing mass loss across campaigns of up to 0.6 g from a single tile. The mass of splashed melted Be on the upper walls was also estimated using the high-resolution images of wall components taken after each campaign. The results agree with the total material loss estimated by tile weighing (~130 g). Morphological and structural analysis performed on Be melt layers revealed a multilayer structure of re-solidified material composed mainly of Be and BeO with some heavy metal impurities Ni, Fe, W. IBA analysis performed across the affected tile ridge in both poloidal and toroidal direction revealed a low D concentration, in the range 1–4  ×  10¹⁷ D atoms cm⁻².

We present a new theoretical approach, based on the Hamiltonian formalism, to investigate the stability of islands in phase space, generated by trapping of energetic particles (EPs) in plasma waves in a tokamak. This approach is relevant to MHD modes driven by EPs (EP-MHD) such as toroidal Alfvén eigenmodes (TAEs), EP-driven geodesic acoustic modes (EGAMs) or fishbones. A generic problem of a single isolated EP-MHD mode is equivalent to and hence can be replaced by a 2D Hamiltonian dynamics in the vicinity of the phase space island. The conventional Langmuir wave/bump-on-tail problem is then used as a representative reduced model to describe the dynamics of the initial EP-MHD.

Solving the Fokker–Planck equation in the presence of pitch angle scattering, velocity space diffusion and drag and retaining plasma drifts in a model, we find a 'perturbed' equilibrium, associated with these phase space islands. Its stability is then explored by addressing the Vlasov/Fokker–Planck–Poisson system. The Lagrangian of this system provides the dispersion relation of the secondary modes and allows an estimate of the mode onset. The secondary instabilities have been confirmed to be possible but under certain conditions on the primary island width and in a certain range of mode numbers. The threshold island width, below which the mode stability is reached, is calculated. The secondary mode growth rate is found to be maximum when the associated resonant velocity approaches the boundary of the primary island. This, in turn, leads to a conclusion that the onset of the secondary mode can be prevented provided the primary wave number is the lowest available.

Collective ion cyclotron emission (ICE) at the ion cyclotron frequency and its harmonics is a potential passive diagnostic of the fast-ion distribution in fusion reactors. ICE is observed in most plasmas in the DIII-D tokamak and is most strongly excited by the fast ions from neutral beam injection. The conventional outboard-edge ICE is detected in H-mode plasmas. However, weaker centrally-localized ICE is measured in L-mode plasmas, including those with negative triangularity shapes. Similar ICE spectra are found with both ICE diagnostics systems, the dedicated magnetic probes and the instrumented antenna straps. Many differences in the behavior of this central ICE are generated by varying the deuterium beam injection angle into deuterium plasmas. The co-current 'near-perpendicular' beam excites the most central ICE from the co-current beams, with this emission detected from the fundamental to the fifth ICE harmonic. However, the counter-current 'near-tangential' beam destabilizes the highest amounts of centrally-localized ICE. This emission is spectrally broader than that driven by the co-current beams and is observed up to its seventh ICE harmonic. The central ICE excited by this co-current beam correlated strongly on the local electron density and related parameters (plasma current and neutron rate) and increased with deeper fast-ion loss boundaries towards the magnetic axis. This was also the case with second harmonic ICE driven by the counter-current beam but not with its stronger third harmonic emission. The central ICE harmonics destabilized by both beams are observed to have different temporal dynamics. The central ICE amplitude responds rapidly to transient MHD events; it dropped and recovered in less than a millisecond at each sawtooth event. ICE frequency splitting is triggered by both the co-current and counter-current 'near-tangential' beams. The data presented in this paper provide opportunities to test and validate models of excitation of ICE by energetic ions.

Toroidal computation of the plasma response to the n  =  2 (n is the toroidal mode number) resonant magnetic perturbation field, based on an H-mode plasma in DIII-D, is carried out for the purpose of investigating the role of the ideal versus resistive plasma response models while scanning the plasma safety factor (q) at the edge. Both response models, implemented in the MARS-F code (Liu et al 2000 Phys. Plasmas7 3681), show significant amplification of the kink-peeling response in certain q-windows. A longstanding issue addressed in this work is the sensitivity of the q-window versus smoothing of the X-point geometry of the plasma separatrix. For this purpose, scan of the safety factor in 2D space (q₉₅, q_a) is carried out, where the q-value at 95% of the equilibrium poloidal flux, q₉₅, is scanned by varying the total plasma current, whilst the edge safety factor q_a is varied by gradually smoothing the X-point geometry at fixed total plasma current. Transition to the edge-peeling amplification domain is well described by simple analytic curves relating q₉₅ and q_a in the (q₉₅, q_a) space. These analytic curves not only help to quantify the sensitivity of the q₉₅-window versus q_a variation, but also establish the q₉₅ window in the asymptotic limit of infinite q_a (i.e. in the presence of true X-points). The 'resonant' q₉₅ values (3.6, 4.05, 4.5) are found to roughly exhibit 1/n periodicity at infinite q_a. Both ideal and resistive plasma response models, as well as the computed field at different pick-up locations, yield the same analytic curves describing transition to strong edge-peeling response. Detailed analysis of the computed plasma response and comparison with experiments are also performed.

The W7-X scrape-off layer (SOL) with its characteristic magnetic island chain has been investigated using electric probes mounted on a reciprocating manipulator close to the outboard mid-plane. A survey of the W7-X configuration space shows that the presence and particular topology of magnetic islands significantly affects the SOL profiles of electron temperature, density, electric field and plasma flows. Particularly relevant for divertor operation, very wide SOL heat flux profiles have been observed in some magnetic configurations, which we link to the presence of magnetic islands. In these situations, the islands can feature a local minimum of the plasma potential accompanied by a direction reversal of driven dynamics measured by probe arrays.

This paper presents a summary of recent progress pertaining to the manufacturing and inspection of ITER gyrotrons and the operation system in QST. Major achievements are as follows. (i) The final design of the ITER gyrotron was accomplished and manufacturing of two ITER gyrotrons was completed. Then operation test in QST prior the shipment to ITER has started with ITER relevant high voltage power supply configuration. The 1st ITER gyrotron has achieved 1.05 MW operation for 300 s with 51% efficiency. Measured cooling channel waveform of 300 s pulse demonstrated thermally stable condition representing sufficient cooling performance for 1 MW CW operation; (ii) 5 kHz modulation operation was demonstrated up to 200 s at with  >0.8 MW at flat top of pulses; (iii) 300 s/1 MW operation was repeated for 20 shots with successful 19 shots which demonstrating  >95% of reliability requirement. These results lead to success of ITER EC H&CD system construction and commissioning toward first plasma.

We present simulations of ignition and burn based on the Highfoot and high-density carbon indirect drive designs of the National Ignition Facility for three regimes of alpha-heating—self-heating, robust ignition and propagating burn—exploring hotspot power balance, perturbations and hydrodynamic scaling. A Monte-Carlo particle-in-cell charged particle transport package for the radiation-magnetohydrodynamics code Chimera was developed for this purpose, using a linked-list type data structure.

The hotspot power balance between alpha-heating, electron thermal conduction and radiation was investigated in 1D for the three burn regimes. Stronger alpha-heating levels alter the hydrodynamics: sharper temperature and density gradients at hotspot edge; and increased hotspot pressures which further compress the shell, increase hotspot size and induce faster re-expansion. The impact of perturbations on this power balance is explored in 3D using a single Rayleigh–Taylor spike. Heat flow into the perturbation from thermal conduction and alpha-heating increases by factors of , due to sharper temperature gradients and increased proximity of the cold, dense material to the main fusion regions respectively. The radiative contribution remains largely unaffected in magnitude.

Hydrodynamic scaling with capsule size and laser energy of different perturbation scenarios (a short-wavelength multi-mode and a long-wavelength radiation asymmetry) is explored in 3D, demonstrating the differing hydrodynamic evolution of the three alpha-heating regimes. The multi-mode yield increases faster with scale factor due to more synchronous compression producing higher temperatures and densities, and therefore stronger bootstrapping of alpha-heating. The perturbed implosions exhibit differences in hydrodynamic evolution due to alpha-heating in addition to the 1D effects, including: reduced perturbation growth due to ablation from both fire-polishing and stronger thermal conduction; and faster re-expansion into regions of weak confinement, which can result in loss of confinement.

Low-frequency fishbone instability driven by passing fast ions via wave-particle resonance is studied within the framework of energetic particle mode (EPM), where and are toroidal transit frequency and poloidal transit frequency of passing fast ions respectively. The effects of finite orbit width of energetic particles are responsible for the low-frequency fishbone of EPM type. It is found that magnetic shear at the q  =  1 radius plays an important role in the instability whereas the effect of background plasma beta is weak. In particular, for the case of co-injection of beam ions, there exists a critical magnetic shear below which the beam ion beta threshold for EPM excitation is very small. For moderate or higher magnetic shear, the beam ion beta threshold is much higher, on order of a few percent. These results are consistent with the observation of the low-frequency fishbone in the HL-2A tokamak.

The 'Super H-Mode' regime is predicted to enable pedestal height and fusion performance substantially higher than standard H-Mode operation. This regime exists due to a bifurcation of the pedestal pressure, as a function of density, that is predicted by the EPED model to occur in strongly shaped plasmas above a critical pedestal density. Experiments on Alcator C-Mod and DIII-D have achieved access to the Super H-Mode (and Near Super H) regime, and obtained very high pedestal pressure, including the highest achieved on a tokamak (p _ped ~ 80 kPa) in C-Mod experiments operating near the ITER magnetic field. DIII-D Super H experiments have demonstrated strong performance, including the highest stored energy in the present configuration of DIII-D (W ~ 2.2–3.2 MJ), while utilizing only about half of the available heating power (P_heat ~ 7–12 MW). These DIII-D experiments have obtained the highest value of peak fusion gain, Q_DT,equiv ~ 0.5, achieved on a medium scale (R  <  2 m) tokamak. Sustained high performance operation (β_N ~ 2.9, H₉₈ ~ 1.6) has been achieved utilizing n  =  3 magnetic perturbations for density and impurity control. Pedestal and global confinement has been maintained in the presence of deuterium and nitrogen gas puffing, which enables a more radiative divertor condition. A pair of simple performance metrics is developed to assess and compare regimes. Super H-Mode access is predicted for ITER and expected, based on both theoretical prediction and observed normalized performance, to allow ITER to achieve its goals (Q  =  10) at I_p  <  15 MA, and to potentially enable more compact, cost effective pilot plant and reactor designs.

Reduced activation ferritic/martensitic steels for in-vessel components of a fusion reactor have shown a decrease in plasticity and radiation hardening at low irradiation temperatures. The formation of dislocation loops and embryos of α' phase is considered the main reason for these effects. In this work, Eurofer 97 steel was irradiated with 5.6 MeV Fe²⁺ ions up to 10²⁰ m⁻² at 250, 300 and 400 °C. Transmission electron microscopy study of ion irradiated samples revealed nucleation of dislocation loops. The pair-correlation analysis of atom probe tomography data detected an initial stage of solid solution decomposition. The hardening of ion-irradiated Eurofer 97 was calculated with the dispersed barrier hardening model, taking into account radiation-induced dislocation loops to compare it with the change of yield stress in neutron-irradiated Eurofer 97. According to the obtained results, it can be supposed that the formation of dislocation loops plays the main role in the low-temperature radiation hardening of Eurofer 97 at a dose level up to ~10 dpa.

Iron (Fe) impurity behaviour in ADITYA tokamak plasma has been studied using Fe spectral line emissions in the vacuum ultra violet (VUV) wavelength range. A VUV survey spectrometer has been utilized to record spectral lines at 28.41 nm from Fe¹⁴⁺ , and 33.54 nm and 36.08 nm from Fe¹⁵⁺ . It has been observed that the intensities of the Fe emissions decrease with an increase in plasma electron density. The observed emission and the intensity ratio of Fe¹⁴⁺ and Fe¹⁵⁺ ions from two discharges having relatively low and high plasma density are modelled using an impurity transport code. It is found that the observed data could be modelled using the same ratio of the convective velocity to the diffusion coefficient D profile, but with two different Fe concentrations. The ratio varies from the value of   −  0.22 m⁻¹ at the plasma normalized radius to a maximum value of   −  0.35 m⁻¹ at . The obtained diffusion coefficient value at the plasma core region is explained in terms of neo-classical transport, indicating that Fe impurity transport follows the same behaviour in the core plasma of the ADITYA tokamak.

Within the EUROfusion MST1 work package, a series of experiments has been conducted on AUG and TCV devices to disentangle the role of plasma fueling and plasma shape for the onset of small ELM regimes. On both devices, small ELM regimes with high confinement are achieved if and only if two conditions are fulfilled at the same time. Firstly, the plasma density at the separatrix must be large enough (), leading to a pressure profile flattening at the separatrix, which stabilizes type-I ELMs. Secondly, the magnetic configuration has to be close to a double null (DN), leading to a reduction of the magnetic shear in the extreme vicinity of the separatrix. As a consequence, its stabilizing effect on ballooning modes is weakened.

During the 2016 NSTX-U experimental campaign, locked modes in the plasma edge presented clear evidence of the presence of error fields. Extensive metrology and plasma response modeling with IPEC and M3D-C1 have been conducted to understand the various sources of error fields in NSTX-U as built in 2016, and to determine which of these sources have the greatest effect on the plasma. In particular, modeling finds that the error field from misalignment of the toroidal field (TF) coils may have a significant effect on the plasma. The response to the TF error field is shown to depend on the presence of a q  =  1 surface, in qualitative agreement with experimental observations. It is found that certain characteristics of the TF error field present new challenges for error field correction. Specifically, the error field spectrum differs significantly from that of coils on the low-field side (such as the NSTX-U error field correction coils), and does not resonate strongly with the dominant kink mode, thus potentially requiring a multi-mode correction. Furthermore, to mitigate heat fluxes using poloidal flux expansion, the pitch angle at the divertor plates must be small (∼). It is shown that uncorrected error fields may result in potentially significant local perturbation to the pitch angle. Estimates for coil alignment tolerances in NSTX-U are derived based on consideration of both heat flux and core resonant fields independently.

The effect of the divertor pumping on plasmas has been investigated in the Large Helical Device (LHD). A divertor pump which reaches high pumping speed (67  ±  5 m³ s⁻¹ in hydrogen) and high pumping capacity (58 000 Pa m³ in hydrogen) has been developed at LHD. The global particle confinement time (), which is expressed as τ_p/(1  −  R), where τ_p is the particle confinement time and R is the global recycling coefficient, is used for comparison of operation with and without divertor pumping. After the fueling was stopped, shorter values were obtained in low density plasmas with divertor pumping in comparison with the values without divertor pumping. In this paper, the acceleration of RF wall conditioning by the divertor pumping is also discussed. The effect of the divertor pumping on high density plasmas has been investigated. In high density plasmas obtained by pellet fueling, not only edge density but also core density is reduced by the divertor pumping due to shallower pellet penetration. The control of neutral pressure by the feedback operation of pellet injection and gas puffing likely will be a candidate for the edge density control.

The features of plasma energy transfer to material surfaces during plasma–surface interactions (PSIs) in the presence of a strong magnetic field are investigated within the recently developed quasi-stationary plasma accelerator, QSPA-M. This novel PSI test-bed facility can reproduce edge localized mode (ELM) impacts, both in terms of heat load and particle flux to the surface, and provide plasma transportation in an external magnetic field, which mimics the divertor conditions. Investigations of energy transfer to the material surface have been performed for varied plasma heat load and external magnetic field values. Calorimetry, optical emission spectroscopy and high-speed imaging were applied for PSI characterization. For perpendicular plasma incidence, it has been shown that the transient plasma layer is formed in front of the surface by the stopped head of the plasma stream even for rather small plasma heat loads, which do not result in surface melting. The plasma density in this near-surface layer is much higher than in the impacting stream. It leads to the arisen screening effect for energy transfer to the surface. For B  =  0, the thickness of the screening layer is less than 3 cm, but it increases to 15 cm when B  =  0.8 T. The shielding effect due to the formation of a dense plasma layer in front of the exposed surface should be favorable for material performance, being important for decreasing the overall erosion of plasma-facing components during a large number of repetitive ELMs.

The precise delivery of mass to burning plasmas is an area of growing interest in magnetic fusion (MF). The amount of mass that is necessary and sufficient can vary depending on such parameters as the type of atoms involved, the type of applications, plasma conditions, mass injector, and injection timing. Motivated by edge localized mode (ELM) control in H-mode plasmas, disruption mitigation and other applications in MF, we report the progress and new possibilities in mass delivery based on hollow pellets. Here, a hollow pellet refers to a spherical shell mass structure with a hollow core. Based on an empirical model of pellet ablation, coupled with BOUT++ simulations of the ELM triggering threshold, hollow pellets are found to be attractive in comparison with solid spheres for ELM control. By using hollow pellets, it is possible to tailor mass delivery to certain regions of edge plasmas while minimizing core contamination and reducing the total amount of mass needed. We also include the experimental progress in mass delivery experiments, in situ diagnostics and hollow pellet fabrication, and emphasize new experimental possibilities for ELM control based on hollow pellets. A related application is the disruption mitigation scheme using powder encapsulated inside hollow shells. Further experiments will also help to resolve known discrepancies between theoretical predictions and experiments in using mass injection for ELM control and leading to better predictive models for ELM stability and triggering.

He-induced W nanofuzz growth over the W divertor target is one of the main limiting factors affecting the current design and development of fusion reactors. In this paper, based on He reaction rate model in W, we simulate the growth and evolution of He nanobubbles during W nanofuzz formation under fusion-relevant He⁺ irradiation conditions. Our modeling unveils the existence of He nanobubble-enriched W surface layer (<10 nm), formed due to the He diffusion in W crystal into defect sites. At an elevated temperature, the growth of He bubbles in the W surface layer prevents He atoms diffusing into the deep layer (>10 nm). The formation of W nanofuzz at the surface is attributed to surface bursting of high-density He bubbles with their radius of ~4 nm, and an increase in the surface area of irradiated W. Our findings have been well confirmed by the experimental measurements.

Energetic ions are well confined in the core of the reversed field pinch and a substantial population develops rapidly during tangential neutral beam injection. A saturated fast ion density is observed within several milliseconds and coincides with the onset of energetic particle mode activity. First measurements of the fusion product profile determine the fast ion density and pressure profiles, the latter of which when combined with neutral particle analysis of the energy distribution. These are key measurements in explaining the limiting behavior of the fast ion profile in response to a combination of effects. Tearing mode activity affects the fast ion profile, and the fast ions noticeably influence the core-most tearing mode. The nonlinear situation settles into a marginally stable state where the EPM transports fast ions down a steep gradient to a region where their confinement is limited by stochastic orbit and charge exchange losses. In the state at marginal stability (achieved in one set of experimental conditions with full 1 MW neutral beam power), a critical fast ion beta gradient is measured that limits the core , or nearly four times the core thermal , but less than a strictly classical calculation predicts.

We present experimental observations suggesting that non-diffusive avalanche-like transport events are a prevalent and universal process in the electron heat transport of tokamak plasmas. They are observed in the low confinement mode and the weak internal transport barrier plasmas in the absence of magnetohydrodynamic instabilities. In addition, the electron temperature profile corrugation, which indicates the existence of the shear flow layers, is clearly demonstrated as well as their dynamical interaction with the avalanche-like events. The measured width of the profile corrugation is around , implying the mesoscale nature of the structure.

The first energetic particle experiments in negative triangularity tokamak plasmas have been carried out on DIII-D. Alfvén eigenmode (AE) activity and associated fast ion transport comparable to that in positive triangularity is observed during the current ramp portion of all plasmas with early beam heating indicating negative triangularity does not confer a special advantage with respect to AE induced transport. In these discharges, a range of mode activity driven by the sub-Alfvénic () 80 kV neutral beams is found including beta induced Alfvén acoustic eigenmodes, beta induced Alfvén eigenmodes, reversed shear Alfvén eigenmodes (RSAEs) and toroidicity induced Alfvén eigenmodes (TAEs). Mode intermittency and possibly chirping appears to be more common than comparable positive triangularity and/or oval discharges but overall, the unstable spectra and mode amplitudes observed on magnetics, CO2 interferometry, and electron cyclotron emission in a set of matched positive and negative triangularity cases at moderate beam power is similar. Large levels of Alfvén eigenmode induced fast ion transport are found in both positive and negative triangularity with up to central fast ion pressure deficits relative to classical predictions early during the current ramp phase for discharges with 3 MW injected 80 kV neutral beam power. The deficit in both cases is reduced toward zero as the current penetrates and eventually reaches classical levels during current flattop with . Fast ion transport in negative triangularity plasmas measured using a beam modulation technique show very similar levels to those measured in oval plasmas over the range of tested beam powers (–7 MW). This similarity is found despite quite different unstable spectra in the highest beam power case where a mixture of coherent and quasi-coherent modes in the TAE frequency range are observed in negative triangularity and a spectrum of more typical narrowband TAEs and RSAEs is observed for the oval plasmas.

The retention of radioactive tritium in the reactor wall is a safety issue and should be kept to a minimum. We investigated the influence of helium, argon, neon and nitrogen as plasma impurities on the deuterium retention in tungsten in the linear plasma devices PSI-2 and PISCES-A. Following mixed plasmas were produced: pure D, D  +  He, D  +  Ar, D  +  Ne, D  +  N and D  +  He  +  Ar, with impurity fractions between 3% and 10%. Experiments were performed at tungsten sample temperatures of 500 and 770 K. The incident ion energy was 70 eV, above the tungsten sputtering threshold for argon and nitrogen, but below it for deuterium and helium. For neon, in addition, it was varied between 20 and 70 eV, below and above the tungsten sputtering threshold, respectively. The admixture of helium reduced the deuterium retention by a factor of 3–100, with a stronger reduction at a higher sample temperature. In the D  +  He  +  Ar case the retention was similar as for pure D. Argon sputtered the near-surface helium nanobubble layer and thus overrode the effect of helium. The effect of neon is sensitive to the incident ion energy. Addition of nitrogen increased the deuterium retention by a factor of ~10 and ~100 for 500 and 770 K, respectively. In general, the effect of impurities on the deuterium retention appears to be sensitive to the properties of the affected near-surface layer of tungsten. Admixed species, i.e. helium, can form a layer with open porosity, which serves as an additional release channel for deuterium thus decreasing the retention. However, if the process is dominated by sputtering, as for argon, such a layer cannot be formed. The nitrogen enriched layer, in contrast, serves as a desorption barrier for deuterium increasing the retention.

Recent Experimental Advanced Superconducting Tokamak (EAST) experiments have successfully demonstrated a long-pulse steady-state scenario with improved plasma performance through integrated operation since the last IAEA FEC in 2016. A discharge with a duration over 100 s using pure radio frequency (RF) power heating and current drive has been obtained with the required characteristics for future long-pulse tokamak reactors such as good energy confinement quality (H_98y2 ~ 1.1) with electron internal transport barrier inside ρ  <  0.4, small ELMs (frequency ~100–200 Hz), and good control of impurity and heat exhaust with the tungsten divertor. The optimization of X-point, plasma shape, the outer gap and local gas puffing near the low hybrid wave (LHW) antenna were integrated with global parameters of B_T and line-averaged electron density for higher current drive efficiency of LHW and on-axis deposition of electron cyclotron heating in the long-pulse operation. More recently, a high β_P RF-only discharge (β_P ~ 1.9 and β_N ~ 1.5, /n_GW ~ 0.80, f _bs ~ 45% at q₉₅ ~ 6.8) was successfully maintained over 24 s with improved hardware capabilities, demonstrating performance levels needed for the China Fusion Engineering Test Reactor steady-state operation. A higher energy confinement is observed at higher β_P and with favorable toroidal field direction. Towards the next goal (⩾400 s long-pulse H-mode operations with ~50% bootstrap current fraction) on EAST, an integrated control of the current density profile, pressure profile and radiated divertor will be addressed in the near future.

Within this research we present a method to speed up the simulation of convective heat loads onto the plasma-facing components (PFCs) of the Wendelstein 7-X (W7-X) stellarator by a factor of approximately with the same statistical precision as compared to the previous standard. The geometric models developed for this are also designed to unravel the complex 3D PFCs onto a 2D picture-like input format which gives access to the full set of image analysis tools like for example wavelet analysis or the applicability of convolutional neural network (CNN) architectures.

The significant speedup of heat load calculation allows to simulate a massive data set of heat loads for approximately magnetic configurations in edge rotational transform-radial axis shift space. To first order, plasma dynamic effects like toroidal current development as well as beta effects mainly influence this space which motivates the simulation in this scope.

A criterion to evaluate the safety of a magnetic configuration with respect to the convective heat load onto the plasma facing components has been developed taking statistical fluctuations of the simulation into account. This criterion, applied to the introduced data set, provides a map relevant for discharge planning and machine safety.

The methods and concepts introduced herein could contribute to the safety evaluation of magnetic confinement devices in general and are not specific for W7-X.

Core ion cyclotron emission (ICE) signals have recently been observed in beam-heated plasmas on several tokamaks. In this manuscript we present experimental evidence in support of core ICE being generated by the magnetoacoustic cyclotron instability (MCI), itself driven by the velocity-space inversion of sub-Alfvénic beam-injected ions. The observed core ICE amplitude evolution in beam-heated plasmas is consistent with the MCI and is a result of competition between the beam ion fraction build-up (which increases the instability growth rate) and the slowing down of the dominant beam ion velocity component (which stabilizes the MCI). The oblique propagation angle is constrained to lie in the range 75–78°, where the lower bound is governed by the fast wave reabsorption losses on plasma electrons and bulk ions and the upper bound is governed by the experimentally observed ICE amplitude growth time. Although the oblique MCI is a local theory and does not provide information on the radial mode localization, the calculated instability growth rates are broadly consistent with the observed dynamics of the core ICE amplitude on the ASDEX Upgrade tokamak.

The EU DEMO reactor is expected to be among the first applications of fusion for electricity generation in the near future and the design of its magnet system is of central importance as driving power plant performance, budget and production efficiency. In this purpose activities were led by CEA in the framework of EUROfusion Consortium to contribute to the EU DEMO magnet system design. It encompassed design activities (dimensioning and development of associated modelling tools) with R&D (design and tests of mock-ups). The CEA design activity was mainly oriented towards Toroidal Field (TF) coils system to propose a conceptual option (pancake-wound, no radial plates) established with a semi-analytical CEA tool that considers the inter-dependent electromagnetic and mechanical behaviors. Then the proposed design is consolidated by detailed analyses: Thermo-hydraulics evaluation by coupling THEA, TRAPS and CAST3M softwares respectively for thermo-hydraulics, electromagnetic and thermal items. The outcomes obtained in normal and off-normal regimes are exposed and discussed in the paper; Mechanics evaluation with the most stressed zones identified and their criticity evaluated, in particular in the insulation zones. Design optimization analyses were conducted on jacket shape, together with investigations on the thermo-mechanic hotspot criterion. Further to the TF system, the central solenoid design was addressed and an optimization analysis will be presented and discussed. On another side, CEA also conducted R&D activities, mostly regarding the TF system with hydraulic tests at variable void fraction to explore its impact on helium friction and a full-scale TF conductor sample design and manufacture.

In future fusion reactors, tungsten is considered as the main candidate material for plasma-facing components. However, the intrinsic brittleness of tungsten is of great concern during operation. To overcome this drawback, tungsten fiber-reinforced tungsten composites (W_f/W) are being developed relying on extrinsic toughening principles. Tungsten (W) fibers with extremely high tensile strength and ductility are used to reinforce a tungsten matrix.

In this work, field assisted sintering technology (FAST) is used to produce W_f/W material. Mechanical characterizations including Charpy impact and 3-point bending tests are performed. Based on the 3-point bending test results, the W_f/W materials can facilitate a promising pseudo-ductile behavior even at room temperature, similar to fiber reinforced ceramic composites. Fracture energy density and fracture toughness together with the crack-resistance curves (R-curves) are measured. Compared to conventional pure tungsten, W_f/W shows significant improvement in fracture toughness.

Laser lights with relativistic intensities and pulse lengths exceeding the picosecond (ps) have been recently made available. Laser–plasma interactions with such a parameter regime belong to the mesoscale between kinetic and fluid regimes, and thus theories developed for sub-ps laser–plasma interactions are not straightforwardly applicable to those for the multi-ps regime. We here study the generation of high-energy electrons in ps relativistic laser–foil interactions by using the particle-in-cell (PIC) simulation. We show that the dynamics of the laser hole boring, which stops during over-ps laser irradiation, is a key to generate the high energy electrons. An energy distribution with a high-energy tail cannot be fitted by a single Maxwellian function, and is likely to be a nonthermal distribution through a stochastic interaction via the recirculation of electrons in the expanding foil plasma. The present study of superthermal electron generation can be a basis for ps laser applications such as fast ignition-based laser fusion, laser ion acceleration, and short-pulse x-ray generation.

A total power injection up to 0.3 GJ has been achieved in EAST long pulse H-mode operation of 101.2 s with an ITER-like water-cooled tungsten (W) mono-block divertor, which has steady-state power exhaust capability of 10 MWm⁻². The peak temperature of W target saturated at 12 s to the value T ~ 500 °C with a heat flux ~3.3 MW m⁻² being maintained during the discharge. By tailoring the 3D divertor plasma footprint through edge magnetic topology change, the heat load was broadly dispersed and thus peak heat flux and W sputtering were well controlled. Active feedback control of H-mode detachment with D₂ fuelling or divertor impurity seeding has been achieved successfully, with excellent compatibility with the core plasma performance. Active feedback control of radiative power utilizing neon seeding was achieved with f _rad  =  18%–41% in H-mode operation, exhibiting potential for heat flux reduction with divertor and edge radiation. This has been further demonstrated in DIII-D high β_P H-mode scenario within the joint DIII-D/EAST experiment using impurity seeding from the divertor volume. Steady-state particle control and impurity exhaust has been achieved for long pulse H-mode operation over 100 s with the W divertor by leveraging the effect of drifts and optimized divertor configuration, coupled with strong pumping and extensive wall conditioning. Approaches toward the reduction of divertor W sourcing, which is of crucial importance for a metal-wall tokamak, are also explored. These advances provide important experimental information on favourable core-edge integration for high power, long-pulse H-mode operation in EAST, ITER and CFETR.

For many years, machine learning tools have proved to be very powerful disruption predictors in tokamaks. On the other hand, the vast majority of the techniques deployed assume that the input data is independent and is sampled from exactly the same probability distribution for the training set, the test set and the final real time deployment. This hypothesis is certainly not verified in practice, since the experimental programmes evolve quite rapidly, resulting typically in ageing of the predictors and consequent suboptimal performance. This paper describes various adaptive training strategies that have been tested to maintain the performance of disruption predictors in non-stationary conditions. The proposed approaches have been implemented using new ensembles of classifiers, explicitly developed for the present application. The improvements in performance are unquestionable and, given the difficulties encountered so far in translating predictors from one device to another, the proposed adaptive methods from scratch can therefore be considered a useful option in the arsenal of alternatives envisaged for the next generation of devices, particularly at the very beginning of their operation.

Tungsten–vanadium (W–V) alloys with 5 and 10 wt.% V made using hot-pressing sintering have been subjected to helium plasma at 1050 K with ion energy of 100 eV for one hour in a linear plasma generator STEP. The W–V alloys consist of three typical phase regions, i.e. the W-enriched region, the W–V solid solute region and the V-enriched region. Slight surface erosion on the W–V solid solute region and severe surface erosion on the V-enriched region are observed on the exposed surface. Sputtering during the He plasma loading contributed to the surface erosion in the W–V solid solute region and the V-enriched region. Slight sputtering on W–V solid solute region is expected given that the lighter alloying element increases the sputtering yield of the W-based material. Moreover, fuzz nanostructure formed on the surface of the W–V alloys under the applied irradiation conditions. The W–V solid solute region exhibits severe fuzz growth in comparison with the W-enriched region when exposed to He plasma, especially in the low flux area. The intense attraction interaction between He and substitutional V atoms aggravates the fuzz growth on the W–V solid solute region. Furthermore, the aggravation effect of V on fuzz nanostructure evolution is mitigated in the high flux area due to the enhanced preferential V sputtering.

The first ITER toroidal field coil structures (TFCS) have been successfully completed by the Japan Domestic Agency with the solving of major technical challenges, and further fabrication is ongoing. There were five major challenges for the TFCS fabrication. This paper describes these challenges and the solutions established during fabrication and how to implement to the ITER TFCS.

The axial merging method is one of the candidates to provide a center-solenoid-free start-up of high-beta spherical tokamak (ST) plasma. Two initially formed STs merge through magnetic reconnection in the presence of the guide (toroidal) magnetic field, which is perpendicular to the reconnection (poloidal) magnetic field. During ST merging start-up, electrons are effectively accelerated near the reconnection point where the reconnection electric field is almost parallel to the magnetic field. In order to evaluate the effectivity of this acceleration process on electron heating, the temporal and spatial distributions of generated energetic electrons are observed by a soft x-ray fast imaging system equipped on the UTST device. The energetic electrons were generated not only in the vicinity of the reconnection point but transiently in the inboard-side downstream region until the static electric field by charge separation grew to cancel the reconnection electric field component parallel to the magnetic field line. Adequate control of the downstream condition could enhance the generation of energetic electrons and provide a more effective conversion from the released magnetic energy to electron energy.

Ion heating/transport and its fine structure formation process through magnetic reconnection have been investigated by high guide field tokamak merging experiments in TS-3 and TS-3U. In addition to the previously reported demonstration of high-temperature plasma startup without center solenoid, the detailed fine structure formation process of reconnection heating has been revealed using new 96CH/320CH ultra-high-resolution 2D ion Doppler tomography diagnostics. By identifying the double-axis field configuration with the X-point on the midplane using in situ magnetic probe diagnostics, the detailed measurement successfully revealed that the ion temperature profile forms two types of characteristic heating structure, both around the X-point and downstream. The former is affected by the Hall effect to form a tilted heating profile, while the latter is affected by the transport process which a forms a poloidal double-ring-like structure. The achieved ion heating mostly depends on the reconnecting component of the magnetic field, and the contribution of the guide field to decrease the heating efficiency tends to be saturated in the high guide field regime. Under the influence of better toroidal confinement with higher guide field, the downstream ion heating is transported vertically, mostly by parallel heat conduction, and finally forms a poloidal ring-like hollow distribution aligned with the closed flux surface at the end of merging.

For the realization of the ITER neutral beam (NB) system, a method for achieving the acceleration voltage of 1 MV for the vacuum insulated beam source (VIBS) has been developed. Additionally, the design basis for the 1 MV vacuum insulation has also been developed by integrating previous empirical scaling of the voltage holding capability for plane and coaxial electrodes with new scaling for the corner region where the locally-concentrated electric field is generated. The scaling clarified that a product of electric field and voltage obtained by electric field analysis in the beam source vessel was much higher than the allowable level. Therefore, in order to improve the voltage holding capability, electrostatic shields having intermediate potential have been proposed. Consequently, it was found that the VIBS needs to be surrounded by more than three intermediate shields instead of a single gap insulation of 1 MV. Reliability of the design basis of the developed intermediated shields has been experimentally verified by improving the 1.3 m single gap insulation in the voltage holding test. As a result, the sustainable voltage has been significantly improved from 0.7 MV with the single gap to 1 MV with shielding up to the power supply limit. This result ensures the realization of the 1 MV VIBS in the ITER NB system.

A detached divertor plasma is predicted to be necessary to mitigate the heat fluxes and ion energy, and so sputtering, at the divertor plate of a fusion power plant. It has proven difficult in current experiments, however, to prevent the detachment front, once formed, from running up to the main plasma resulting in deterioration of pedestal performance. The lithium vapor box divertor localizes a dense cloud of lithium vapor away from the main plasma, in order to induce detachment at a stable, distant location. The vapor localization is created by local evaporation and nearby condensation, a configuration inaccessible for non-condensing seed impurities. This paper provides simulations of lithium vapor flow using the SPARTA direct simulation Monte Carlo (DSMC) code, in which the lithium vapor dynamics are governed by Li–Li collisions that should dominate in regions with little plasma. Inertia is included in this code, permitting the formation of shocks. The plasma model implemented in SPARTA is based on UEDGE results for the location of ionization and recombination. We find that a simplified lithium vapor box configuration, without baffles, provides robust stabilization of the detachment front and acceptable lithium vapor flow to the main chamber. Lithium is evaporated from a capillary porous surface on the private-flux side of the divertor, distant from the main plasma, and is condensed on the other side walls of the divertor chamber. The condensed lithium flows back to the evaporator through 2 cm internal diameter pipes. The capillary pressure differential at the surface of the evaporator is capable of overcoming MHD back-force in the piping, with a great deal of margin if a sandwich flow channel insert is implemented. Very little lithium should be accumulated on the first wall of the main chamber if it is maintained at 600 °C, as expected for a power-producing fusion system.

A highly reproducible spontaneous non-inductive low recycling no-ELM regime has been achieved in the EAST superconducting tokamak with the water-cooled tungsten divertor and molybdenum first wall. The salient feature is that after the L-H transition, emission intensity, divertor neutral pressure and the separatrix electron density all decrease, and the amplitude of ELMs decreases simultaneously until they disappear completely. This scenario exhibits a distinct density threshold, which is dependent on plasma current and wall condition. ELITE/NIMROD simulations indicate that the density at the foot of pedestal and ion diamagnetic effect may play a key role in the achievement of no-ELM operation; the low pedestal foot density might have a strong ion diamagnetic stabilizing effect, which can stabilize the medium-n and high-n Peeling-Ballooning modes. In order to achieve this low pedestal foot density condition, a number of techniques have been employed in EAST, i.e. through extensive lithium wall conditioning and active control of plasma configuration and strike point position. The low recycling no-ELM regime can be extended into a small ELMs regime by increasing the pedestal foot density to facilitate impurity control for long pulse operation, which has been demonstrated in EAST with a pulse length exceeding 101 s. Such a small ELMs regime provide a possible approach toward long-pulse steady-state H-mode operation in future devices.

Nuclear inventory simulations have a vital role to play in the planning and execution of future fusion experiments and power plants. They are able to predict the transmutation (burn-up) response of material compositions under neutron irradiation, thus providing information about the build-up of impurities that could impact on material performance. The inventory evolution also quantifies the radiological response of a material by tracing the production (and decay) rates of radioactive nuclides. This information can be used to plan maintenance schedules at nuclear facilities, satisfy nuclear regulators during reactor planning, and quantify the waste disposal needs during reactor decommissioning.

However, the validity of inventory simulations must be verified to give confidence in the predictions. This paper describes a validation and verification (V&V) benchmark exercise that tests the quality of nuclear code predictions. Such benchmarks are an important part of the development and release of the FISPACT-II inventory code and its associated input nuclear data libraries. This paper describes the latest V&V based on the fusion decay-heat measurements performed at the Japanese FNS facility. Rigorous and detailed assessment techniques, focussed on the complex breakdown of decay-heat contributions from individual radionuclides, have been employed to interpret the simulated results, benchmark the data against the experimental measurements, and to compare results from different international nuclear data libraries. Example results are presented and discussed for nickel, iron, niobium, tungsten, chromium, and osmium, using FISPACT-II simulations performed with TENDL-2017, JEFF-3.3, ENDF/B-VIII.0, EAF2010, and IRDFF-1.05 nuclear cross section libraries.

Multiple turbulent plasma states in the edge transport barriers formation are studied on JT-60U. Following a slow L–H transition, which causes significant reduction in the ion thermal transport in the pedestal towards the neoclassical level with a weak negative E_r value, we found clear and fast changes in the particle transport in association with the change in E_r towards a strong negative value at the later H-phase. It has also been suggested that there is a clear difference in responses between ion and electron transports due to the change in the inhomogeneity in E_r. This observation suggests the presence of multiple types of turbulent fluctuations in the H-mode plasma state, which affects the ion energy and other channels of transport differently. The nonlinear dynamics that induce the E_r-bifurcation are also analyzed in terms of the radial current generation. These analyses contribute to understanding of the L–H transition condition.

Predictability of burning plasmas is a key issue for designing and building credible future fusion devices. In this context, an important effort of physics understanding and guidance is being carried out in parallel to JET experimental campaigns in H and D by performing analyses and modelling towards an improvement of the understanding of DT physics for the optimization of the JET-DT neutron yield and fusion born alpha particle physics. Extrapolations to JET-DT from recent experiments using the maximum power available have been performed including some of the most sophisticated codes and a broad selection of models. There is a general agreement that 11–15 MW of fusion power can be expected in DT for the hybrid and baseline scenarios. On the other hand, in high beta, torque and fast ion fraction conditions, isotope effects could be favourable leading to higher fusion yield. It is shown that alpha particles related physics, such as TAE destabilization or fusion power electron heating, could be studied in ITER relevant JET-DT plasmas.

Based on an approximate modeling of scrape-off-layer (SOL) dynamics and edge turbulence, it is shown that a new type of limit cycle oscillation (LCO) or one-step L–H transition can be triggered from the outer-well E  ×  B flow shear originating from the SOL region. From its close relevance to SOL properties, this model appears to provide a possible explanation of the L–H threshold power dependence on magnetic or divertor geometry.

We investigate the penetration threshold of resonant magnetic perturbation (RMP) by the external coils in the Large Helical Device (LHD) for various plasma aspect ratio configurations. The qualitative dependence on the collisionality is opposite to that in a high plasma aspect configuration; this is a quite unique property first found in the LHD. We also investigate the threshold dependence on the ion species, and find that the threshold in deuterium discharges is much smaller than that in hydrogen discharges. In all cases the thresholds are higher as the poloidal rotation becomes faster, which shows that poloidal rotation is the dominant driver to the RMP shielding. This is qualitatively consistent with the torque balance model between the electromagnetic and poloidal viscous torques. In a configuration of the LHD, the dependence of the threshold on the density is qualitatively similar to that in Ohmic tokamak plasmas, but the beta dependence is opposite to that of tokamaks. The difference arises from the cause of the viscous torque.

Polycrystalline tungsten (W) samples were simultaneously irradiated by 10.8 MeV W ions and exposed to 300 eV deuterium (D) ions at different temperatures ranging from 450 K to 1000 K. After simultaneous W ion irradiation and D ion exposure the samples were additionally exposed to low-energy D ions at 450 K in order to populate all the defects created beforehand. The amount of damage created was evaluated by measuring D depth profiles and D thermal desorption spectra. Results were compared with data obtained in a sequential experiment where samples were first irradiated by 10.8 MeV W ions and only afterwards exposed to 300 eV D ions at 450 K to populate the created defects. At 450 K we observed a two times higher maximum D concentration for the simultaneous case as compared with the sequential case. At 600 K and 800 K the ratio between the simultaneous and sequential case decreases to about 1.6 and 1.2, respectively, and increases again to a factor of 2 at 1000 K. We attribute this dependence on temperature to the change in the concentration of mobile and trapped D during the simultaneous exposures, which is in line with theoretical calculations predicting that trapped D in a vacancy prevents vacancy annihilation with self-interstitials.

Two series of measurements were performed in the JSI TRIGA research reactor in 2014 and 2017 to validate the ⁵⁵Mn(n,γ)⁵⁶Mn cross-sections and experimentally investigate the relationship between the ⁵⁵Mn(n,γ)⁵⁶Mn reaction and the rate of tritium production through the ⁶Li(n,t)⁴He reaction. Indeed, previously observed similarities between the sensitivity profiles of the neutron reaction of tritium production on lithium, ⁶Li(n,t)⁴He, and those of the ⁵⁵Mn(n,γ)⁵⁶Mn reaction in tritium breeder modules indicated that the latter reaction could be used as an effective monitor of tritium production, at least for short-term monitoring (the half-life of ⁵⁶Mn being 2.579 h). However, experimental verification, improvements and validation of the ⁵⁵Mn(n,γ)⁵⁶Mn cross-sections are needed in order to meet the required accuracy. Foils of certified reference material Al–1%Mn, as well as LiF thermoluminescent detectors and Li₂O samples were irradiated, both bare and under cadmium, to study the potential use of the ⁵⁵Mn(n,γ)⁵⁶Mn reaction for monitoring tritium production in fusion devices. Additionally, Al–0.1%Au was also irradiated for comparison, the ¹⁹⁷Au(n,γ)¹⁹⁸Au reaction cross-section being a standard. In order to obtain complementary information for data validation purposes, the irradiations were performed in positions within the JSI TRIGA reactor with different neutron spectra, i.e. in the central channel, the pneumatic tube and the F19 position, both in the outer 'F' ring of the reactor core and in the IC-40 irradiation channel located in the graphite reflector surrounding the reactor core. Bare and cadmium-covered irradiations were needed to subtract the contribution of epithermal neutrons to the ⁵⁵Mn(n,γ)⁵⁶Mn reaction. Calculations of the reaction rates were performed using the Monte Carlo code MCNP6.1 with a detailed model of the JSI TRIGA reactor, with the samples, the irradiation capsules and covers being modelled explicitly. The uncertainties involved in the measurements and the calculations were carefully evaluated. The principal objective was to study the energy response and correlations between the ⁵⁵Mn(n,γ)⁵⁶Mn reaction in irradiated Al–1%Mn and the ⁶Li(n,t)⁴He reaction in irradiated LiF and Li₂O. Good consistency between the measured and calculated ⁵⁵Mn(n,γ)⁵⁶Mn and ¹⁹⁷Au(n,γ)¹⁹⁸Au reaction rates, in most cases within the uncertainty bars, was observed.

Divertor Thomson scattering (DTS) and laser-induced fluorescence (LIF) are both laser aided diagnostics well suited to combination with common probing and collecting optics that are the most sophisticated and expensive part of any ITER optical diagnostic system. The combination of DTS and LIF are used for simultaneous measurement of local electron (T_e, n_e), ion (T_i, n_HeII) and atom (n_HeI, n_H(D,T)) parameters and provide basic information on rates of electron and ion processes to allow basic understanding of the physics of divertor plasma detachment. The measured parameters permit the calculation of rates of ionization and recombination using T_e, n_e, T_i, n_i, n_HeI and n_H(D,T); emission intensity—T_e, n_e, n_i, n_HeI and n_H(D,T); frictional force of the plasma flow due to collisions with neutrals—T_i, n_i, T₀, n_HeI and n_H(D,T) and pressure of the incoming plasma flow—T_e, n_e, T_i and n_i. The paper discusses the benefits of DTS and LIF integration, suggests new approaches to the estimation of DTS capability, LIF implementation and possibilities for further diagnostic development.

Experiments to combine a radiating divertor or mantle with high power, high β_N plasma operation have successfully reduced the divertor heat flux by 40% but have also encountered several challenging problems. For example, injecting either neon or argon seed impurities have led both to significant fuel dilution of the core plasma and to the emergence of harmful tearing mode activity which compromised plasma beta and energy confinement time. Increased divertor closure resulted in more effective control over the injected seed impurity inventory, although it was also observed that small radial variations in the placement of the outer separatrix strike point within the DIII-D 'closed' divertor could lead to clear differences in the impurity build-up inside the core plasma. Active particle pumping of argon seed impurities through the divertor leg on the high-field side of a double-null divertor configuration showed little effect on seed impurity inventory and only modest control over the deuterium fuel inventory. Typically, 80% or greater of injected argon was removed by the corresponding cryo-pump on the low-field (outboard) side of that divertor, with the remaining argon being removed by the outboard cryo-pump from the opposite divertor. For single-null divertor cases, the contribution of the inner (high-field) cryo-pump to impurity and main-ion control appears to be complicated by the levels of divertor recycling and degree of divertor detachment. In general, divertor design that impedes the escape of seed impurities, plasma shaping that allows a higher stability limit, and electron cyclotron deposition profiles that prevent tearing mode onset and impede impurity accumulation in the core are under consideration for addressing the issues presented in this paper.

A new small angle slot (SAS) divertor concept has been developed to enhance neutral cooling across the divertor target by coupling a closed slot structure with appropriate target shaping. Initial tests on DIII-D find a strong interplay between such anticipated 'SAS' effects and cross-field drifts, favouring operation with the ion B  ×  ∇B drift away from the X-point, as currently employed for advanced tokamaks. This offers the following key improvements relative to DIII-D's open lower divertor or partially-closed upper divertor: (i) SAS allows for transition to low temperature moderately detached divertor conditions with T_e  ≲  10 eV at very low main plasma densities, lower than are usually attainable at all in DIII-D high confinement (H-mode) plasmas as used in these tests; (ii) Pedestal performance and core confinement are significantly improved with SAS. The final confinement collapse associated with the onset of X-point MARFE (multifaceted asymmetric radiation from the edge) following deep detachment occurs at significantly higher pedestal densities, thus widening the window of H-mode operation compatible with a dissipative divertor. For operation with the ion B  ×  ∇B drift toward the X-point, the divertor plasma transitions to a bifurcative detached state at much higher densities, similar to other divertor configurations in DIII-D. These results highlight the strong interplay between divertor closure and drifts, and point to an interesting divertor optimization path to explore that offers potential for future fusion reactors.

We study the influence of grain boundaries (GBs) on radiation-induced vacancies, as well as on the hydrogen (H) behavior in tungsten (W) samples with different grain sizes in the temperature range from 300 K to 573 K, both experimentally and by computer simulations. For this purpose, coarse-grained and nanostructured W samples were sequentially irradiated with carbon (C) and H ions at energies of 665 keV and 170 keV, respectively. A first set of the implanted samples was annealed at 473 K and a second set at 573 K. Object kinetic Monte Carlo simulations were performed to account for experimental outcomes. Results show that the number of vacancies for nanostructured W is always larger than for monocrystalline W samples in the whole studied temperature range and that the number of vacancies is only reduced in samples with a large density of grain boundaries and at temperatures high enough to activate the vacancy motion (around 573 K). Results also indicate that the migration of H along vacancy free grain boundaries is more effective than along the bulk, and that the retained H is trapped in vacancies located within the grains. These results are used to explain the experimental outcomes.

This paper compares the gyrokinetic instabilities and transport in two representative JET pedestals, one (pulse 78697) from the JET configuration with a carbon wall (C) and another (pulse 92432) from after the installation of JET's ITER-like Wall (ILW). The discharges were selected for a comparison of JET-ILW and JET-C discharges with good confinement at high current (3 MA, corresponding also to low ) and retain the distinguishing features of JET-C and JET-ILW, notably, decreased pedestal top temperature for JET-ILW. A comparison of the profiles and heating power reveals a stark qualitative difference between the discharges: the JET-ILW pulse (92432) requires twice the heating power, at a gas rate of e s⁻¹, to sustain roughly half the temperature gradient of the JET-C pulse (78697), operated at zero gas rate. This points to heat transport as a central component of the dynamics limiting the JET-ILW pedestal and reinforces the following emerging JET-ILW pedestal transport paradigm, which is proposed for further examination by both theory and experiment. ILW conditions modify the density pedestal in ways that decrease the normalized pedestal density gradient a/L_n, often via an outward shift in relation to the temperature pedestal. This is attributable to some combination of direct metal wall effects and the need for increased fueling to mitigate tungsten contamination. The modification to the density profile increases , thereby producing more robust ion temperature gradient (ITG) and electron temperature gradient driven instability. The decreased pedestal gradients for JET-ILW (92432) also result in a strongly reduced shear rate, further enhancing the ion scale turbulence. Collectively, these effects limit the pedestal temperature and demand more heating power to achieve good pedestal performance. Our simulations, consistent with basic theoretical arguments, find higher ITG turbulence, stronger stiffness, and higher pedestal transport in the ILW plasma at lower .

Tungsten, the material used in the plasma-facing components (PFCs) of nuclear fusion reactors, develops a fuzz-like surface morphology under typical reactor operating conditions. This fragile 'fuzz' surface nanostructure adversely affects reactor performance and operation. Developing predictive models, capable of simulating the spatiotemporal scales relevant to the fuzz formation process is essential for understanding the growth of such extremely complex surface features and improving PFC and reactor performance. Here, we report the development of an atomistically-informed, continuous-domain model for the onset of fuzz formation in helium plasma-irradiated tungsten and validate the model by comparing its predictions with measurements from carefully designed experiments. Our study demonstrates that fuzz forms in response to stress induced in the near-surface region of PFCs as a result of plasma exposure and helium gas implantation. Our model sets the stage for detailed descriptions of this complex fuzz formation phenomenon and similar phenomena observed in other materials.

Heating neutral beam (HNB) injectors, necessary to achieve burning conditions and to control plasma instabilities in ITER, are characterized by such demanding parameters that a neutral beam test facility (NBTF) dedicated to their development and optimization is being realized in Padua (Italy) with direct contributions from the Italian government (through Consorzio RFX as the host entity) and the ITER international organization (with kind contributions from the ITER domestic agencies of Europe, Japan and India) and technical and scientific support from various European laboratories and universities. The NBTF hosts two experiments: SPIDER, devoted to ion source optimization for the required source performance, and MITICA, the full-size prototype of the ITER HNB, with an ion source identical to SPIDER.

This paper gives an overview of the progress towards NBTF realization, with particular emphasis on issues discovered during this phase of activity and on solutions adopted to minimize the impact on the schedule and maintain the goals of the facilities. The realization of MITICA is well advanced; operation is expected to start in 2023 due to the long procurement time of the in-vessel mechanical components. The beam source power supplies, operating at 1 MV, are in an advanced phase of realization; all high-voltage components have been installed and the complex insulation test phase began in 2018. At the same time, construction and installation of SPIDER plant systems was successfully completed with their integration into the facility. The mechanical components of the SPIDER ion source were installed inside the vessel and connected to the plants. Integrated commissioning with the control, protection and safety systems ended positively and the first experimental phase has begun. The first results of the SPIDER experiment, with data from operational diagnostics, and the plans for the 1 MV insulation tests on the MITICA high-voltage components are presented.

The scrape off layer (SOL) plasma in a Tokamak device and its coupling with the edge dictate the performance of a discharge to a high degree—especially as all plasma has to go through the SOL, which is the main exhaust channel for the hot plasma. This contribution provides an introduction to the modelling efforts of the plasma dynamics in the SOL and the edge regions. We employ a fully dynamical fluid model, the HESEL code. HESEL is equipped with synthetic diagnostic tools as probe arrays, Li-beam spectroscopy, and Gas Puff Imaging. Using the synthetic probe arrays to measure the electron and ion heat advection and conduction we obtain the upstream power fall-off length for a broad range of plasma parameters and by applying non-linear fitting procedures we derive the scaling of the fall-off length. The obtained results are in agreement with recent experimental observations from L-mode ASDEX Upgrade data. A workflow for generating synthetic Lithium beam data, where the fluctuation data from HESEL are passed to the RENATE code are discussed using experimental results from ASDEX Upgrade.

The expansion of the DIII-D tokamak operational space via the addition of new non-axisymmetric coil rows is investigated using the linear MHD plasma response code MARS-F. The plasma response to magnetic perturbations imposed by five distinct low- and high-field side coil geometries is modeled for toroidal mode numbers up to n  =  6. A set of linear (: resonant, kink mode amplitude) and quadratic (: core and edge integrated NTV torque) plasma metrics are defined and compared directly for the various coil geometries. Coil size is also scanned as a secondary optimization variable. An in-vessel coil located at the low-field side midplane is identified as the most efficient plasma response actuator across all metrics and toroidal mode numbers. The multi-row combination of this new midplane (M) coil with the existing DIII-D non-axisymmetric coil set is considered in numerical optimization studies to quantify the impact of this new coil on modeled plasma response metrics. The addition of the M coil is shown to significantly increase the actuator efficiency in maximizing individual plasma response metrics as well as maximizing multi-modal ratios of response metrics, e.g. maximizing resonant modes or edge NTV torque while minimizing kink modes or core NTV torque. The ex-vessel midplane (C) coil is also observed to have an outsize effect on multi-modal optimization due to its uniquely narrow poloidal spectra, and the expansion of optimization space due to the combined use of the two existing in-vessel coil rows with both the M and C coils is notably larger than either coil's individual contribution even though both share similar poloidal locations. This four-row system is expected to greatly extend abilities to understand and optimize the control of transients and rotation on the DIII-D tokamak.

During International Thermonuclear Experimental Reactor operation, due to plasma–wall interaction, particles/dust will be created in sizes ranging from nanometers to tens of microns. The dust properties, especially their ability to be covered by a thin oxide electrostatic insulating layer, and surface topology deeply affect their tritium inventory. Consequently, physico-chemical properties specific to tritiated tungsten particles and consequence on particle behavior in the facility and environment must be carefully assessed. For size-relevant tungsten particles, the measured tritium inventory is ~10 GBq g⁻¹. However, it varies with the particle specific surface area. Due to tritium beta decay and the oxide-insulating layer, dust exhibits a positive electrostatic self-charging. For a 5 µm particle in diameter with a 10 GBq g⁻¹ tritium inventory, self-charging rate could lead to 5.5 10⁴ elementary electric charges per day. These electrostatic properties could change the adhesion of dust on walls. In the case of a single particle, the adhesion will be reinforced due to image and dielectric forces. However, if the tritiated particle is part of an aggregate, the adhesion remains unknown. Due to the limited free path of the β emission in material, the tritium inventory carried by airborne particles cannot be measured in real time by conventional continuous radioactive aerosols monitors, and a new measurement strategy is needed for atmospheric surveillance in the workplace and of facility exhaust. Toxicity studies dealing with exposure to untritiated/tritiated tungsten particles of 100 nm have been undertaken. It was shown that these particles are rapidly dissolved in biologic media. Finally, after collection, dust must be confined to avoid its spreading into the environment. Different technical solutions are presented in this paper.

In this comment, we point out possible critical numerical flaws of recent particle simulation studies (Jiang et al 2016 Nucl. Fusion56 126017; Peng et al 2018 Nucl. Fusion58 026007) on the electrical gas breakdown in a simple one-dimensional periodic slab geometry. We show that their observations on the effects of the ambipolar electric fields during the breakdown, such as the sudden reversal of the ion flow direction, could not be real physical phenomena but result from numerical artifacts violating the momentum conservation law. We show that an incomplete implementation of the direct-implicit scheme can cause the artificial electric fields and plasma transports resulting in fallacies in simulation results. We also discuss that their simple plasma model without considering poloidal magnetic fields seriously mislead the physical mechanism of the electrical gas breakdown because it cannot reflect important dominant plasma dynamics in the poloidal plane (Yoo et al 2018 Nat. Commun. 9 3523).