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

Compressional Alfvén eigenmodes (CAEs) driven by energetic ions have been observed in magnetic fusion experiments. In this paper, we show that the modes can also be driven by runaway electrons formed in post-disruption plasma, which may explain kinetic instabilities observed in DIII-D disruption experiments with massive gas injection. The spatial structure is calculated, as are the frequencies which are in agreement with experimental observations. Using a runaway electron distribution function obtained from a kinetic simulation, the mode growth rates are calculated and found to exceed the collisional damping rate when the runaway electron density exceeds a threshold value. The excitation of CAEs poses a new possible approach to mitigate seed runaway electrons during the current quench and surpassing the avalanche. [ABSTRACT FROM AUTHOR]

During the ITER design phase, the focus of ITER boundary plasma modeling activities has been on divertor performance under baseline H-mode, fusion power operation (FPO) conditions. However, early ITER operation will be primarily with hydrogen fuel in L-mode, in the pre-fusion power operation 1 (PFPO-1) phase. Here, the SOLPS-ITER code is used to evaluate divertor performance during this non-active phase. To verify the assumptions used in the existing high power simulation database, gas throughput scans were performed for two types of divertor surface material (beryllium and tungsten) and two gas puff locations (divertor and main chamber). The adoption of beryllium target surfaces simulates the effect of main chamber material erosion and migration and, along with main chamber gas injection, is the current default for the high power database. Depending on the divertor surface material, the atom to molecule ratio of the recycled neutral particles varies. This modifies the momentum and power loss mechanisms arising from plasma–neutral interactions. However, since the effect of atomic and molecular reactions are compensatory, the 'total' power and momentum losses are relatively insensitive to the target surface material. Similarly, the impact of gas puff location on divertor plasma parameters is not significant, though main chamber injection provides an additional ionization source in the upstream scrape-off layer (SOL) and leads to moderate changes in the upstream density and far SOL parameters. However, these effects can be neglected within the available range of the gas puff and pump rates in ITER. Since beryllium and tungsten are materials at both extremes in terms of surface reflection properties, the conclusions may be applicable to other divertor surface materials. An important additional finding of the study is that the insensitivity of upstream density to divertor neutral pressure found in the FPO database is also recovered in these PFPO-1 simulations. [ABSTRACT FROM AUTHOR]

Disruption mitigation remains a critical and unresolved issue for ITER. Measurement uncertainties preventing quantification of a system's efficacy remains a significant hurdle in producing and validating a viable disruption mitigation system. This study addresses this issue through the creation of a dataset on the ASDEX Upgrade tokamak targeted at developing analysis techniques to quantify mitigation that are applicable irrespective of the disruption mitigation system being studied. These experiments used a range of thermal energy fractions to address concerns for ITER from JET massive gas injection (MGI) experiments showing a decrease in efficacy with increasing thermal energy fraction. The dataset produced in this study used MGI valves in two toroidal locations. The high resolution foil bolometers and AXUV diode arrays were used to infer radiation emission profiles at varying toroidal distances from the injection location. The fueling efficiency of the two systems was found to be comparable and toroidal asymmetries over the entire disruption were found to be negligible. The AXUV diodes were cross-calibrated with the foil bolometers and used to estimate the thermal energy radiated and magnetic energy coupled to the vessel structure and coils. It was estimated that 75–95% of the thermal energy was radiated and an almost constant 60% of the total magnetic energy was coupled. Radiated energy fractions of 0.8–1.0 were calculated and no decrease as a function of thermal energy fraction was found. A ± 20% variation in the coupled magnetic energy was explored and it did not alter this trend. [ABSTRACT FROM AUTHOR]

We present the first successful simulation of an induced disruption in ASDEX Upgrade from massive material injection (MMI) up to established runaway electron (RE) beam, thus covering pre-thermal quench, thermal quench and current quench (CQ) of the discharge. For future high-current fusion devices such as ITER, the successful suppression of REs, e.g. through MMI, is of critical importance to ensure the structural integrity of the vessel. To computationally study the interplay between MMI, background plasma response, and RE generation, a toolkit based on the 1.5D transport code coupling ASTRA-STRAHL is developed. Electron runaway is described by state-of-the-art reduced kinetic models in the presence of partially ionized impurities. Applied to argon MMI in ASDEX Upgrade discharge #33 108, key plasma parameters measured experimentally, such as temporal evolution of the line averaged electron density, plasma current decay rate and post-CQ RE current, are well reproduced by the simulation presented. Impurity ions are transported into the central plasma by the combined effect of neoclassical processes and additional effects prescribed inside the q = 2 rational surface to explain experimental time scales. Thus, a thermal collapse is induced through strong impurity radiation, giving rise to a substantial RE population as observed experimentally. [ABSTRACT FROM AUTHOR]

Disruptions are a major threat to future tokamaks including ITER. They generate excessive electromagnetic forces, heat loads and multi-MeV runaway electrons. The runaway electron beam carries the risk of in-vessel component damage and even the structures beyond them. Thus, prevention of the runaway beam generation or the mitigation of the developed beam is of prime importance. In JET ITER-like wall, the runaway electron beams triggered by massive gas injection (MGI) coexists with a cold background plasma. Lines corresponding to the higher ionization states of argon are observed in VUV spectra outside of the runaway region suggesting a hot background plasma.Using the quantitative analysis of the VUV spectroscopy, the temperature profiles of the background plasmas are estimated using a synthetic line ratios method. The background plasmas at JET-ILW are found to be hotter than other tokamaks where mitigation of the runaway electron beam was unconditionally successful. The volume-averaged Te is found to increase linearly with the gas amount used to trigger the disruption and the electron density in the far scrape-off layer. It is independent of other background plasma properties. A 0D/1D power balance of the post-disruption physical systems is made using the characteristics of the background plasma. The collisional power loss of the runaway electron beam is the primary power source heating the background plasma. [ABSTRACT FROM AUTHOR]

The paper describes under which plasma and machine conditions runaway electrons (REs) are generated during ASDEX Upgrade plasma disruptions. The REs are created by argon injection to investigate methods of current and energy dissipation. The RE beams are stable and last up to a few hundred milliseconds. The experimental findings are described and made available for the validation of theoretical models. In the following, a simple 0D fluid model is used to simulate the observed RE current magnitude and time behavior. In spite of its simplicity, the model is consistent with the measured RE current. The injection of heavy gases into the RE beam is then discussed in some detail: the injected gas penetrates into the low density background plasma and through the beam without ionizing significantly. Therefore its effect on the REs can be tested directly and it is found to be consistent with the known Coulomb collisional theory. [ABSTRACT FROM AUTHOR]

This work provides a qualitative description of the pellet response to the ambient runaway electrons. For ITER-relevant parameters, our estimates suggest that the cryogenic pellets will be sublimated instantly at the edge of the runaway beam. The subsequent rapid expansion of the sublimated material spreads the impurities over the poloidal cross-section of a tokamak on a millisecond time scale prior to the complete ionization of the expanding cloud. The injected solid pellet turns into a rapidly expanding gas cloud before it reaches the core of the runaway beam. As a result, the pellet acts similar to the massive gas injection. [ABSTRACT FROM AUTHOR]

The paper presents an analysis of disruptions occurring during JET-ILW plasma operations covering the period from the start of ILW (ITER-like wall) operation up to completion of JET operation in 2016. The total number of disruptions was 1951 including 466 with deliberately induced disruptions. The average rate of unintended disruptions was 16.1 %, which is significantly above the ITER target at 15 MA. The pre-disruptive plasma parameters are: plasma current Ip = (0.82–3.38) MA, toroidal field BT = (0.98–3.4) T, safety factor q95 = (1.52–9.05), plasma internal inductance li = (0.58–1.86), Greenwald density limit fraction FGWL = (0.04–1.61), with 720 X-point plasma pulses from a subset of 1420 unintended disruption shots. Massive gas injection (MGI) has been routinely used in protection mode both to terminate pulses when the plasma is at risk of disruption and to mitigate against disruption effects. The MGI was mainly triggered by the n = 1 locked mode (LM) amplitude exceeding a threshold or by the disruption itself, namely, either dIp/dt (specifically, a fast drop in Ip) or the toroidal loop voltage exceeding threshold values. For mitigation purposes, only the LM was used as a physics precursor and threshold on the LM signal was used to trigger the MGI prior to disruption. Long lasting LM (≥ 100 ms) do exist prior to disruption in 75% of cases. However, 10% of non-disruptive pulses have a LM which eventually vanished without disruption. The plasma current quench (CQ) may result in 3D configurations, termed as asymmetrical disruptions, which are accompanied by sideways forces. Unmitigated vertical displacement events (VDEs) generally have significant plasma current toroidal asymmetries. Unmitigated non-VDE disruptions also have large plasma current asymmetries presumably because there is no plasma vertical position control during the CQ and so they too are subject to large vertical displacements. MGI is a reliable tool to mitigate 3D effects and correspondingly sideways forces during the CQ. The vessel structure loads depend on the force impulse and force time behaviour, including their rotation. The toroidal rotation of 3D configuration may cause resonance with the natural frequencies of the vessel components in large tokamaks such as ITER. The JET-ILW amplitude-frequency interdependence of toroidal rotation of 3D configurations is presented. [ABSTRACT FROM AUTHOR]

This paper investigates the effect of the ITER-like wall (ILW) on runaway electron (RE) generation through a comparative study of similar slow argon injection JET disruptions, performed with different wall materials. In the carbon wall case, a RE plateau is observed, while in the ITER-like wall case, the current quench is slower and the runaway current is negligibly small. The aim of the paper is to shed light on the reason for these differences by detailed numerical modelling to study which factors affected the RE formation. The post-disruption current profile is calculated by a one-dimensional model of electric field, temperature and runaway current taking into account the impurity injection. Scans of various impurity contents are performed and agreement with the experimental scenarios is obtained for reasonable argon and wall impurity contents. Our modelling shows that the reason for the changed RE dynamics is a complex, combined effect of the differences in plasma parameter profiles, the radiation characteristics of beryllium and carbon, and the difference of the injected argon amount. These together lead to a significantly higher Dreicer generation rate in the carbon wall case, which is less prone to being suppressed by RE loss mechanisms. The results indicate that the differences are greatly reduced above ∼50% argon content, suggesting that significant RE current is expected in future massive gas injection experiments on both JET and ITER. [ABSTRACT FROM AUTHOR]

Following the completion in May 2011 of the shutdown for the installation of the beryllium wall and the tungsten divertor, the first set of JET campaigns have addressed the investigation of the retention properties and the development of operational scenarios with the new plasma-facing materials. The large reduction in the carbon content (more than a factor ten) led to a much lower Zeff (1.2–1.4) during L- and H-mode plasmas, and radiation during the burn-through phase of the plasma initiation with the consequence that breakdown failures are almost absent. Gas balance experiments have shown that the fuel retention rate with the new wall is substantially reduced with respect to the C wall. The re-establishment of the baseline H-mode and hybrid scenarios compatible with the new wall has required an optimization of the control of metallic impurity sources and heat loads. Stable type-I ELMy H-mode regimes with H98,y2 close to 1 and βN ∼ 1.6 have been achieved using gas injection. ELM frequency is a key factor for the control of the metallic impurity accumulation. Pedestal temperatures tend to be lower with the new wall, leading to reduced confinement, but nitrogen seeding restores high pedestal temperatures and confinement. Compared with the carbon wall, major disruptions with the new wall show a lower radiated power and a slower current quench. The higher heat loads on Be wall plasma-facing components due to lower radiation made the routine use of massive gas injection for disruption mitigation essential. [ABSTRACT FROM AUTHOR]

This paper presents a summary of the papers presented at the 24th IAEA Fusion Energy Conference (San Diego, CA, October 2012) on magnetic confinement theory and modelling. [ABSTRACT FROM AUTHOR]

The DIII-D research program is addressing key ITER research needs and developing the physics basis for future steady-state tokamaks. Pellet pacing edge-localized mode (ELM) control in the ITER configuration reduces ELM energy loss in proportion to 1/fpellet by inducing ELMs at up to 12× the natural ELM rate. Complete suppression of ELMs with resonant magnetic perturbations has been extended to the q95 expected for ITER baseline scenario discharges, and long-duration ELM-free QH-mode discharges have been produced with ITER-relevant co-current neutral-beam injection (NBI) using external n = 3 coils to generate sufficient counter-Ip torque. ITER baseline discharges at βN ∼ 2 and scaled NBI torque have been maintained in stationary conditions for more than four resistive times using electron cyclotron current drive (ECCD) for tearing mode suppression and disruption avoidance; active tracking with steerable launchers and feedback control catch these modes at small amplitude, reducing the ECCD power required to suppress them. Massive high-Z gas injection into disruption-induced 300–600 kA 20 MeV runaway electron (RE) beams yield dissipation rates ∼10× faster than expected from e–e collisions and demonstrate the possibility of benign dissipation of such REs should they occur in ITER. Other ITER-related experiments show measured intrinsic plasma torque in good agreement with a physics-based model over a wide range of conditions, while first-time main-ion rotation measurements show it to be lower than expected from neoclassical theory. Core turbulence measurements show increased temperature fluctuations correlated with sharply enhanced electron transport when exceeds a critical-gradient scale length. In H-mode, data show the pedestal height and width growing between ELMs with ∇P at the computed kinetic-ballooning limit, in agreement with the EPED model. Successful modification of a neutral-beam line to provide 5 MW of adjustable off-axis injection has enabled sustained operation at βN ∼ 3 with broader current and pressure profiles at higher qmin than previously possible, though energy confinement is lower than expected. Initial experiments aimed at developing integrated core and boundary solutions demonstrated heat flux reduction using enhanced edge radiation from neon injection and innovative divertor geometries (e.g. snowflake configuration). [ABSTRACT FROM AUTHOR]

An overview of the recent research work on the J-TEXT tokamak over the last two years is presented. A series of experiments and simulations of the interaction between resonant magnetic perturbations (RMPs) and plasma were carried out on the J-TEXT tokamak. The results show that the m/n = 2/1 (m and n are the poloidal and toroidal mode numbers, respectively) mode locking is obtained with sufficiently large RMPs. And suppression of the m/n = 2/1 tearing mode by moderate magnetic perturbation amplitude is also observed. With experimental parameters as input, both mode locking and mode suppression by RMPs are simulated by nonlinear numerical modelling based on reduced magnetohydrodynamic equations. The simulations are in good agreement with the experimental observations. Density modulation using gas puffing is carried out on J-TEXT to evaluate the particle transport parameters in a typical J-TEXT discharge, including diffusion coefficient and convective velocity. Inverse sawtooth-like activity caused by neon gas injection is observed. The inverse sawtooth-like activity occurs only when the amount of neon impurity exceeds a threshold. Nevertheless, other impurities such as helium and argon cannot trigger such events. With the aid of a soft x-ray detector array, the runaway electron beam following disruptions is visible directly. A high-resolution far infrared polarimeter/interferometer, based on a three-wave technique, was developed and it observes the perturbations associated with sawtooth and tearing mode activities; the first result of the current density profile reconstruction is provided. An x-ray imaging crystal spectrometer is designed to receive the Kα line of Ar XVII and its satellites. The electron temperature obtained from line ratios of the W line to its satellites is 750 eV, and the ion temperature deduced from the Doppler broadening of the W line is 330 eV. [ABSTRACT FROM AUTHOR]

Experiments with boron-coated side limiters of two antennas operated together in 2012 showed that the side limiters are responsible for more than half of the increased W content in the plasma. Together with the contribution from the other limiter tiles, not replaced in 2012, the limiters account for at least two thirds of the Wcontent. A modified test two-strap ion cyclotron range of frequency (ICRF) antennas in ASDEX Upgrade with broad limiters and narrow straps has shown an improved operation with full W wall in 2011/2012 campaigns with up to a 40% lower rise of W concentration allowing more stable operation at low deuterium gas injection rate. Limiter spectroscopy measurements indicate up to a 40% reduction of the rise of the W sputtering yield during ICRF power, measured under the assumption of negligible influence of geometry variations and reflections on the measurements. The boron limiters on two antennas together with the improved broad-limiter antenna allowed a successful ICRF operation in 2012. As a part of long-term strategy of antenna design development, two three-strap antennas with phase and power balance control for reduction of E∥ are planned for installation in the future. [ABSTRACT FROM AUTHOR]

Shattered pellet injection (SPI) has been adopted as the baseline disruption mitigation system for ITER, as the radiative payload penetration into DIII-D plasmas from SPI is superior to those using the massive gas injection (MGI) method. Because of the substantial differences in the energy content of ITER plasma and those in present experiments, reliable 3D MHD modeling, benchmarked against present experiments is needed to project to ITER plasmas. In support of these needs, the depth of SPI fragment penetration in DIII-D plasmas was investigated by injecting SPI into two discharges with vastly different energy content and pedestal height. 400 Torr-L pure Ne fragmented pellets at a velocity of about 200 m s−1 were injected into a 0.2 MJ L-mode discharge and a 2 MJ super H-mode discharge. Results show deep penetration of SPI fragments into low-energy plasmas in DIII-D. SPI fragment penetration is reduced as the plasma energy content increases, with some discharges exhibiting penetration that is confined to the outer regions of the plasma. The injected SPI fragments are also spread out over a distance of about 20 cm, which results in some fragments arriving near the end of or after the thermal quench is over. [ABSTRACT FROM AUTHOR]

The effect of a periodic train of short gas-puff pulses on the rotation frequency and amplitude of drift-tearing modes has been studied in ADITYA/ADITYA-U tokamak. The short gas puffs, injecting approximately molecules of fuel gas (hydrogen) at one toroidal location, are found to concomitantly decrease the drift-tearing mode rotation frequency and the mode amplitude during the period of injection and then recover back to its initial values when the gas pulse is over. This leads to a periodic modulation of the rotation frequency and amplitude of the drift-tearing modes that is correlated with the periodicity of the gas pulse injection. The underlying mechanism for this change in the mode characteristic appears to be related to gas puff induced change in the radial profile of the plasma pressure in the edge region that brings about a reduction in the diamagnetic drift frequency. Detailed experimental measurements and BOUT++ code simulations support such a reduction in diamagnetic drift frequency. Our results reveal a close interaction between the edge dynamics and core MHD phenomena in a tokamak that could help us better understand the rotation dynamics and amplitude pulsations of magnetic islands. [ABSTRACT FROM AUTHOR]

The operational range and the reactor relevance of the TCV experiments are being enhanced by two sets of major upgrades. The first includes the installation of neutral beam injection (NBI) and new electron cyclotron (EC) auxiliary heating sources, to reach ITER relevant beta values and vary the electron to ion temperature ratio. A 15–30 keV, 1 MW tangential NBI system has been operational on TCV since 2015. A second beam of 1 MW, 50–60 keV ion energy, also aligned tangentially but opposite to the first beam, is foreseen to approach beta limits, vary the applied torque through zero and probe suprathermal ion physics. For the EC power, two 0.75 MW gyrotrons at the second harmonic have been installed. The next step will add two 1 MW dual frequency gyrotrons, one of which is currently being commissioned. These heating upgrades will increase the total available power for high-density plasmas from 1.25 MW to 5.0 MW. The rest of the upgrade consists of installing an in-vessel structure to form a divertor chamber of increased closure, to reach higher neutral divertor densities and impurity compression and thereby extend TCV divertor regimes toward more reactor relevant conditions for conventional and advanced divertor configurations. Graphite gas baffles will be installed inside the TCV vessel to delineate divertor and main chamber regions. The first set of baffles features 32 tiles on the high and 64 tiles on the low-field side, with geometry guided by simulations performed using the SOLPS-ITER code. The baffles are expected to be effective for a wide range of divertor configurations, including snowflake and super-X divertors, yet maintain plasma close to the inner wall for improved passive stabilization. The baffle dimensions may be varied in the future to modify the divertor closure. Control of the plasma, neutral and impurity densities will be achieved by a combination of toroidally distributed gas injection valves and impurity seeding, and a possible addition of cryo-condensation pumps. Significant diagnostic developments will be undertaken, to better characterize the divertor plasma, measure power and particle deposition at the strike points, and, specifically, improve our physics understanding of the detachment process. [ABSTRACT FROM AUTHOR]

Disruption means a sudden loss of confinement during a discharge in fusion reactors. Due to the huge electromagnetic loading and thermal loading on the facility and a large number of runaway electrons generated during disruptions, it is essential to find a method to predict the disruptions, so that measures like massive gas injection can be taken to mitigate or to avoid these harmful effects. In this research, a machine learning model mainly based on a 1.5-dimensional convolutional neural network, which is good at dealing with signals from multi-channels with great divergence, is trained to predict disruptions in the HL-2A tokamak. The disruption predictor uses shots 20000–29999 in HL-2A to train the machine learning model, and uses shots 30000–31999 to optimize hyper parameters. When tested on shots 32000–36000 in HL-2A, it reaches a true positive rate of 92.2% and a true negative rate of 97.5% with 30 ms before the disruption. [ABSTRACT FROM AUTHOR]

Locked modes (LMs) will be one of the major causes of disruptions in the ITER tokamak. Disruption mitigation systems (DMSs), such as massive gas injection (MGI) or shattered pellet injection (SPI), are expected to be deployed in pre-disruption discharges with large pre-existing locked modes. A series of target plasmas with an m/n  =  2/1 locked mode induced by resonant magnetic perturbation (RMP) penetration was terminated by an MGI on the J-TEXT tokamak. The penetration of the injected impurities during the process was diagnosed using a fast frame visible camera and a multi-channel polarimeter-interferometer (POLARIS) in combination. The electron temperature evolution during thermal quench (TQ) is also shown in detail. It is found that both the phase and width of the 2/1 mode have an effect on MGI shutdown dynamics. When the mode is larger than the critical width, the penetration depth and assimilation of impurities can be enhanced during pre-TQ, leading to a faster quenching process if the relative phase between the O-point of the 2/1 mode and the MGI valve is   +90°. Conversely, the penetration depth and assimilation of impurities are suppressed, leading to a slower TQ when the relative phase is   −90°. The toroidal radiation asymmetry is worse with the locked mode. The results suggest that the 3D effect between the injected impurities and the 2/1 locked mode is important during the disruption mitigation process. [ABSTRACT FROM AUTHOR]

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 τp/(1  −  R), where τp 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. [ABSTRACT FROM AUTHOR]

In view of improving ion cyclotron range of frequencies (ICRF) wave coupling in EAST (experimental advanced superconducting tokamak) necessary to obtain high-power discharges, two series of experiments in L-mode were devoted to the use of local gas injection as a tool to increase density in front of the antennas. During the first session, the quantity of deuterium simultaneously puffed from five different poloidal positions on each side of one ICRF antenna was gradually increased. Steady gas puffing on this antenna (~8.1020 e s−1) increased antenna coupling resistance by 100%, but puffing more gas did not provoke any further improvement. Plasma central electron density was not kept constant when using local puff but kept increasing during the discharges. During the second session, the global density was well-controlled and remained constant, the distance between the separatrix and the first wall was increased by 2 cm allowing, the gas to spread better, and puffing more gas this time lead to better wave coupling. The same amount of gas was then puffed from three different locations and changes in SOL (scrape-off layer) parameters were measured from several locations. Extensive analysis of the edge diagnostics made it possible to observe an electron density increase and a temperature decrease caused by the ionization of neutrals in the SOL. Heat loads on the divertor targets were slightly mitigated by gas injection, which is favorable from the perspective of long pulse scenario development. Injection at mid-plane led to the best results with the highest coupling simultaneously achieved for both antennas and smallest heat loads on the divertor target (below 1 MW m−2). As far as wave coupling efficiency is concerned, most experimental trends were also successfully reproduced by antenna modelling with the RAPLICASOL finite element code with experimental density profiles from reflectometry as the main input. [ABSTRACT FROM AUTHOR]

The injection of a large amount of impurities is one of the possible ways of mitigating disruption in large-scale tokamaks. The deposition of impurities at the center of the plasma is the key to the radiation of plasma energy and suppression of runaway. The interaction of the gas jet with the rational surfaces has been studied by scanning the plasma current. The experimental results show that the injection of a massive amount of argon can cool the plasma from the edge to the core region, and the cooling process is accompanied by different magnetohydrodynamic (MHD) modes when the gas jet reaches the corresponding rational surfaces. It is observed that with different edge safety factors and electron density, gas injection can induce different poloidal modes at first. Then, the poloidal mode traverses to lower m (where m is the poloidal mode number) MHD activities until a 2/1 mode is initiated and a thermal quench is started. The experimental results show that the penetration of a gas jet across the rational surfaces is faster in the plasmas with pre-existing large 2/1 tearing modes, which indicates that the 2/1 mode plays an important role in the penetration process. Disruptions triggered by supersonic molecular beam injection display a slower cooling process compared with massive gas injection, which can be divided into four stages. The dominant poloidal mode transition from m  =  3 to m  =  2 is associated with electron temperature recovery. [ABSTRACT FROM AUTHOR]

We report on the first demonstration of dissipation of fully avalanched post-disruption runaway electron (RE) beams by shattered pellet injection in the DIII-D tokamak. Variation of the injected species shows that dissipation depends strongly on the species mixture, while comparisons with massive gas injection do not show a significant difference between dissipation by pellets or by gas, suggesting that the shattered pellet is rapidly ablated by the relativistic electrons before significant radial penetration into the runaway beam can occur. Pure or dominantly neon injection increases the RE current dissipation through pitch-angle scattering due to collisions with impurity ions. Deuterium injection is observed to have the opposite effect from neon, reducing the high-Z impurity content and thus decreasing the dissipation, and causing the background thermal plasma to completely recombine. When injecting mixtures of the two species, deuterium levels as low as  ∼10% of the total injected atoms are observed to adversely affect the resulting dissipation, suggesting that complete elimination of deuterium from the injection may be important for optimizing RE mitigation schemes. [ABSTRACT FROM AUTHOR]

The paper is dedicated to the characterisation of multi-strap ICRH antenna coupling to plasma. Relevance of traditional concept of coupling resistance to antennas with mutually coupled straps is revised and the importance of antenna port excitation consistency for application of the concept is highlighted. A method of antenna S-matrix measurement in presence of plasma is discussed allowing deeper insight into the problem of antenna-plasma coupling. The method is based entirely on the RF plant hardware and control facilities available at JET and it involves application of variable phasing between the antenna straps during the RF plant operations at  >100 kW. Unlike traditional techniques relying on low-power (~10 mW) network analysers, the applied antenna voltage amplitudes are relevant to practical conditions of ICRH operations; crucially, they are high enough to minimise possible effects of antenna loading non-linearity due to the RF sheath effects and other phenomena which could affect low-power measurements. The method has been successfully applied at JET to conventional 4-port ICRH antennas energised at frequencies of 33 MHz, 42 MHz and 51 MHz during L-mode plasma discharges while different gas injection modules (GIMs) were used to maintain comparable plasma densities during the pulses. The S-matrix assessment and its subsequent processing yielding ‘global’ antenna coupling resistances in conditions of equalised port maximum voltages allowed consistent description of antenna coupling to plasma at different strap phasing, operational frequencies and applied GIMs. Comprehensive experimental characterisation of mutually coupled antenna straps in presence of plasma also provided a unique opportunity for in-depth verification of TOPICA computer simulations. [ABSTRACT FROM AUTHOR]

High pressure noble gas injection is a promising technique to mitigate the effect of disruptions in tokamaks. In this paper, results of mitigation experiments with low-Z massive gas injection (helium) on the EAST tokamak are reported. A fast valve has been developed and successfully implemented on EAST, with valve response time  ⩽150 μs, capable of injecting up to particles, corresponding to 300 times the plasma inventory. Different amounts of helium gas were injected into stable plasmas in the preliminary experiments. It is seen that a small amount of helium gas () can not terminate a discharge, but can trigger MHD activity. Injection of 40 times the plasma inventory impurity () can effectively radiate away part of the thermal energy and make the electron density increase rapidly. The mitigation result is that the current quench time and vertical displacement can both be reduced significantly, without resulting in significantly higher loop voltage. This also reduces the risk of runaway electron generation. As the amount of injected impurity gas increases, the gas penetration time decreases slowly and asymptotes to (∼7 ms). In addition, the impurity gas jet has also been injected into VDEs, which are more challenging to mitigate that stable plasmas. [ABSTRACT FROM AUTHOR]

The pedestal structure of type I ELMy H-modes has been analysed for JET with the ITER-like Wall (JET-ILW). The electron pressure pedestal width is independent of ρ* and increases proportionally to  √βpol,PED. Additional broadening of the width is observed, at constant βpol, PED, with increasing ν* and/or neutral gas injection and the contribution of atomic physics effects in setting the pedestal width cannot as yet be ruled out. Neutral penetration alone does not determine the shape of the edge density profile in JET-ILW. The ratio of electron density to electron temperature scale lengths in the edge transport barrier region, ηe, is of order 2–3 within experimental uncertainties. Existing understanding, represented in the stationary linear peeling–ballooning mode stability and the EPED pedestal structure models, is extended to the dynamic evolution between ELM crashes in JET-ILW, in order to test the assumptions underlying these two models. The inter-ELM temporal evolution of the pedestal structure in JET-ILW is not unique, but depends on discharge conditions, such as heating power and gas injection levels. The strong reduction in pe,PED with increasing D2 gas injection at high power is primarily due to clamping of half way through the ELM cycle and is suggestive of turbulence limiting the Te pedestal growth. The inter-ELM pedestal pressure evolution in JET-ILW is consistent with the EPED model assumptions at low gas rates and only at low beta at high gas rates. At higher beta and high gas rate the inter-ELM pedestal pressure evolution is qualitatively consistent with the kinetic ballooning mode (KBM) constraint but the peeling–ballooning (P–B) constraint is not satisfied and the ELM trigger mechanism remains as yet unexplained. [ABSTRACT FROM AUTHOR]

A new 1D radial fluid code, IMAGINE, is used to simulate the penetration of gas into a tokamak plasma during a massive gas injection (MGI). The main result is that the gas is in general strongly braked as it reaches the plasma, due to mechanisms related to charge exchange and (to a smaller extent) recombination. As a result, only a fraction of the gas penetrates into the plasma. Also, a shock wave is created in the gas which propagates away from the plasma, braking and compressing the incoming gas. Simulation results are quantitatively consistent, at least in terms of orders of magnitude, with experimental data for a D2 MGI into a JET Ohmic plasma. Simulations of MGI into the background plasma surrounding a runaway electron beam show that if the background electron density is too high, the gas may not penetrate, suggesting a possible explanation for the recent results of Reux et al in JET (2015 Nucl. Fusion55 093013). [ABSTRACT FROM AUTHOR]

Runaway currents following disruptions have an important effect on the first wall in current tokamaks and will be more severe in next generation tokamaks. The behavior of runaway currents in massive gas injection (MGI) induced disruptions have been investigated in the J-TEXT tokamak. The cold front induced by the gas jet penetrates helically along field lines, preferentially toward the high field side and stops at a location near the q  =  2 surface before the disruption. When the cold front reaches the q  =  2 surface it initiates magnetohydrodynamic activities and results in disruption. It is found that the MGI of He or Ne results in runaway free shutdown in a large range of gas injections. Mixture injection of He and Ar (90% He and 10%Ar) consistently results in runaway free shutdown. A moderate amount of Ar injection could produce significant runaway current. The maximum runaway energy in the runaway plateau is estimated using a simplified model which neglects the drag forces and other energy loss mechanisms. The maximum runaway energy increases with decreasing runaway current. Imaging of the runaway beam using a soft x-ray array during the runaway current plateau indicates that the runaway beam is located in the center of the plasma. Resonant magnetic perturbation (RMP) is applied to reduce the runaway current successfully during the disruption phase in a small scale tokamak, J-TEXT. When the runaway current builds up, the application of RMP cannot decouple the runaway beam due to the lower sensitivity of the energetic runaway electrons to the magnetic perturbation. [ABSTRACT FROM AUTHOR]

The pre–thermal–quench (PTQ) phase of the massive gas injection (MGI) scenario to terminate the tokamak plasma discharge is studied by means of one-dimensional (1D) transport simulations. This phase is characterized by the cold–front penetration in the hot plasma after the gas has been released from the valves, and before the actual thermal quench takes place, with consequent plasma disruption at lower stored energy. The comparison between the simulations and the ASDEX Upgrade (AUG) experiments allows to gain insight in the observed dependencies and time scales. Despite the genuine three-dimensional structure of the problem, it is shown that the 1D simulations are already giving experimentally relevant answers, the reason for which will be discussed in detail. Influence of unknown parameters and simplifying assumptions are also discussed. [ABSTRACT FROM AUTHOR]

This paper summarizes experimental results from recent studies on intrinsic rotation at the plasma edge of the TCABR tokamak. These results were obtained after upgrading the number of channels of the rotation diagnostic to three. The measurements were carried out in the collisional (Pfirsch–Schluter) regime and the rotation profiles of the ions were obtained from the Doppler shifts of the impurity carbon lines, CIII (464.74 nm), and CVI (529.06 nm). Results on the correlation between toroidal rotation at the plasma edge and direction of gas injection are also presented. They indicate that the direction of gas injection has a small effect on rotation; the velocity of the background neutral hydrogen is affected by direct momentum transfer from the injected gas (also hydrogen), while the carbon ions' velocity is affected by inward radial friction force between the injected gas atoms and ions, increasing their velocity in the opposite sense of the plasma current. [ABSTRACT FROM AUTHOR]

Disruptions are a major operational concern for next generation tokamaks, including ITER. They may generate excessive heat loads on plasma facing components, large electromagnetic forces in the machine structures and several MA of multi-MeV runaway electrons. A more complete understanding of the runaway generation processes and methods to suppress them is necessary to ensure safe and reliable operation of future tokamaks. Runaway electrons were studied at JET-ILW showing that their generation dependencies (accelerating electric field, avalanche critical field, toroidal field, MHD fluctuations) are in agreement with current theories. In addition, vertical stability plays a key role in long runaway beam formation. Energies up to 20 MeV are observed. Mitigation of an incoming runaway electron beam triggered by massive argon injection was found to be feasible provided that the injection takes place early enough in the disruption process. However, suppressing an already accelerated runaway electron beam in the MA range was found to be difficult even with injections of more than 2 kPa.m3 high-Z gases such as krypton or xenon. This may be due to the presence of a cold background plasma weakly coupled to the runaway electron beam which prevents neutrals from penetrating in the electron beam core. Following unsuccessful mitigation attempts, runaway electron impacts on beryllium plasma-facing components were observed, showing localized melting with toroidal asymmetries. [ABSTRACT FROM AUTHOR]

Measurements from the DIII-D tokamak show that toroidal radiation asymmetries during fast shutdown by massive gas injection (MGI) are largely driven by magnetohydrodynamic modes during the thermal quench. The phenomenology of these modes, which are driven unstable by profile changes as the thermal energy is quenched, is described based on detailed magnetic measurements. The toroidal evolution of the dominantly perturbation is understood to be a function of three parameters: the location of the MGI port, pre-MGI plasma rotation, and error fields. The resulting level of radiation asymmetry in these DIII-D plasmas is modest, with a toroidal peaking factor (TPF) of for the total thermal quench energy and for the peak radiated power, both of which are below the estimated limit for ITER (TPF 2) (Sugihara et al 2007 Nucl. Fusion47 337). [ABSTRACT FROM AUTHOR]

Simulations of massive gas injection for disruption mitigation in DIII-D are carried out to compare the toroidal peaking of radiated power for the cases of one and two gas jets. The radiation toroidal peaking factor (TPF) results from a combination of the distribution of impurities and the distribution of heat flux associated with the mode. When ignoring the effects of strong uni-directional neutral beam injection and rotation present in the experiment, the injected impurities are found to spread helically along field lines preferentially toward the high-field-side, which is explained in terms of a nozzle equation. Therefore when considering the plasma rest frame, reversing the current direction also reverses the toroidal direction of impurity spreading. During the pre-thermal quench phase of the disruption, the toroidal peaking of radiated power is reduced in a straightforward manner by increasing from one to two gas jets. However, during the thermal quench phase, reduction in the TPF is achieved only for a particular arrangement of the two gas valves with respect to the field line pitch. In particular, the relationship between the two valve locations and the 1/1 mode phase is critical, where gas valve spacing that is coherent with 1/1 symmetry effectively reduces TPF. [ABSTRACT FROM AUTHOR]

This paper describes the first development and implementation of a closed loop edge localized mode (ELM) frequency controller using gas injection as the actuator. The controller has been extensively used in recent experiments on JET and it has proved to work well at ELM frequencies in the 15–40 Hz range. The controller responds effectively to a variety of disturbances, generally recovering the requested ELM frequency within approximately 500 ms. Controlling the ELM frequency has become of prime importance in the new JET configuration with all metal walls, where insufficient ELM frequency is associated with excessive tungsten influx. The controller has allowed successful operation near the minimum acceptable ELM frequency where the best plasma confinement can be achieved. Use of the ELM frequency controller in conjunction with pellet injection has enabled investigations of ELM triggering by pellets while maintaining the desired ELM frequency even when pellets fail to trigger ELMs. [ABSTRACT FROM AUTHOR]

Experiments of disruption mitigation with massive gas injection are conducted in ASDEX Upgrade with fast valves located close to the plasma. The valves and the dedicated experiment are described in this paper. The dependence of the fuelling efficiency on plasma and gas parameters is documented and discussed. Several sources of uncertainties affecting its evaluation and physical interpretation have been addressed. An actual fuelling efficiency of 40% has been reached for neon injection with valves close to the plasma and for gas quantities relevant for the thermal and current quench mitigation of ITER. Refuelling the plasma after thermal quench is shown to be feasible; this result opens the possibility of raising the density in a runaway beam and therefore of increasing the collisional drag on and the radiative energy losses of the fast electrons. [ABSTRACT FROM AUTHOR]