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Energy dissipation in the plasma edge is key for future tokamaks. The potential of neon as radiating seeding species in disconnected double null (DDN) configuration is assessed in EAST discharges in high confinement mode (H-mode). As the separation between the two separatrices in the studied DDN discharges is minimum 1.5 cm, the configuration is effectively a single null configuration, and the benefits of the double null topology are minimal. Neon seeding, on the other hand, has a favourable effect: both the target heat flux and the divertor temperature decrease more than five-fold with increased seeding rate in high-recycling conditions. Interpretive edge plasma simulations with SOLPS-ITER in support of ongoing transport analysis are presented. For the unseeded case the numerical results agree with the experimental data within a factor two for the target temperature conditions and measured neutral pressures in the active divertor. The key for achieving good agreement is a suitable selection of coefficients for anomalous transport and neutral conductances between the upper cryopump and the main chamber.

Hot spot is a serious challenge limiting long pulse operation with LHCD (lower hybrid current drive) in tokamak. In order to mitigate the hot spot in guard limiter and improve LHCD capability, effect of edge plasma density on inducing hot spot and current drive has been studied in EAST. The temperature in the guard limiter of the LH antenna, inducing hot spot directly, increases with edge density and LH power. Studies show that the hot spot is mainly ascribed to the heat flux in front of LH antenna. Further simulation indicates that such spot corresponds to the peak position of edge density due to local LH electric field. In addition, due to the stronger parametric instability (PI) behavior in the case of higher edge density, the current drive capability decreases with edge density. Strike-point splitting behaviors appear with density increase, in agreement with current profiles in the edge region and the reduction of total driven current, suggesting that more power is deposited in edge region, which does more contribution to hot spot. Above studies offer one possible idea to optimize the edge density so as to satisfy the coupling, mitigate the heat flux in the guard limiter, and improve current drive capability for fusion device.

Lithium Beam Emission Spectroscopy (Li-BES) diagnostic has been developed on the Experimental Advanced Superconducting Tokamak (EAST), aiming at providing the edge electron density profile and density fluctuation. This system has been upgraded to 32 × 4 (radial × poloidal) channels since the 2016 campaign with a radial observation range of about 30 cm at the midplane on the low field side. The spatial resolution of Li-BES diagnostic is about 1 cm. The whole optical system of Li-BES diagnostic has been shifted 9 cm away from the tokamak axis perpendicular to the flange at horizontal D-port on EAST since the 2017 campaign. It’s confirmed that transmissivities of Avalanche PhotoDiode (APD) channels in the edge region are improved by comparing the relative light intensity calibration factors of the same row of APD channels before and after the shift. Spatial calibration and relative light intensity calibration of this system have been performed. Based on Schweinzer’s algorithm [J. Schweinzer et al., Plasma Phys. Control. Fusion 34, 1173 (1992)], the edge electron density profile is reconstructed and shown to be consistent with those provided by other diagnostics on EAST. The electron density profile evolution during the plasma current ramp-down process is observed with Li-BES diagnostic in both SOL region and pedestal region. With an injected Li-beam under the high frequency chopping control, the temporal resolution of the reconstructed electron density profile can be up to 10 μ s , which reveals temporal behaviors of fast transient events on EAST, such as the L-H transition and ELM crashes.

Liquid lithium (Li) can partly ameliorate lifetime and power-exhaust issues of plasma facing components (PFCs) by enabling a self-healing, self-replenishing surface with a reduced susceptibility to neutron damage in future fusion devices. To assess operational stability and heat-exhaust capability under tokamak exposure, two generations of continuously flowing liquid Li (FLiLi) limiters on the concept of a thin flowing Li film have been successfully designed and tested in high performance discharges in EAST. The design uses a circulating Li layer with a thickness of

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

Plasma-wall interactions at the first wall in fusion experimental devices are critical for the life time of plasma-facing components and tokamak operation. In-situ measurements of these plasma-wall interaction processes are required in long pulse devices in order to understand better the underlying mechanisms as well as monitoring of the tritium content of the first wall as an immediate safety application. In this work we present measurements of the deuterium retention in the first wall of EAST using laser-induced ablation spectroscopy. The diagnostic is applied during plasma operation and the dynamic retention for graphite and tungsten as plasma-facing materials are compared, showing more than three times higher short-term retention in graphite than in tungsten. Laser-based analysis is applied during a shift of the vertical positions of the plasma column in limited as well as in the start-up phase of discharges in diverted magnetic configuration. Plasma core densities of ne = (1.5–3.2) × 1019/m3 are chosen to vary the ion flux to the first wall and study saturation effects of deuterium on the surface first wall materials during tokamak plasma operation. A linear dependence of the retention from ion fluence is observed for graphite as PFC up to 3 × 1018D+/cm2, followed by a constant retention up to fluences of 3.5 × 1019D+/cm2 for nearby first wall temperatures of TPFC ≈ 350 K.

Both experimental modeling and numerical modeling were carried out on Experimental Advanced Superconductive Tokamak (EAST) to study the radiative divertor plasma behavior with neon (Ne) seeding. Steady H-mode radiation feedback control regimes (frad ∼ 18%–36%) were achieved by synergistic working of a divertor puff and supersonic molecular beam injection with Ne in the 2017 EAST campaign. Ne seeding effectively promoted divertor detachment while only causing a slight loss to the plasma stored energy in the experiment. To better understand the experimental result, analysis of the detachment process is introduced in this work. Two time slices namely before and after Ne seeding during the radiative divertor experiment in Shot #71021 were simulated by Scrape Off Layer Plasma Simulator (SOLPS)-ITER code. The simulation results for upstream plasma agreed well with the experiment, and the divertor profile agreed well on the outer targets when ignoring drifts. Radiation distribution was also calculated by SOLPS-ITER, showing that there was a high radiation region near the X-point, which is consistent with the experimental measurements.

Radio frequency (RF) power in the ion cyclotron range of frequencies (ICRF) is one of the primary auxiliary heating techniques for Experimental Advanced Superconducting Tokamak (EAST). In this paper, we will present an overview of the main experimental results from recent campaign on EAST. 1) A dedicated experiment for more than 10MW operations has been carried out on EAST. The major aspects linked to the use of the EAST auxiliary heating systems: ICRF,LHCD and NBI are presented with emphasis on ICRF and LH. The challenges for ICRF coupling on EAST will be described at the low densities typically used for long pulse operation with lithium wall conditioning. 2)This paper also summarizes effects of ICRF power on LH coupling and modification of heat loads on EAST. 3) Co-current rotation driven by ICRF are observed in both L-mode and H-mode discharges. Co-current rotation can be generated by ICRF heating in LHCD plasma. Different rotation behavior observed in LHCD discharges. ICRF can also induce counter-current rotation in LHCD L-mode discharges. 4) A high-frequency magnetic probes (HFBs) system has been developed and successfully applied to detect ion cyclotron emission (ICE) in D-NBI injection experiments in the 2018 experimental campaign of EAST. ICE driven by fast D-NBI ions and fusion born ions are observed on EAST.5) In Hydrogen minority in Deuterium plasma, the interaction of fast wave and fast D-NBI ions is detected on EAST. Acceleration of fast D ions is confirmed by increased neutron yield.

The redistribution of the divertor flux caused by the synergy of the supersonic-molecular-beam-injection (SMBI) and the magnetic perturbations induced by lower hybrid waves (LHWs), has been observed on the Experimental Advanced Superconducting Tokamak (EAST) [ Li J. {\it et al.} 2013 Nature Phys. 9 817]. To reveal the physical mechanism behind, first simulations with good qualitative agreements to the experimental findings are performed by utilizing a self-consistent fluid 3D edge plasma Monte-Carlo code coupled to a kinetic neutral particle transport code. The ions and electrons originating from the ionization of injected neutral particles in the plasma edge flow along the magnetic flux tube towards to the divertor, thus directly increasing the divertor flux on the split strike lines in the footprint. Combining this with the multi-lobe structure of the edge magnetic topology, actively controlling the divertor flux can be realized by adjusting the SMBI position or the phase of the magnetic perturbations.

A newly developed radial directional electron probe (DEP) has been applied to the unidirectional electron current measurement on EAST tokamak. The DEP consists of two radial arrays of channels which have opposite directions and align along the local magnetic field line. Each radial array has 6 holes with a radial interval of 5 mm. Every channel has a hole with 0.5 mm radial width, 3 mm depth and 15° poloidal opening angle. The graphite collector embedded inside the hole is biased to positive potential to repel low energy ions, and high energy ions are blocked by the hole surface because their Larmor radii are larger than the radial width of hole. In consequence, the ion current collected by the DEP collector can be ignored in contrast with the electron current, as demonstrated by the I-V characteristics in a DEP commissioning experiment. The difference of collected current between two opposite channels signifies the unidirectional electron current in the flux tube. In a lower hybrid wave (LHW) modulation experiment, the amplitude and radial structure of unidirectional electron current induced by LHW is measured directly by this radial DEP array, and the LHW filament current covers over 20 mm radial region with a maximum of 20 A/cm2.

The localized coupling among geodesic acoustic mode (GAM), tearing modes (TMs) and twin counter-propagating beta-induced Alfvén eigenmodes (BAEs) waves has been investigated in the experimental advanced superconducting tokamak. Before the appearance of TMs, typical continuous GAM is observed through the multi-channel Doppler backscattering (DBS) diagnostic. The twin BAEs can be excited after the burst of magnetic islands, which are localized to the q = 4 rational surface that is confirmed by the measurement of DBS array, where the GAM and twin BAEs are observed synchronically at R ≈ 2.23 m (normalized radius ρ ≈ 0.8). One reasonable excitation mechanism is proposed that the twin BAEs can be excited by the nonlinear interaction between GAM and magnetic islands. As the width of magnetic islands increases, the electromagnetic twin BAEs increase synchronically with the decreasing of electrostatic GAM, strongly suggesting that the electromagnetic components are pumped from three-wave interaction between electrostatic GAM and magnetic islands.

EAST was equipped with W-coated divertor plates and dome for the upper divertor, the remaining plasma facing components include the plasma facing components at lower divertor consisted of Carbon materials, at the baffles and main chamber walls consisted of Molybdenum materials. In order to study the control of W impurity on EAST with W divertor by an additional radiation losses, the radiative tungsten divertor with Ar impurity seeding for the control of tungsten impurity is studied by SOLPS-ITER simulation. The computational region includes the core, SOL and private flux region. The real geometry and material of plasma facing components (PFCs) are taken into account. The plasma species in the simulation include D 0 , D + 1 , e - , He 0 - He + 2 , C 0 - C + 6 , Mo 0 - Mo + 42 , W 0 - W + 74 and Ar 0 - Ar + 18 . The simulation results show that with the increase of Ar seeding rate at the dome the plasma temperature at the target plates and the concentration of W impurity in the plasma decrease. The simulation plays an important role in the study of radiative tungsten divertor towards the higher heating power and steady state operation.

A tangential fast visible camera has been set up in EAST tokamak for the study of edge MHD instabilities such as ELM. To determine the 3-D information from CCD images, Tsai's two-stage technique was utilized to calibrate the high-speed camera imaging system for ELM study. By applying tiles of the passive stabilizers in the tokamak device as the calibration pattern, transformation parameters for transforming from a 3-D world coordinate system to a 2-D image coordinate system were obtained, including the rotation matrix, the translation vector, the focal length and the lens distortion. The calibration errors were estimated and the results indicate the reliability of the method used for the camera imaging system. Through the calibration, some information about ELM filaments, such as positions and velocities were obtained from images of H-mode CCD videos.

This paper discusses a control system for the Experimental Advanced Superconducting Tokamak (EAST) that suppresses neoclassical tearing modes (NTMs) (3/2 and 2/1) by applying electron cyclotron (EC...

To study the dynamic characteristics of filamentary structure during the edge-localized mode (ELM) burst on the experimental advanced superconducting tokamak, the BOUT++ six-field two-fluid model is used to simulate the temporal and spatial evolutions of the ELM filaments. In the peak gradient region at the outer mid-plane, the radial speed of filaments is changed frequently and varied from 0.11 km/s to 0.79 km/s, and the poloidal speed oscillates in the range of 0.61–0.85 km/s. The width of the ELM filaments is increased from 7.76 mm to 26.7 mm and grows rapidly to 41.99 mm from t = 0.2717 ms to 0.2831 ms after the start of the burst. After that, the width decreases gradually and stabilizes at 25–30 mm when it comes to the saturation phase. Further research indicates that the filament propagation speed is increased during their outward movement. The width gradually decreases when the filaments move outward, and it is proportional to electron temperature (Te) perturbation. All the results are consistent with the experimental data presented by Chen et al., IEEE Trans. Plasma Sci. 47, 799 (2019).

Newly developed large-area pixelated two-dimensional detector and two-crystal assemblies were deployed for the first time on tokamaks to enable time-resolved Bragg-diffracted x-ray imaging with good framing rate and water-cooling capabilities for in-vacuum long-pulse operations. High-quality helium-like (He-like) and hydrogen-like (H-like) argon spectra have been observed simultaneously for the first time on a single detector for a wide range of plasma parameters to infer both ion temperature and rotation profiles and support studies on spontaneous rotation, impurity transport, and RF physics. Since tokamak plasmas rotate in both the poloidal (θ) and toroidal (ϕ) directions, a reliable wavelength calibration is needed to account for the correct Doppler shift as well as to compute the spectrometer's instrumental function. Lyα lines emitted from Cd x-ray tubes are proposed to be used as "markers" to provide an in situ calibration of the EAST's X-ray imaging crystal spectrometer systems measuring He- and H-like argon spectra. The first lab test indicated that the X-ray tube can excite strong Lyα lines at 15 kV voltage and 1 mA current when the crystal is shined for 10 min. Other indirect calibration methods using locked-mode discharge scenarios were also studied as complementary methods.

The plasma current profile modification in the edge region is observed in LHCD (lower hybrid current drive) plasmas for the first time in EAST, correlated with an increased level of PI (parametric instability) activities with the decrease in LH (lower hybrid) source frequency or the increase in plasma density. With the increase in the PI strength, an increase in the edge current is observed despite of a reduction in current drive capability. In these plasmas, strike-point splitting behaviors are also observed, implying the modification of the magnetic geometry in the scrape-off-layer. Results show the effects of LH frequency and plasma density on power dissipation in the edge region, suggesting a role of anomalous wave power damping on the modification of current profile at the plasma edge. These novel results are significant in that they give insight for the first time into how nonlinear wave-plasma interactions such as PI may directly impact the edge current profile, the control of which is critical in order to achieve optimized modes of operation in a steady state fusion reactor. © 2018 IAEA, Vienna.

Stationary long pulse plasma of high electron temperature was produced on EAST for the first time through an integrated control of plasma shape, divertor heat flux, particle exhaust, wall conditioning, impurity management, and the coupling of multiple heating and current drive power. A discharge with a lower single null divertor configuration was maintained for 103 s at a plasma current of 0.4 MA, q95 ≈7.0, a peak electron temperature of >4.5 keV, and a central density ne(0)~2.5×1019 m−3. The plasma current was nearly non-inductive (Vloop

A laser based method combined with spectroscopy, such as laser-induced breakdown spectroscopy (LIBS) and laser-induced ablation spectroscopy (LIAS), is a promising technology for plasma-wall interaction studies. In this work, we report the development of in situ laser-based diagnostics (LIBS and LIAS) for the assessment of static and dynamic fuel retention on the first wall without removing the tiles between and during plasma discharges in the Experimental Advanced Superconducting Tokamak (EAST). The fuel retention on the first wall was measured after different wall conditioning methods and daily plasma discharges by in situ LIBS. The result indicates that the LIBS can be a useful tool to predict the wall condition in EAST. With the successful commissioning of a refined timing system for LIAS, an in situ approach to investigate fuel retention is proposed.

A microwave reflectometry system operating in the V-band frequency with extraordinary mode polarization has been developed on the EAST tokamak. The reflectometry system, using a voltage-controlled oscillator (VCO) source driven by an arbitrary waveform generator with high temporal resolution, can operate for the density profile measurement. The result of the bench test shows that the output frequency of the VCO has a linear dependence on time. The dispersion of reflectometry system is determined and reported in this paper. The evolution of a pedestal density profile during the L-H transition is observed by the reflectometry in H-mode discharges on EAST tokamak. A frequency synthesizer is used to replace the VCO as microwave source for density fluctuation measurements. The level of density fluctuation in the pedestal shows an abrupt decrease when the plasma enters into H-mode. A coherent mode with a frequency of about 100 kHz is observed and the mode frequency decreases gradually as the pedestal evolves.

A series of discharges achieving high- β N ( > 1.5 ) H-mode phases were obtained on EAST with both LHW and NBI as the heating power, a tungsten divertor, an unfavorable ▽ B × B direction configuration for the L-H transition and a lower toroidal field ( B T ∼ 1.6 T ). The experimental power threshold with the unfavorable direction configuration appeared to be higher than previous results with favorable directions in 2014 and 2015, although the tungsten divertor configurations were believed to benefit the L-H transition compared to a graphite divertor. The proportions of LHW and NBI power modulated the ratio of electron temperature to ion temperature, accompanied with an optimal ratio that caused a minimum P l o s s . The WCM might have prolonged the dwell time. The power threshold increased with increasing L ∥ and lower OLL lead to lower power threshold. Better energy confinement was observed during the high- β N H-mode phases.

The reflectometry diagnostics on experimental advanced superconducting tokamak (EAST) are composed of Q-band (32–56 GHz), V-band (48–76 GHz) and W-band (72–110 GHz) microwave reflectometers with X-mode polarization. The three reflectometers with separate frequency bands were launched independently in previous experiments and now are combined together by applying advanced microwave technology. A quasi-optical (QO) combiner/de-combiner by using frequency selective surfaces (FSSs) has been built to combine the three bands so that these waves (32–110 GHz) can be transmitted to the antenna by using one single oversized waveguide. Two double-ridged horns are used for launcher and receiver. The received waves are decoupled by using a QO de-combiner, the same with the combiner. The designed parameters and laboratory test results of the upgraded system have been presented. The density pedestal evolution during the edge localized modes (ELMs) in EAST plasma have been presented as an example for the application of the updated reflectometers.

By using the specific co-neutral beam injection (co-NBI) and counter-NBI systems on EAST, an alternating flow shear discharge has been obtained to study the impact of the flow shear on ballooning-driven edge localized modes (ELMs) at a fixed high collisionality (). The results reveal that the increased flow shear can significantly mitigate ELMs, or even totally suppress ELMs when the shear is large enough. Our simulations with BOUT support the observations on EAST, and further indicate that the increased can both reduce the linear growth rate of the ballooning mode and shorten its growth time (phase coherence time, PCT). The enhanced nonlinear interactions shorten the PCT of the ballooning mode, as validated by the bispectrum study on EAST. All those studies suggest a new way to control ELMs.

The radiation distribution and radiated energy loss during ELM (Edge Localized Mode) are preliminarily investigated in the Experimental Advanced Superconducting Tokamak (EAST). The radiated power is measured by fast bolometer which is based on compact, multichannel photodiodes array with fast response time of ∼μs. Most of the ELMs are type-III ELMs for the present H-mode discharges. It is found that only a small part of the ELM-induced energy loss is dissipated by radiation. Radiation distribution in divertor region during ELM is compared with that of between ELMs. The radiation profiles are always peaked near the X-point. Power balance is also presented for the ELMy H-mode discharge.

Experiments with a small amount of Ar puffed from the lower outer target into L-mode plasma in double null (DN), lower single null (LSN), and upper single null (USN) configurations studied the effects of magnetic configurations and the seeded location relative to the X-point on the performance of Ar seeding. With Ar seeded, the peak heat flux and the integrated power load onto the target decreased most significantly in the LSN configuration, although the increase of the line integrated radiation in the main plasma was the lowest in this case and the highest in the USN configuration. The radiation fraction in the divertor rose from about 10% to 23% of the heating power in LSN, and from around 15% to 22% in DN, while it almost did not change in the USN configuration with Ar seeding. The line emission of Ar XVI increased following the puff and then started to degrade near the end of LSN and USN phases. No such degradation was seen in the DN phase, meaning that some Ar dwelt in the core till the LSN phase beg...

An ICRF system with power up to 6.0 MW and a LHCD system up to 4MW have been applied for heating and current drive experiments on EAST. Intensive lithium wall coating was intensively used to reduce particle recycling and Hydrogen concentration in Deuterium plasma, which is needed for effective ICRF and LHCD power absorption in high density plasmas. Significant progress has been made with ICRF heating and LHW current drive for realizing the H-mode plasma operation in EAST. In 2010, H-mode was generated and sustained by LHCD alone, where lithium coating and gas puffing launcher mouth were applied to improve the LHCD power coupling and penetration into the core plasmas at high density of H-modes. During the last two experimental campaigns, ICRF Heating experiments were carried out at the fixed frequency of 27MHz, achieving effective ions and electrons heating with the H Minority Heating (H-MH) mode, where electrons are predominantly heated by collisions with high energy minority ions. The H-MH mode gave the best plasma performance, and realized H-mode alone in 2012. Combination of ICRF and LHW power injection generated the H-mode plasmas with various ELMy characteristics. The first successful application of the ICRF Heating in the D (He3) plasma was also achieved. The progress on ICRF heating, LHCD experiments and their application in achieving H-mode operation from last two years will be discussed in this report.

Oxidation treatment is one of the most promising methods for amorphous tritiated carbon layers removal, which is important for tritium removal for ITER. A series of experiments of oxidation wall conditioning, including thermo-oxidation, glow discharge (O-GDC) and radio frequency wave associated oxidation (O-ICR), were carried out in a limiter machine – HT-7 with 20% carbon walls, and in an ITER-relevant full superconducting tokamak – EAST with metal walls or full carbon walls for divertor configuration. This paper will give a review of the oxidation wall conditionings on HT-7 and EAST.

Hot spots induced by lower hybrid wave in experimental advanced superconducting tokamak tokamak have caused high performance experiment disruption and serious damages to the guard limiters. Experimental and theoretical analyses have been carried out to study its physical mechanism. Plasma density scan experiments indicate that the wall temperature within the hot spots enhanced by a factor of 5 and increases with the plasma density near the antenna. A lower hybrid current drive (LHCD)-only density climb experiment shows that the carbon impurity decreases to a minimum value at certain plasma density and then increases with the line averaged plasma density. A model has been developed to explain the mechanism of sputtering of graphite tiles due to hot spots as the plasma density near the LHCD antenna and the time increases. A theoretical scaling of the heat flux driven by LHCD is also presented and is consistent with the experimental scaling in the Tore Supra tokamak. The simulation results show that the total sputtering flux density has a minimum at a certain plasma density and gradually increases as the plasma density increases or decreases away from the minimum value, and the increase in parallel heat flux near the antenna would enhance the sputtering flux density. The sputtering flux density trend is qualitatively consistent with the density scan experiments. The simulated temporal evolution of sputtered flux implies that the chemical sputtering could be a candidate for the carbon impurity explosion.

A detailed analysis of the synchrotron radiation intensity and energy of runaway electrons is presented for the Experimental Advanced Superconducting Tokamak (EAST). In order to make the energy of the calculated runaway electrons more accurate, we take the Shafranov shift into account. The results of the analysis show that the synchrotron radiation intensity and energy of runaway electrons did not reach the maximum at the same time. The energy of the runaway electrons reached the maximum first, and then the synchrotron radiation intensity of the runaway electrons reached the maximum. We also analyze the runaway electrons density, and find that the density of runaway electrons continuously increased. For this reason, although the energy of the runaway electrons dropped but the synchrotron radiation intensity of the runaway electrons would continue rising for a while.

The Experimental Advanced Superconducting Tokamak (EAST) was the first shaped tokamak of mega-ampere scale to achieve plasma utilizing a fully superconducting poloidal field coil system, and it is ...

Plasma equilibrium reconstruction plays an important role in the tokamak plasma research. With a high temporal and spatial resolution, the POlarimeter-INTerferometer (POINT) system on EAST has provided effective measurements for 102s H-mode operation. Based on internal Faraday rotation measurements provided by the POINT system, the equilibrium reconstruction with a more accurate core current profile constraint has been demonstrated successfully on EAST. Combining other experimental diagnostics and external magnetic fields measurement, the kinetic equilibrium has also been reconstructed on EAST. Take the pressure and edge current information from kinetic EFIT into the equilibrium reconstruction with Faraday rotation constraint, the new equilibrium reconstruction not only provides a more accurate internal current profile but also contains edge current and pressure information. One time slice result using new kinetic equilibrium reconstruction with POINT data constraints is demonstrated in this paper and the result shows there is a reversed shear of q profile and the pressure profile is also contained. The new improved equilibrium reconstruction is greatly helpful to the future theoretical analysis.

Plasma was first obtained successfully on the EAST Tokamak in the autumn of 2006. This was an ohmically heated plasma with rf-assisted pre-ionization for plasma startup. Many diagnostics systems were set up to satisfy the requirement for the description of basic plasma performance. Three hard X-ray detectors have been installed to monitor the runaway electrons with a tangential view into the plasma on the equatorial plane. A BGO (Bi4Ge3O12) scintillator installed in the forward direction was used to detect the hard X-ray emission in the energy range 0.5–7 MeV while two CdTe detectors in the backward and forward directions are used to detect the hard X-ray emission in the energy range 0.3-1.2 MeV. Phenomena observed with these detectors in different discharge conditions are presented here.

The first plasma was obtained in the EAST on September 26th, 2006. Single-null (SN) and double-null (DN) diverted plasmas were achieved successfully in the EAST tokamak on January 22nd, 2007. The employed plasma diagnostics for first plasma study of EAST are as follows: a vertical one-channel far-infrared (FIR) hydrogen cyanide (HCN) laser interferometer for measuring the line average density, a 10-channel soft X-ray array for intensity measurement, a 16-channel heterodyne Electron Cyclotron Emission (ECE) for measuring the electron temperature profile, a 8-channel XUV bolometer array to measure plasma radiation losses, a 3-channel hard X-ray array for intensity measurement, an electromagnetic measurement system, a 35-channel H α radiation array, 20 probes for divertor plasma, a one-channel visible bremsstrahlung emission, an impurity optical spectrum measurement system and two optical spectroscopic multi-channel analyzers (OMA). The first experimental results of diagnostic systems are summarized in this Letter.

The first plasma discharges were successfully achieved on the experimental advanced superconducting tokamak (EAST) in 2006. The sawteeth behaviours were observed by means of soft x-ray diagnostics and ECE signals in the EAST. The displacement and radius of the q = 1 surface was studied and compared with the result of equilibrium calculation. The density sawtooth oscillation was also observed by the HCN laser interferometer diagnostics. The structure of the EAST operational region was studied in detail. Plasma performance was obviously improved by the boronization and wall conditioning. It was observed that lower qa and a wider stable operating region is extended by the GDC boronization.

Using a tangentially viewing x-ray imaging crystal spectrometer, substantial co-current rotation driven by lower hybrid current drive（LHCD） at 4.6 GHz is observed on EAST tokamak. This study presents plasma rotation behaviors with 4.6 GHz LHCD injection. Typically, the 10-20 km/s co-current rotation change and the transport of rotation velocity from edge to core are observed. The relationship between plasma parameters and rotation is also investigated, indicating that rotation decreases with increasing internal inductance（li） and increases with increasing safety factor（q0）. Hysteresis between rotation and Te plasma stored energy is observed, suggesting different response times between the electron heating and rotation acceleration by LHCD. A comparison between the rotations driven by 4.6 G LHCD and 2.45 G LHCD on EAST is also presented, in which higher frequency LHCD could induce more rotation changes.

In this article, we present the design of a four-channel Ultra-Fast Charge eXchange Recombination Spectroscopy (UF-CXRS) diagnostic on the EAST tokamak. This diagnostic is based on the active charge exchange to measure ion temperature with a time resolution at the order of 1 μ s and a spatial resolution of 1 cm. The design of the main components is carefully discussed, including the dichroic mirror light path, the fiber, the spectrometer and the customized lens. The theoretical estimation of the emission photon flux has been carried out, proving the feasibility of the whole design.

The internal transport barrier (ITB) has been obtained in ELMy H-mode plasmas by neutron beam injection and lower hybrid wave heating on the Experimental Advanced Superconducting Tokamak (EAST). The ITB structure has been observed in profiles of ion temperature, electron temperature, and electron density within ρ < 0.5. It was also observed that the ITB formation is stepwise. Due to the ITB formation, the confinement quality H 98y2 increases from 1 to 1.1 and the normalized beta, β N, increases from 1.5 to near 2. The fishbone activity observed during the ITB phase suggests the central safety factor q(0) ~ 1. Transport coefficients are calculated by particle balance and power balance analysis, showing an obvious reduction after the ITB formation.

The first EAST (Experimental Advanced Superconducting Tokamak) plasma ignited in 2006 with non-actively cooled steel plates as the plasma-facing materials and components (PFMCs) which were then upgraded into full graphite tiles bolted onto water-cooled copper heat sinks in 2008. The first wall was changed further into molybdenum alloy in 2012, while keeping the graphite for both the upper and lower divertors. With the rapid increase in heating and current driving power in EAST, the W/Cu divertor project was launched around the end of 2012, aiming at achieving actively cooled full W/Cu-PFCs for the upper divertor, with heat removal capability up to 10 MW m−2. The W/Cu upper divertor was finished in the spring of 2014, consisting of 80 cassette bodies toroidally assembled. Commissioning of the EAST upper W/Cu divertor in 2014 was unsatisfactory and then several practical measures were implemented to improve the design, welding quality and reliability, which helped us achieve successful commissioning in the 2015 Spring Campaign. In collaboration with the IO and CEA teams, we have demonstrated our technological capability to remove heat loads of 5000 cycles at 10 MW m−2 and 1000 cycles at 20 MW m−2 for the small scale monoblock mockups, and surprisingly over 300 cycles at 20 MW m−2 for the flat-tile ones. The experience and lessons we learned from batch production and commissioning are undoubtedly valuable for ITER (International Thermonuclear Experimental Reactor) engineering validation and tungsten-related plasma physics.

Experimental Advanced Superconducting Tokamak (EAST) aims to demonstrate steady-state advanced high-performance H-mode plasmas with an ITER-like configuration, plasma control and heating schemes. The plasma-facing components in EAST are actively cooled, providing good conditions for researching long-pulse and high-energy discharges. A long-pulse high-performance plasma discharge (#59892 discharge) of up to 103 s with a core electron temperature of up to 4.5 keV was sustained with an injected energy exceeding 0.22 GJ in the 2015–2016 experimental campaign. A calorimetric measurement utilizing the temperature increment of cooling water is carried out to calculate the heat load on the strike point region of the lower divertor during long-pulse discharges in EAST. For the long-pulse and high-energy discharges, the comparison of the measurement results for the heat load measured by divertor Langmuir probes and the calorimetry diagnostic indicates that most of the heat load is delivered to the divertor panels as plasma, not radiation, and charge exchange neutrals. The ratio of the heat load on the strike point region of the lower divertor to the total injected energy is on average 42.5% per discharge with the lower single null divertor configuration. If the radiated energy loss measured by the fast bolometer diagnostic is taken into consideration, the ratio is found to be 61.6%. The experimental results and the analysis of the physics involved in these discharges are reported and discussed.

ITER and future superconducting fusion machines need efficient wall conditioning techniques for routine operation in between shots in the presence of permanent high magnetic field for wall cleaning, surface isotope exchange and to control the in-vessel long term tritium retention. Ion Cyclotron Wall Conditioning (ICWC) based on the ICRF discharge is fully compatible and needs the presence of the magnetic field. The present paper focuses on the principal aspects of the ICWC discharge performance in large-size fusion machines: (i) neutral gas RF breakdown with conventional ICRF heating antennas, (ii) antenna coupling with low density (similar to 10(17) m(-3)) RF plasmas and (iii) ICWC scenarios with improved RF plasma homogeneity in the radial and poloidal directions. All these factors were identified as crucial to achieve an enhanced conditioning effect (e.g. removal rates of selected "marker" masses). All the observed effects are analyzed in terms of RF plasma wave excitation/absorption and compared with the predictions from I-D RF full wave and 0-D RF plasma codes. Numerical modeling and empirical extrapolation from the existing machines give good evidence for the feasibility of using ICWC in ITER with the main ICRF antenna.

EAST has been equipped with two high power lower hybrid current drive (LHCD) systems with operating frequencies of 2.45 GHz and 4.6 GHz. Comparative LHCD experiments with the two different frequencies were performed in the same conditions of plasma for the first time. It was found that current drive (CD) efficiency and plasma heating effect are much better for 4.6 GHz LH waves than for the one with 2.45 GHz. High confinement mode (H-mode) discharges with 4.6 GHz LHCD as the sole auxiliary heating source have been obtained in EAST and the confinement is higher with respect to that produced previously by 2.45 GHz. A combination of ray-tracing and Fokker-Planck calculations by using the C3PO/LUKE codes was performed in order to explain the different experimental observations between the two waves. In addition, the frequency spectral broadening of the two LH wave operating frequencies was surveyed by using a radio frequency probe. Published by AIP Publishing.

In toroidal magnetic fusion devices, fast-ion D-alpha diagnostic (FIDA) is a powerful method to study the fast-ion feature. The fast-ion characteristics can be inferred from the Doppler shifted spectrum of Dα light according to charge exchange recombination process between fast ions and probe beam. Since conceptual design presented in the last HTPD conference, significant progress has been made to apply FIDA systems on the Experimental Advanced Superconducting Tokamak (EAST). Both co-current and counter-current neutral beam injectors are available, and each can deliver 2-4 MW beam power with 50-80 keV beam energy. Presently, two sets of high throughput spectrometer systems have been installed on EAST, allowing to capture passing and trapped fast-ion characteristics simultaneously, using Kaiser HoloSpec transmission grating spectrometer and Bunkoukeiki FLP-200 volume phase holographic spectrometer coupled with Princeton Instruments ProEM 1024B eXcelon and Andor DU-888 iXon3 1024 CCD camera, respectively. This paper will present the details of the hardware descriptions and experimental spectrum.

The pellet-induced snake oscillation was observed by soft x-ray (SXR) diagnostic in EAST for the first time after a fueling-sized pellet penetrated the q = 1 surface. The snake phenomenon has a long lifetime with a helicity of m = 1 and n = 1. Basic behaviors of the snake, including the triggering condition, interaction with the sawtooth and snake rotation frequency, were discussed in detail by multiple core diagnostics. The snake location was also analyzed through observation of the vertical SXR arrays and raw SXR brightness profiles. It is clear that the snake resided in a broad region between the magnetic axis and the q = 1 surface derived from equilibrium reconstruction. This investigation is beneficial for the understanding of the snake formation for EAST and future devices, like ITER and DEMO.

Ion cyclotron radio frequency (ICRF) wave heating is a primary method to heat ions in the Experimental Advanced Superconducting Tokamak (EAST). Through neutron diagnostics, effective ion heating was observed in hydrogenminority heating (MH) scenarios. At present, investigation of deuterium–deuterium (DD) fusion neutrons is mostly based on time-resolved flux monitor and spectrometer measurements. When the ICRF was applied, the neutron intensity became one order higher. The H/H + D ratio was in the range of 5–10%, corresponding to the hydrogen MH dominated scenario, and a strong high energy tail was not displayed on the neutron spectrum that was measured by a liquid scintillator. Moreover, ion temperature in the plasma center (T i) was inversely calculated by the use of neutron source strength (S n) and the plasma density based on classical fusion reaction equations. This result indicates that T i increases by approximately 30% in L-mode plasma, and by more than 50% in H-mode plasma during ICRF heating, which shows good agreement with x-ray crystal spectrometer (XCS) diagnostics. Finally, the DD neutron source strength scaling law, with regard to plasma current (I P) and ICRF coupling power (P RF) on the typical minority heating condition, was obtained by statistical analysis.

Increasing the density in a tokamak is limited by the so-called density limit, which is generally performed as an appearance of disruption causing loss of plasma confinement, or a degradation of high confinement mode which could further lead to a H → L transition. The L-mode and H-mode density limit has been investigated in EAST tokamak. Experimental results suggest that density limits could be triggered by either edge cooling or excessive central radiation. The L-mode density limit disruption is generally triggered by edge cooling, which leads to the current profile shrinkage and then destabilizes a 2/1 tearing mode, ultimately resulting in a disruption. The L-mode density limit scaling agrees well with the Greenwald limit in EAST. The observed H-mode density limit in EAST is an operational-space limit with a value of . High density H-mode heated by neutral beam injection (NBI) and lower hybrid current drive (LHCD) are analyzed, respectively. The constancy of the edge density gradients in H-mode indicates a critical limit caused perhaps by e.g. ballooning induced transport. The maximum density is accessed at the H → L transition which is generally caused by the excessive core radiation due to high Z impurities (Fe, Cu). Operating at a high density () is favorable for suppressing the beam shine through NBI. High density H-mode up to could be sustained by 2 MW 4.6 GHz LHCD alone, and its current drive efficiency is studied. Statistics show that good control of impurities and recycling facilitate high density operation. With careful control of these factors, high density up to 0.93 stable H-mode operation was carried out heated by 1.7 MW LHCD and 1.9 MW ion cyclotron resonance heating with supersonic molecular beam injection fueling.

The lower hybrid wave current drive (LHCD) is one of the efficient methods of driving the non-inductive current required for Tokamak operating in steady-state. Residual loop voltage exists in Tokamak when the non-inductive current is not fully driven. Residual loop voltage also accelerates the fast electrons generated by the lower hybrid wave (LHW), which can drive extra current and combine with the current driven by the LHW. It is generally difficult to separate these two different components of driven current in the experiment. In this paper, the currents driven by LHCD and residual loop voltage are separated directly by solving the Fokker–Plank equation numerically. The fraction of the current driven by residual loop voltage compared to the current driven by LHW is evaluated on the experimental advanced superconducting tokamak (EAST). The current driven by residual loop voltage is several percent of the currents driven by the LHCD when the residual loop voltage is small, but it increases with the residual loop voltage up to 25% when the residual loop voltage is about 2 V. The hot electrical conductivity is deduced from the net current driven by the residual loop voltage. Its distribution profile is related to the fast electron distribution driven by LHW.