Increasing Plasma Current Research Articles

EuroPED-NN: uncertainty aware surrogate model

This work successfully generates an uncertainty-aware surrogate model of the EuroPED plasma pedestal model using the Bayesian neural network with noise contrastive prior (BNN-NCP) technique. This model is trained using data from the JET-ILW pedestal database and subsequent model evaluations, conforming to EuroPED-NN. The BNN-NCP technique has been proven to be a suitable method for generating uncertainty-aware surrogate models. It matches the output results of a regular neural network while providing confidence estimates for predictions as uncertainties. Additionally, it highlights out-of-distribution regions using surrogate model uncertainties. This provides critical insights into model robustness and reliability. EuroPED-NN has been physically validated, first, analyzing electron density ne(ψpol=0.94) with respect to increasing plasma current, Ip , and second, validating the Δ−βp,ped relation associated with the EuroPED model. This affirms the robustness of the underlying physics learned by the surrogate model. On top of that, the method was used to develop a EuroPED-like model fed with experimental data, i.e. an uncertainty aware experimental model, which is functional in JET database. Both models have been also tested in ∼50 AUG shots.

Hydrogen removal by electron cyclotron wall conditioning with neon gas and its impact of tokamak plasma start-up on the QUEST spherical tokamak

Electron cyclotron wall conditioning with neon gas (Ne-ECWC) has been performed on the normal conducting spherial tokamak QUEST with metal walls under a trapped particle configuration with O-mode EC waves including X-mode polarization with a frequency of 8.2 GHz and an injection power of 16 kW. The Ne-ECWC removes hydrogen from the wall with small neon retention. The Ne-ECWC decreases hydrogen recycling at the following tokamak discharges, contributing to an improvement of the following tokamak plasma start-up: the plasma current increases and the start-up timing of the plasma current shifts forward. However, defects such as voids and bubbles are formed on tungsten surface exposed to the Ne-ECWC plasma.

Experimental and simulation studies on ELM characteristics under a plasma current ramping up discharge on EAST

Previous experimental results show that the poloidal mode spacing of the filamentary structures increases and the dominant toroidal mode number decreases in the edgelocalized mode (ELM) rising phase with increasing plasma current. In addition, the experimental results in this paper show that the energy loss ratio of the pedestal (ΔW/Wped) decreases as the edge safety factor (q95) increases. The BOUT++ three-field two-fluid model can reproduce the experimental results and provide a possible explanation mechanism. The pedestal density plays an important role in the characteristics of filamentary structures as the current ramps up. On the one hand, the resistivity related to the pedestal density drives the instability of the peeling–ballooning mode, and the resistive effect is stronger in the high current case, making the dominant toroidal mode number lower and the corresponding poloidal mode spacing wider in the high current case. A low q95 corresponds to a high pedestal collision rate and a high pedestal energy loss ratio. On the other hand, the ELM crash process is dominated by resistivity, so the ratio of pedestal energy loss caused by ELM is not inversely proportional to the pedestal collision rate.

The timescale of thermal quench during disruptions in EAST

Since 2015, the timescale of thermal quench (TQ) during disruptions on the EAST tokamak has been examined using electron cyclotron emission and soft x-ray diagnostic systems. The database includes both major disruptions (MDs) and hot vertical displacement events (VDEs), where the TQ duration of the former is within 56–788 μs, and the latter is approximately within 100–3000 μs. In particular, for MDs, the lower bound of TQ duration, indicating the minimum durations at different current plateaus, decreases as the plasma current increases. This decrease is due to the connection length shortening and the plasma temperature increasing. For MDs, two typical TQ processes, single-stage TQ and double-stage TQ, are characterized by different magnetic perturbations. In single-stage TQ, a fast-loss stage is triggered by magnetic perturbation exceeding 4.3 × 10−3 T with a fast growth rate of 1.5 × 10−2 μs−1. In contrast, fast quench is triggered by a slightly smaller magnetic perturbation of 3.6 × 10−3 T in double-stage TQ, and the growth rate 5.3 × 10−3 μs−1 is an order of magnitude smaller than single-stage TQ. For hot VDEs, the plasma temperature collapses step by step from the edge to the core, and every progressive collapse corresponds to a magnetic perturbation, whose growth rate is approximately equal to or less than double-stage TQ. The whole process of TQ energy release can be divided into the transport in a stochastic field within the separatrix and in the scrape-off layer and, according to the typical parameters of EAST, TQ duration in MDs is roughly estimated to be 245 μs by an approximate formula, which is consistent with the experimental results.

Study on divertor detachment and pedestal characteristics in the DIII-D upper closed divertor

Experiments performed in DIII-D demonstrate that higher plasma current and heating power combined with impurity seeding facilitate the achievement of divertor detachment with a higher pedestal pressure and higher plasma performance in H-mode plasmas with a baffled closed divertor compared with an open divertor. Dedicated experiments were carried out to study the impact of power, plasma current and impurity seeding on divertor detachment with ion directed into the divertor favorable for the L–H transition. With a factor of three variation in heating power and with only D2 puffing, no significant difference in the separatrix density at detachment onset was found. The higher heating power leads to higher impurity concentration and wider scrape-off layer (SOL) width, and reduces the detachment onset density to one similar to that in lower-power plasmas. Higher current requires higher pedestal and line-averaged densities to achieve divertor detachment; however, the increase in separatrix density at increasing plasma current is found to be less pronounced. Initial calculations found that both power scan and plasma current scan datasets are qualitatively consistent with theory after considering the change in impurity concentration and heat flux width. This also motivates the future extensive study of transport and divertor impurity behavior in order to have a quantitative comparison between experiment and theory. Compared with an open divertor, a closed divertor facilitates detachment onset at ∼40% lower line-averaged plasma density. Additional N2 seeding facilitates the achievement of detachment at a lower separatrix density and thus a higher pedestal temperature, which is beneficial for advanced tokamak scenarios. Higher heating power requires a higher N2 puffing rate to achieve the same degree of detachment, while a higher N2 puffing rate leads to lower detachment onset line-averaged density, both of which agree with theory. In contrast to the narrower pedestal in an open divertor approaching detachment, the pedestal density width in a closed divertor increases with density. The density gradient increases with line-averaged density at higher plasma current, but remains nearly unchanged at lower plasma current. In particular, compared with discharges with low power, at high heating power the pedestal density gradient is much weaker, while the SOL density is significantly higher and wider. At the same plasma current, both pedestal pressure gradient and temperature gradient decrease linearly with the line-averaged density but remain similar across different heating powers. Even with different plasma current and heating power, the normalized pressure gradient remains identical. As a result, achievement of divertor detachment with a higher pedestal pressure and higher plasma performance is shown in a closed divertor, which is important for improving core–edge integration as one of the critical issues for future tokamak fusion reactors.

Research on Arc Morphology and Keyhole Behavior of Molten Pool in Magnetically Controlled Plasma-GMAW Welding

In the magnetically controlled Plasma-GMAW welding process, the composite arc forms a keyhole in the workpiece to be welded. In order to explore the effect of process parameters on arc coupling, weld pool and keyhole, and the behavior characteristics of keyhole, the arc behavior and side weld pool information were collected using a welding arc acquisition system and a high-speed camera during bead-on-plate welding. The arc image is processed by pseudo-color enhancement technology, and the collected molten pool information is analyzed by boundary extraction algorithm and coordinate conversion algorithm, and the molten pool boundary and keyhole entrance width are obtained. It is found that the coupling degree of the two arcs increases with the increase in plasma current, GMAW current and magnetic field intensity. With the increase in plasma current, the size of keyhole inlet increases; with the increase of GMAW current, the size of keyhole inlet decreases, and the wave crest increases. With the increase of magnetic field intensity, the intensity of metal oscillation between the two arcs increases, and so does the wave crest.

Current quench and vessel currents characterisation at the COMPASS tokamak

The characterisation of plasma current quench and understanding of its underlying physical processes play a crucial role when designing large fusion devices such as ITER. For the first time, an extensive analysis of the COMPASS tokamak disruption database is presented. A unique set of magnetic diagnostics allows the investigation of local toroidal and poloidal vessel currents, including currents flowing along the open magnetic field lines from the plasma to the vacuum vessel (VV) (i.e., halo currents). Area-normalised current quench times are in agreement with the ITPA 1.67 ms m−2 lower limit. Extremely fast I p quench rates of up to 0.6 MA ms−1 are observed during runaway electron campaigns at COMPASS, which are under the ITPA lower limit. Vertical movement of the plasma column is accelerated by the I p quench during major disruptions. Toroidal vessel currents of around 2% – 4% of the predisruptive plasma current are observed during I p quench. Net poloidal eddy currents are obtained by Mirnov coils and diamagnetic loop, reaching 3% of . Using the Mirnov coils it is shown that the halo current magnitude grows and its poloidal profile broadens with increasing plasma current I p . Geometric features of the VV structure and in-vessel component positions on the poloidal vessel current measurements are discussed.

Wettability control of paper through substitution between the hydroxyl group and carbon elements using argon-carbon plasma treatment

The surface of paper was treated to control wettability using argon-carbon plasma generated by a direct current-pulsed (frequency: 60 kHz) sputtering device with parallel-arrangement magnets. After plasma treatment, the C–C bond percentage in the C 1s region was inversely proportional to the C–OH bond percentage, while the ratio of C=O and O–C=O bonds only changed slightly. The change ratios for the increase in C–C and the decrease in the C–OH bonds were approximately the same, and the change ratio increased from approximately 5%–11% with an increase in the plasma current. The reduction ratio of the contact angle decreased from 0.945 to 0.182 (°/s) with increasing plasma current. These results indicate that the water resistance of the paper surface was enhanced after the simple plasma treatment, and wettability could be easily controlled by changing the substitution ratio between C–OH and C–C elements, achieved by altering the plasma current.

Identification of kink instability in 3D helical flux ropes at VEST

Local helicity injection (LHI) is a non-inductive startup and current drive method via Taylor relaxation for the spherical torus. In achieving Taylor relaxation, it has been suggested that kink instability in 3D helical flux ropes plays an important role. However, the role and occurrence of kink instability during LHI have yet to be validated. Experimentally, determining the kink mode in a flux rope relies on measuring internal information using a probe. However, for LHI, the 3D geometry complicates this measurement process. Here, we propose a new approach for determining the kink modes of 3D helical flux ropes without any internal probe measurements. It is confirmed by this approach that flux ropes exhibit two different kink modes. With increasing plasma current in the flux ropes, a transition from the coherent internal kink mode to the external kink mode is observed. Kink mode properties such as rotating frequency calculated from the kink theory agree well with the magnetic signature driven by the kink mode. During the LHI experiment in the versatile experiment spherical torus, three distinguishable phases are confirmed by the approach, consistent with NIMROD simulation. Before driving the toroidal plasma current, the external kink mode is observed for 3D helical flux ropes. As the toroidal plasma current increases, the external kink mode disappears while generating broadband internal modes instead of coherent internal kink of flux ropes. Decoupling between the toroidal plasma and flux rope results in both decay of toroidal plasma current and re-appearance of the external kink mode in the flux ropes.

Start-Up Studies of GLAST-III Spherical Tokamak in the Presence of Poloidal Field

GLAST-III (glass spherical tokamak) is a small device having an aspect ratio of two (R/a = 2) with major radius R = 20 cm and minor radius a = 10 cm. Experiments are performed to examine the effect of increasing the poloidal field (PF) produced by poloidal coils as well as to study the effect of the inductance of PF coils on the plasma current formation. To investigate the GLAST-III plasma, several diagnostic tools such as Rogowski coil, loop voltage, photodiode, Optical emission spectroscopy, high-speed camera, and Langmuir probe are used. After optimizing the working pressure, the first experiment is performed using toroidal field coils and central solenoid along with radio frequency (RF) pre-ionization source. A plasma current of 1 kA is achieved for 1.2 ms. In later experiments, the effect of the poloidal magnetic field is also included. It is observed that with increasing the PF, the plasma current increases, attain a maximum value (5 kA) and then after a critical value of the applied poloidal magnetic field, the plasma current starts decreasing. Similarly, with the inclusion of inductor in series with the PF coils, the plasma current value is reduced and its pulse duration is increased. Both optical emission and Langmuir probe diagnostics results show a similar trend of the plasma current.

ELM-induced cold pulse propagation in ASDEX Upgrade

In ASDEX Upgrade, the propagation of cold pulses induced by type-I edge localized modes (ELMs) is studied using electron cyclotron emission measurements, in a dataset of plasmas with moderate triangularity. It is found that the edge safety factor or the plasma current are the main determining parameters for the inward penetration of the Te perturbations. With increasing plasma current the ELM penetration is more shallow in spite of the stronger ELMs. Estimates of the heat pulse diffusivity show that the corresponding transport is too large to be representative of the inter-ELM phase. Ergodization of the plasma edge during ELMs is a possible explanation for the observed properties of the cold pulse propagation, which is qualitatively consistent with non-linear magneto-hydro-dynamic simulations.

Up/down impurity density asymmetries in C-Mod plasmas

Brightness profiles of x-ray emission from H-like Ar17+ exhibit a distinct up/down asymmetry under certain operating conditions in C-Mod plasmas, indicating that impurity densities are not constant on flux surfaces with r/a between ∼0.8 and ∼0.95. In L- and I-mode plasmas, there is an x-ray brightness excess, up to a factor of 8, on the side opposite to the ion B × B drift direction. This effect is not observed in H-mode plasmas, presumably due to edge impurity transport being dominated by a strong inward pinch, which is absent in L- and I-mode. The magnitude of the asymmetry in L- and I-mode decreases with increasing plasma current, similar to the observed decrease in radial impurity diffusivity. In I-mode, where the codependence between electron density and temperature can be broken with ICRF heating power, the asymmetry magnitude is found to decrease with increasing density and with increasing edge temperature at fixed density. These measurements exhibit some qualitative features of neo-classical expectations but the observed asymmetry magnitude is much larger than predicted and some scalings with plasma parameters are not seen. The up/down asymmetry appears to be largest when the cross field impurity diffusivity is the highest.

Suppression of fast-ion-driven MHD instabilities by ECH/ECCD on Heliotron J

Experiments of suppressing fast-ion-driven MHD instabilities such as energetic particle modes (EPMs) and global Alfvén eigenmodes (GAEs) have been made using a second harmonic x-mode electron cyclotron heating and current drive (ECCD) in the helical-axis heliotron device, Heliotron J. ECCD experiments show that the GAEs destabilized by fast ions of neutral beam injection with the observed frequency around 140 kHz are fully stabilized, and the EPMs with the observed frequency around 90 kHz are suppressed when the EC-driven plasma current flowing in the counter direction reaches approximately 0.7 kA. The low magnetic shear under the vacuum condition is modified into positive magnetic shear when counter-ECCD is applied, and the amplitude of GAEs and EPMs decreases with an increase in EC-driven plasma current. These results indicate that magnetic shear plays a key role in controlling GAEs as well as EPMs. The comparison of the calculation of shear Alfvén spectra with experimental results shows that the increasing continuum damping rate with an increase in local magnetic shear by EC-driven current is important for both EPMs and GAEs. Moreover, the increase in plasma current leads to the inward movement of GAEs. This effect would also contribute to suppression of GAEs because the continuum damping rate increases more and more toward the core. Steady ECH is also found experimentally to be effective for controlling the amplitude of both GAEs and EPMs. The amplitude of EPMs, and especially for GAEs, decreases with an increase in the ECH power under fixed density conditions.

Statistical analysis of the ion flux to the JET outer wall

Statistical analysis of the ion flux to the JET outer-wall is conducted in outer-wall limiter mounted Langmuir probe (OLP) time-series across a wide range of plasma current and line-averaged density during Ohmically heated horizontal target L-mode plasmas. The mean, μ, and the standard deviation, σ, of the ion-saturation current measured by the OLP show systematic variation with plasma current and density. Both increase as either plasma current decreases and/or density increases. Upon renormalization, achieved by subtraction of μ and rescaling by σ, the probability distribution functions (PDFs) of each signal collapse approximately onto a single curve. The shape of the curve deviates from a distribution in the tail of the PDF and is better described by a log-normal distribution. The invariance in the shape of the PDF, which occurs over approximately four decades of the ordinate, is shown to be the result of a balance between the duration time of the average burst wave-form, and the waiting time between bursts, . This implies that the intermittency parameter, , can be considered constant at the JET outer wall during horizontal target Ohmic L-mode operation. This result may be important both for model validation and prediction.

Calculation of prompt loss and toroidal field ripple loss under neutral beam injection on EAST

Neutral beam injection is a major auxiliary heating method in the EAST experimental campaign. This paper gives detailed calculations of beam loss with different plasma equilibria using the guiding center code ORBIT and NUBEAM/TRANSP. Increasing plasma current can dramatically lower the beam ion prompt loss and ripple loss. Countercurrent beam injection gives a much larger prompt loss fraction than co-injection, and ripple-induced collisionless stochastic diffusion is the dominant loss channel.

Impact of Stationary Direct Current in the Central Solenoidal Coil on Tokamak Plasma Formation by Non-induction Heating

Stationary direct current in the central solenoidal coil (DCCS) of tokamak devices can reduce the non-induction heating energy necessary for tokamak plasma formation. The magnetic field energy in the inner region of the central solenoidal coil (CS region) is expelled during the tokamak plasma formation, because the vertical magnetic field intensity generated by the central solenoidal coil and poloidal field coils is partly cancelled by the increase in the toroidal plasma current. Because this magnetic field energy expelled from the CS region is distributed to the tokamak plasma in accordance with the mutual inductance, this expelled energy can drive the toroidal plasma current inductively. This energy expulsion in the CS region can be enhanced by the DCCS without the modification of the tokamak plasma configuration, when the CS coil current has negligible leakage magnetic field in the plasma area. Because the drive of the toroidal plasma current by non-induction heating can be assisted by this inductive c...

Triton burnup measurements in KSTAR using a neutron activation system.

Measurements of the time-integrated triton burnup for deuterium plasma in Korea Superconducting Tokamak Advanced Research (KSTAR) have been performed following the simultaneous detection of the d-d and d-t neutrons. The d-d neutrons were measured using a 3He proportional counter, fission chamber, and activated indium sample, whereas the d-t neutrons were detected using activated silicon and copper samples. The triton burnup ratio from KSTAR discharges is found to be in the range 0.01%-0.50% depending on the plasma conditions. The measured burnup ratio is compared with the prompt loss fraction of tritons calculated with the Lorentz orbit code and the classical slowing-down time. The burnup ratio is found to increase as plasma current and classical slowing-down time increase.

The appearance and propagation of filaments in the private flux region in Mega Amp Spherical Tokamak

The transport of particles via intermittent filamentary structures in the private flux region (PFR) of plasmas in the MAST tokamak has been investigated using a fast framing camera recording visible light emission from the volume of the lower divertor, as well as Langmuir probes and IR thermography monitoring particle and power fluxes to plasma-facing surfaces in the divertor. The visible camera data suggest that, in the divertor volume, fluctuations in light emission above the X-point are strongest in the scrape-off layer (SOL). Conversely, in the region below the X-point, it is found that these fluctuations are strongest in the PFR of the inner divertor leg. Detailed analysis of the appearance of these filaments in the camera data suggests that they are approximately circular, around 1–2 cm in diameter, but appear more elongated near the divertor target. The most probable toroidal quasi-mode number is between 2 and 3. These filaments eject plasma deeper into the private flux region, sometimes by the production of secondary filaments, moving at a speed of 0.5–1.0 km/s. Probe measurements at the inner divertor target suggest that the fluctuations in the particle flux to the inner target are strongest in the private flux region, and that the amplitude and distribution of these fluctuations are insensitive to the electron density of the core plasma, auxiliary heating and whether the plasma is single-null or double-null. It is found that the e-folding width of the time-average particle flux in the PFR decreases with increasing plasma current, but the fluctuations appear to be unaffected. At the outer divertor target, the fluctuations in particle and power fluxes are strongest in the SOL.

Characterization of plasma current quench during disruption in EAST tokamak**Project supported by the National Magnetic Confinement Fusion Science Program of China (Grant Nos. 2014GB103000 and 2013GB102000) and the National Natural Science Foundation of China (Grant No. 11205199 and 11205192).

A preliminary analysis of plasma current quenching is presented in this paper based on the disruption database. It demonstrates that 26.8% of discharges have been disrupted in the last 2012 campaign, in addition, the plasma disruptive rate grows with the increase of plasma current. The best-fit linear and instantaneous plasma current quench rate is extracted from the recent EAST disruptions, showing that an 80%–30% interval of the maximum plasma current is well fit for the EAST device. The lowest area-normalized current quench time is 3.33 ms/m2 with the estimated plasma electron temperature being 7.3 eV,∼9.5 eV. In the disruption case the maximum eddy current goes up to 400 kA, and a fraction of currents are respectively driven on the upper and lower outer plate with nearly 100 MPa–200 MPa stress in the leg.

대기압 플라즈마의 선택적 도핑 공정에서 온도에 의한 인(Phosphorus)의 확산연구

Kwangwoon University of Department of Chemistry, Seoul, Korea(Received October 8, 2014 ; revised October 22, 2014 ; accepted October 23, 2014)Abstract In this study, we propose the application of doping process technology for atmospheric pressure plasma.The plasma treatment means the wafer is warmed via resistance heating from current paths. These pathsare induced by the surface charge density in the presence of illuminating Argon atmospheric plasmas. Fur-thermore, it is investigated on the high-concentration doping to a selective partial region in P type solarcell wafer. It is identified that diffusion of impurities is related to the wafer temperature. For the fixed plasmatreatment time, plasma currents were set with 40, 70, 120 mA. For the processing time, IR(Infra-Red) imagesare analyzed via a camera dependent on the temperature of the P type wafer. Phosphorus concentrationsare also analyzed through SIMS profiles from doped wafer. According to the analysis for doping process,as applied plasma currents increase, so the doping depth becomes deeper. As the junction depth is deeper,so the surface resistance is to be lowered. In addition, the surface charge density has a tendency inverselyproportional to the initial phosphorus concentration. Overall, when the plasma current increases, then it becomeshigher temperatures in wafer. It is shown that the diffusion of the impurity is critically dependent on thetemperature of wafers.