Plasma Current Ip Research Articles

Experimental scaling of divertor heat and particle flux widths in EAST L-mode plasmas

The scalings of divertor heat and particle flux widths have been systematically performed on the Experimental Advanced Superconducting Tokamak (EAST) in radio-frequency (RF) heated L-mode plasma regime under various divertor configurations. The widths were calculated from the measurements of divertor Langmuir probe (LP) arrays and the reciprocating LP diagnostic on the low-field side. A strong inverse scaling of particle and heat flux widths with plasma current Ip (equivalently the poloidal field Bp) has been demonstrated. The measurements of divertor LPs show that the power decay length exhibits λq,div=4.97Ip−0.94 (λq,div=1.46Bp−1.15), in good agreement with the particle decay length λjs=4.74Ip−1.02 (λjs=1.34Bp−1.21). Similar trend measurements have also been demonstrated by the reciprocating LPs. In addition, the scaling with Ip (Bp) appears to be insensitive of the divertor configurations in EAST. The study of heat and particle flux width scaling in EAST is useful for the extrapolation to the future tokamaks such as ITER.

Progress toward commissioning and plasma operation in NSTX-U

The National Spherical Torus Experiment-Upgrade (NSTX-U) is the most powerful spherical torus facility at PPPL, Princeton USA. The major mission of NSTX-U is to develop the physics basis for an ST-based Fusion Nuclear Science Facility (FNSF). The ST-based FNSF has the promise of achieving the high neutron fluence needed for reactor component testing with relatively modest tritium consumption. At the same time, the unique operating regimes of NSTX-U can contribute to several important issues in the physics of burning plasmas to optimize the performance of ITER. NSTX-U further aims to determine the attractiveness of the compact ST for addressing key research needs on the path toward a fusion demonstration power plant (DEMO). The upgrade will nearly double the toroidal magnetic field BT to 1 T at a major radius of R0 = 0.93 m, plasma current Ip to 2 MA and neutral beam injection (NBI) heating power to 14 MW. The anticipated plasma performance enhancement is a quadrupling of the plasma stored energy and near doubling of the plasma confinement time, which would result in a 5–10 fold increase in the fusion performance parameter nτ T. A much more tangential 2nd NBI system, with 2–3 times higher current drive efficiency compared to the 1st NBI system, is installed to attain the 100% non-inductive operation needed for a compact FNSF design. With higher fields and heating powers, the NSTX-U plasma collisionality will be reduced by a factor of 3–6 to help explore the favourable trend in transport towards the low collisionality FNSF regime. The NSTX-U first plasma is planned for the Summer of 2015, at which time the transition to plasma operations will occur.

Microwave experiments on Prairie View Rotamak

A 6 kW/2.45 GHz microwave system has been added on Prairie View Rotamak, and a series of experiments with microwave heating in both O-mode and X-mode configurations have been performed. Effective ionization of hydrogen in the two configurations is observed when filling pressure of the hydrogen gas is under pf=0.1 Pa. Clear oscillations in plasma current Ip and magnetic field BR are excited when microwaves are injected into plasma in the X-mode configuration. The higher the injected microwave power, the sooner the emergence of the magnetic oscillations in BR, which implies the microwave may have decreased the elongation of the plasma. In the experiments, the efficiency of the current drive mechanism due to the injected microwave is about 0.2 kA/kW.

ITER-like current ramps in JET with ILW: experiments, modelling and consequences for ITER

Since the ITER-like wall in JET (JET-ILW) came into operation, dedicated ITER-like plasma current (Ip) ramp-up (RU) and ramp-down (RD) experiments have been performed and matched to similar discharges with the carbon wall (JET-C). The experiments show that access to H-mode early in the Ip RU phase and maintaining H-mode in the Ip RD as long as possible are instrumental to achieve low internal plasma inductance (li) and to minimize flux consumption. In JET-ILW, at a given current rise rate similar variations in li (0.7–0.9) are obtained as in JET-C. In most discharges no strong W accumulation is observed. However, in some low density cases during the early phase of the Ip strong core radiation due to W influx led to hollow electron temperature (Te) profiles. In JET-ILW Zeff is significantly lower than in JET-C. W significantly disturbs the discharge evolution when the W concentration approaches 10−4; this threshold is confirmed by predictive transport modelling using the CRONOS code. Ip RD experiments in JET-ILW confirm the result of JET-C that sustained H-mode and elongation reduction are both instrumental in controlling li.

Co- and counter-current rotation in Tore Supra lower hybrid current drive plasmas

Observations of lower hybrid (LH) radio frequency heating effects on toroidal plasma rotation in L-mode Tore Supra plasmas are reported. A database of more than 50 plasma discharges has been analysed. Core rotation is found to increment in co- or counter-current direction depending on the plasma current (Ip). At low plasma current, the induced rotation is up to +15 km s−1 in the co-current direction, the rotation profile being affected over the whole plasma minor radius. At higher plasma current, an opposite trend is observed, the core plasma rotation incrementing up to −15 km s−1 in the counter-current direction, the profile being affected up to r/a < 0.6 only. At the zero crossing point, which is defined when the plasma rotation profile is not affected by LH power injection, Ip ∼ 0.95 MA. In both low and high Ip cases, rotation increments are found to increase with the injected power. Several mechanisms in competition which can induce co- or counter-current rotation in Tore Supra LHCD plasmas are investigated and typical order of magnitude are discussed. How those effects evolve with plasma parameters and how they compete are important issues addressed in this paper. Rotation increment increase with Ip at fixed LH power is consistent with a dominant standard momentum confinement mechanism related to Ip increase. The co-current change in rotation is consistent with a fast electron ripple loss mechanism, while thermal ripple induced neoclassical friction and absorbed LH wave momentum from resonant electrons are expected to influence the rotation in the counter-current direction. Finally, the numerical simulations show that the radial turbulent momentum transport does impact the rotation behaviour inducing increment in co- or counter-current directions, depending on the plasma current amplitude.

RF calculation and thermal-stress analysis of the ECRH launcher on EAST tokamak

EAST is a medium sized superconducting tokamak with major radius R=1.8m, minor radius a=0.45m, plasma current Ip≤1MA, toroidal field BT≤3.5T and expected plasma pulse length up to 1000s. An electron cyclotron resonance heating (ECRH) launcher for four-beam injection is being installed on EAST tokamak. Four electron cyclotron wave beams which are generated from four sets of 140GHz/1MW/1000s gyrotrons will be injected into the plasma by the spherical focusing mirrors and plane mobile mirrors. The focusing mirrors are spherical to focus Gaussian beams after reflection. Four plane mobile mirrors independently steer continuously in the poloidal and toroidal direction controlled by motors. With the suitable distance between mirrors and appropriate focal length of focusing mirror, the beam radius in the resonance layer of plasma is 31.145mm. The heat from plasma radiation and metal losses is loaded on the mobile mirror. In order to decrease the temperature and thermal stress, the inner equivalent diameter of water channels is 8mm and the suggested water velocity is 4m/s.

On the non-stiffness of edge transport in L-mode tokamak plasmas

Transport analyses using first-principle turbulence codes and 112-D transport codes usually study radial transport properties between the tokamak plasma magnetic axis and a normalized minor radius around 0.8. In this region, heat transport shows significantly stiff properties resulting in temperature scalelength values (R∕LT) that are relatively independent of the level of the radial heat flux. We have studied experimentally in the tokamak à configuration variable [F. Hofmann et al., Plasma Phys. Controlled Fusion 36, B277 (1994)] the radial electron transport properties of the edge region, close to the last closed flux surface, namely, between ρV=V/Vedge=0.8 to 1. It is shown that electron transport is not stiff in this region and high R∕LTe values (∼20–40) can be attained even for L-mode confinement. We can define a “pedestal” location, already in L-mode regimes, where the transport characteristics change from constant logarithmic gradient, inside ρV = 0.8, to constant gradient between 0.8 and 1.0. In particular, we demonstrate, with well resolved Te and ne profiles, that the confinement improvement with plasma current Ip, with or without auxiliary heating, is due to this non-stiff edge region. This new result is used to explain the significant confinement improvement observed with negative triangularity, which could not be explained by theory to date. Preliminary local gyrokinetic simulations are now consistent with an edge, less stiff, region that is more sensitive to triangularity than further inside. We also show that increasing the electron cyclotron heating power increases the edge temperature inverse scalelength, in contrast to the value in the main plasma region. The dependence of confinement on density in ohmic plasmas is also studied and brings new insight in the understanding of the transition between linear and saturated confinement regimes, as well as of the density limit and appearance of a 2/1 tearing mode. The results presented in this paper provide an important new perspective with regards to radial transport in tokamak plasmas which goes beyond L-mode plasmas and explains some previous puzzling results. It is proposed that understanding the transport properties in this edge non-stiff region will also help in understanding the improved and high confinement edge properties.

Plasma current asymmetries during disruptions in JET

A key feature of disruptions during vertical displacement events, discovered in JET in 1996, is the toroidal variation in the measured plasma current Ip, i.e. the plasma current asymmetries, lasting for almost the entire current quench. The unique magnetic diagnostics at JET (full set of poloidal coils and saddle loops recorded either from two toroidally opposite or from four toroidally orthogonal locations) allow for a comprehensive analysis of asymmetrical disruptions with a large scale database. This paper presents an analysis of 4854 disruptions over an 18 year period that includes both the JET carbon (C) wall and the ITER-like (IL) wall (a mixed beryllium/tungsten first wall). In spite of the Ip quench time significantly increasing for the IL-wall compared to C-wall disruptions, the observed toroidal asymmetry time integral (∼ sideways force impulse), did not increase for IL-wall disruptions. The Ip asymmetry has a dominantly n = 1 structure. Its motion in the toroidal direction has a sporadic behaviour, in general. The distributions of the number of rotation periods are found to be very similar for both C- and IL-wall disruptions, and multi-turn rotation was sometimes observed. The Ip asymmetry amplitude has no degradation with rotation frequency for either the C- or IL-wall disruption. Therefore dynamic amplification remains a potentially serious issue for ITER due to possible mechanical resonance of the machine components with the rotating asymmetry.

Progress and improvement of KSTAR plasma control using model-based control simulators

Superconducting tokamaks like KSTAR, EAST and ITER need elaborate magnetic controls mainly due to either the demanding experiment schedule or tighter hardware limitations caused by the superconducting coils. In order to reduce the operation runtime requirements, two types of plasma simulators for the KSTAR plasma control system (PCS) have been developed for improving axisymmetric magnetic controls. The first one is an open-loop type, which can reproduce the control done in an old shot by loading the corresponding diagnostics data and PCS setup. The other one, a closed-loop simulator based on a linear nonrigid plasma model, is designed to simulate dynamic responses of the plasma equilibrium and plasma current (Ip) due to changes of the axisymmetric poloidal field (PF) coil currents, poloidal beta, and internal inductance. The closed-loop simulator is the one that actually can test and enable alteration of the feedback control setup for the next shot. The simulators have been used routinely in 2012 plasma campaign, and the experimental performances of the axisymmetric shape control algorithm are enhanced. Quality of the real-time EFIT has been enhanced by utilizations of the open-loop type. Using the closed-loop type, the decoupling scheme of the plasma current control and axisymmetric shape controls are verified through both the simulations and experiments. By combining with the relay feedback tuning algorithm, the improved controls helped to maintain the shape suitable for longer H-mode (10–16s) with the number of required commissioning shots largely reduced.

Plasma Density Suppression by Electron Cyclotron Wave in Lower Hybrid Wave Driven TST-2 Spherical Tokamak Plasma

For successful plasma current (Ip) ramp-up by the lower hybrid wave (LHW), the plasma density must be kept at a low level. In the TST-2 Ip ramp-up experiment using the LHW, the application of the electron cyclotron wave (ECW) was found to reduce the plasma density by about 10%. This allowed for further Ip ramp-up later in the discharge.

Electron Temperature Measurement by Thomson Scattering in a Low-Aspect-Ratio RFP RELAX

A Thomson scattering diagnostic system has been developed for measuring electron temperatures in low-aspect-ratio reversed-field pinch (RFP) plasmas in REversed-field pinch of Low Aspect ratio eXperiment (RELAX). In the range of plasma currents Ip = 50 - 80 kA, the central electron temperature was around 100 eV and, showed a weakly increasing trend as Ip increases. To estimate the central electron pressure pe0, a density calibration was performed from simultaneous measurements with a 104-GHz microwave interferometer. The maximum pe0 increased with Ip up to ∼ 70 kA to preserve the approximate relation pe0/Bθa2 ∼ constant, where Bθa is the edge poloidal field. In the higher current region, pe0 tended to saturate, which may be improved by optimizing minute control of equilibrium in the higher current region.

Overview of physics results from the conclusive operation of the National Spherical Torus Experiment

Research on the National Spherical Torus Experiment, NSTX, targets physics understanding needed for extrapolation to a steady-state ST Fusion Nuclear Science Facility, pilot plant, or DEMO. The unique ST operational space is leveraged to test physics theories for next-step tokamak operation, including ITER. Present research also examines implications for the coming device upgrade, NSTX-U. An energy confinement time, τE, scaling unified for varied wall conditions exhibits a strong improvement of BTτE with decreased electron collisionality, accentuated by lithium (Li) wall conditioning. This result is consistent with nonlinear microtearing simulations that match the experimental electron diffusivity quantitatively and predict reduced electron heat transport at lower collisionality. Beam-emission spectroscopy measurements in the steep gradient region of the pedestal indicate the poloidal correlation length of turbulence of about ten ion gyroradii increases at higher electron density gradient and lower Ti gradient, consistent with turbulence caused by trapped electron instabilities. Density fluctuations in the pedestal top region indicate ion-scale microturbulence compatible with ion temperature gradient and/or kinetic ballooning mode instabilities. Plasma characteristics change nearly continuously with increasing Li evaporation and edge localized modes (ELMs) stabilize due to edge density gradient alteration. Global mode stability studies show stabilizing resonant kinetic effects are enhanced at lower collisionality, but in stark contrast have almost no dependence on collisionality when the plasma is off-resonance. Combined resistive wall mode radial and poloidal field sensor feedback was used to control n = 1 perturbations and improve stability. The disruption probability due to unstable resistive wall modes (RWMs) was surprisingly reduced at very high βN/li > 10 consistent with low frequency magnetohydrodynamic spectroscopy measurements of mode stability. Greater instability seen at intermediate βN is consistent with decreased kinetic RWM stabilization. A model-based RWM state-space controller produced long-pulse discharges exceeding βN = 6.4 and βN/li = 13. Precursor analysis shows 96.3% of disruptions can be predicted with 10 ms warning and a false positive rate of only 2.8%. Disruption halo currents rotate toroidally and can have significant toroidal asymmetry. Global kinks cause measured fast ion redistribution, with full-orbit calculations showing redistribution from the core outward and towards V∥/V = 1 where destabilizing compressional Alfvén eigenmode resonances are expected. Applied 3D fields altered global Alfvén eigenmode characteristics. High-harmonic fast-wave (HHFW) power couples to field lines across the entire width of the scrape-off layer, showing the importance of the inclusion of this phenomenon in designing future RF systems. The snowflake divertor configuration enhanced by radiative detachment showed large reductions in both steady-state and ELM heat fluxes (ELMing peak values down from 19 MW m−2 to less than 1.5 MW m−2). Toroidal asymmetry of heat deposition was observed during ELMs or by 3D fields. The heating power required for accessing H-mode decreased by 30% as the triangularity was decreased by moving the X-point to larger radius, consistent with calculations of the dependence of E × B shear in the edge region on ion heat flux and X-point radius. Co-axial helicity injection reduced the inductive start-up flux, with plasmas ramped to 1 MA requiring 35% less inductive flux. Non-inductive current fraction (NICF) up to 65% is reached experimentally with neutral beam injection at plasma current Ip = 0.7 MA and between 70–100% with HHFW application at Ip = 0.3 MA. NSTX-U scenario development calculations project 100% NICF for a large range of 0.6 < Ip(MA) < 1.35.

Diamagnetic measurements by concentric loops in the HL-2A tokamak

The diamagnetic concentric loop method in the HL-2A tokamak is described in this article. The system consists of two concentric poloidal loops with different areas enclosing the plasma column and a short time constant differential integrator, RC < 1 ms. The diamagnetic flux in HL-2A ranges from 1 mWb to 2 mWb for typical discharges with plasma current Ip = 100-400 kA. The integrator output ranges from 0.1 V to 0.2 V with time constant RC = 0.5 ms, and differential area ΔS∕Sout ≈ 7%. Using hybrid analog-digital compensation, the integration drift can be well compensated within 5 mV∕10 s, which can meet the requirement of the concentric loop system. In this method, the measurement of differential area ΔS is not required. The vacuum toroidal flux can be compensated by adjusting the resistance in the integration circuit for several discharges with toroidal field only, which minimizes the additional error produced by a measurement of differential area. The diamagnetic concentric loop system improved the signal to noise ratio by using the short time constant integration. The system with a resolution of ±0.2 kJ can be used to study rapid changes in plasma stored energy, such as the additional power absorbed by the plasma, and the energy loss caused by edge localized modes.

Density limit experiments on FTU

One of the main problems in tokamak fusion devices concerns the capability to operate at a high plasma density, which is observed to be limited by the appearance of catastrophic events causing loss of plasma confinement. The commonly used empirical scaling law for the density limit is the Greenwald limit, predicting that the maximum achievable line-averaged density along a central chord depends only on the average plasma current density. However, the Greenwald density limit has been exceeded in tokamak experiments in the case of peaked density profiles, indicating that the edge density is the real parameter responsible for the density limit. Recently, it has been shown on the Frascati Tokamak Upgrade (FTU) that the Greenwald density limit is exceeded in gas-fuelled discharges with a high value of the edge safety factor. In order to understand this behaviour, dedicated density limit experiments were performed on FTU, in which the high density domain was explored in a wide range of values of plasma current (Ip = 500–900 kA) and toroidal magnetic field (BT = 4–8 T). These experiments confirm the edge nature of the density limit, as a Greenwald-like scaling holds for the maximum achievable line-averaged density along a peripheral chord passing at r/a ≃ 4/5. On the other hand, the maximum achievable line-averaged density along a central chord does not depend on the average plasma current density and essentially depends on the toroidal magnetic field only. This behaviour is explained in terms of density profile peaking in the high density domain, with a peaking factor at the disruption depending on the edge safety factor. The possibility that the MARFE (multifaced asymmetric radiation from the edge) phenomenon is the cause of the peaking has been considered, with the MARFE believed to form a channel for the penetration of the neutral particles into deeper layers of the plasma. Finally, the magnetohydrodynamic (MHD) analysis has shown that also the central line-averaged density at the onset of the MHD activity depends only on the toroidal magnetic field.

Engineering design for the HL-2M tokamak components

The HL-2A tokamak will be modified into HL-2M. The Bt at the plasma center (major radius R=1.78m) is 2.2T, the minor radius is 0.65m. The plasma current IP of HL-2M will reach up to 2.5MA, the elongation and triangularity is more than 1.8 and more than 0.5, respectively. The vacuum vessel torus consists of 20 sectors with “D” shaped cross-section and double wall structure. 20 toroidal field coil bundles comprise 140 turns which are designed with demountable joints, the poloidal field coils system consists of 25 coils. The engineering design and calculation for field coil system, vacuum vessel, support structure, etc. are finished, many key issues for manufacture process have been discussed with industry and the fabrication of main components of HL-2M tokamak will be carried out in factories.

Non-inductive plasma initiation and plasma current ramp-up on the TST-2 spherical tokamak

Plasma current (Ip) start-up in a spherical tokamak (ST) by waves in the lower-hybrid (LH) frequency range was investigated on TST-2. A low current (∼1 kA) ST configuration can be formed by waves over a broad frequency range (21 MHz–8.2 GHz in TST-2), but further Ip ramp-up (to ∼10 kA) is most efficient with waves in the LH frequency range. Ip ramp-up to 15 kA was achieved with 60 kW of net RF power PRF in the fast wave (FW) polarization at 200 MHz excited by the inductively coupled combline antenna. X-ray measurements showed that the photon flux and temperature are higher in the direction opposite to Ip, consistent with acceleration of electrons by a uni-directional RF wave. There is evidence that the LH wave is excited nonlinearly by the FW, based on the frequency spectra measured by magnetic probes. Similar efficiencies of Ip ramp-up were obtained with the inductive combline antenna and the dielectric-loaded waveguide array (‘grill’) antenna, and tendencies for the current drive efficiency to increase with plasma current and toroidal field were observed. During operation of the grill antenna, wavevector components were measured by an array of magnetic probes. Results were qualitatively consistent with expectations based on dispersion relations for the FW and the LH wave. A capacitively coupled combline antenna has been developed to improve coupling to the plasma and the wavenumber spectrum of the excited LH wave, and will be tested in 2013.

Investigation of runaway electrons in the current ramp-up by a fully non-inductive lower hybrid current drive on the EAST tokamak

The possibility of using a lower hybrid wave (LHW) to ramp up the plasma current (Ip) from a low level to a high enough level required for fusion burn in the EAST (experimental advanced superconducting tokamak) tokamak is examined experimentally. The focus in this paper is on investigating how the relevant plasma parameters evolve during the current ramp-up (CRU) phase driving by a lower hybrid current drive (LHCD) with poloidal field (PF) coil cut-off, especially the behaviors of runaway electrons generated during the CRU phase. It is found that the intensity of runaway electron emission increases first, and then decreases gradually as the discharge goes on under conditions of PF coil cut-off before LHW was launched into plasma, PF coil cut-off at the same time as LHW was launched into plasma, as well as PF coil cut-off after LHW was launched into plasma. The relevant plasma parameters, including Hα line emission (Ha), impurity line emission (UV), soft x-ray emission and electron density , increase to a high level. The loop voltage decreases from positive to negative, and then becomes zero because of the cut-off of PF coils. Also, the magnetohydrodynamic activity takes place during the CRU driving by LHCD.

High current and low q95 scenario studies for FAST in the view of ITER and DEMO

The Fusion Advanced Study Torus (FAST) has been proposed as a possible European satellite, in view of ITER and DEMO, in order to: (a) explore plasma wall interaction in reactor relevant conditions, (b) test tools and scenarios for safe and reliable tokamak operation up to the border of stability, and (c) address fusion plasmas with a significant population of fast particles. A new FAST scenario has been designed focusing on low-q operation, at plasma current IP=10MA, toroidal field BT=8.5T, with a q95≈2.3 that would correspond to IP≈20MA in ITER. The flat-top of the discharge can last a couple of seconds (i.e. half the diffusive resistive time and twice the energy confinement time), and is limited by the heating of the toroidal field coils. A preliminary evaluation of the end-of-pulse temperatures and of the electromagnetic forces acting on the central solenoid pack and poloidal field coils has been performed. Moreover, a VDE plasma disruption has been simulated and the maximum total vertical force applied on the vacuum vessel has been estimated.

Confinement of energetic ions in a tokamak plasma at magnetic field in the range of 0.7–1.0 T

We have experimentally studied the influence of toroidal magnetic field (BT) and plasma current (Ip) on the capture and confinement of energetic ions (EIs) formed upon ionization of a neutral beam injected in a tokamak. Based on the results of measurements of the flux of 2.45-MeV fusion neutrons, it is concluded that the amount of EIs significantly grows with increasing BT from 0.7 to 1.0 T and Ip from 140 to 180 kA. In addition to the classical Coulomb slowing down, a supplementary channel of EI losses is found that accounts for a 15% decrease in their confinement time.

Dependence of the density limit on the toroidal magnetic field on FTU

The capability of predicting the density limit of a magnetically confined burning plasma is of crucial importance to establish the ultimate performance of a fusion power plant. The Greenwald density limit, commonly used as an empirical scaling law, predicts that the maximum achievable central line-averaged density is given by the relation , where is the average plasma current density and k is the plasma elongation. However, several experiments have pointed out that such a limit can be overcome in the presence of peaked density profiles. This paper proposes a new empirical scaling law for a limiter tokamak operating in the low-energy confinement mode (L-mode) concerning the case of peaked density profiles associated with the presence of multifaceted asymmetric radiation from the edges. This result is based on dedicated experiments performed on the Frascati Tokamak Upgrade (FTU) under extremely clean machine conditions (Zeff = 1.0–1.5), in which the high-density domain is explored in a wide range of values of plasma current (Ip = 500–900 kA) and toroidal magnetic field (BT = 4–8 T). It is found that the maximum achievable central line-averaged density essentially depends on the toroidal magnetic field only and does not depend on the average plasma current density: the behaviour is explained in terms of density profile peaking in the high-density domain. As a confirmation that the limit is an edge limit, it is also shown that a Greenwald-like scaling (i.e. depending on the current density) actually holds for the edge line-averaged density (at r/a ≃ 4/5).