Ohmic Phase Research Articles

NBI optimization on SMART and implications for scenario development

The SMall Aspect Ratio Tokamak (SMART) under commissioning at the University of Seville, Spain, aims to explore confinement properties and possible advantages in confinement for compact/spherical tokamaks operating at negative vs. positive triangularity. This work explores the benefits of auxiliary heating through Neutral Beam Injection (NBI) for SMART scenarios beyond the initial Ohmic phase of operations, in support of the device’s mission. Expected values of electron and ion temperature achievable with NBI heating are first predicted for the current flat-top phase, including modeling to optimize the NBI injection geometry to maximize NBI absorption and minimize losses for a given equilibrium. Simulations are then extended for a selected case to cover the current ramp-up phase. Differences with results obtained for the flat-top phase indicate the importance of determining the plasma evolution over time, as well as self-consistently determining the edge plasma parameters for reliable time-dependent simulations. Initial simulation results indicate the advantage of auxiliary NBI heating to achieve nearly double values of pressure and stored energy compared to Ohmic discharges, thus significantly increasing the device’s performance. The scenarios developed in this work will also contribute to diagnostic development and optimization for SMART, as well as providing test cases for initial predictions of macro- and micro-instabilities.

Transport simulations and electric field in T-10 and TJ-II

A modeling of the transport processes in T-10 tokamak using Astra code for two sets of discharges having ECRH auxiliary heating is presented. Electric potential measurements using HIBP show that radial profiles are negative for ohmic heating but become positive for large heating powers. We first obtain T α and n profiles by adjusting transport coefficients and then try to obtain the radial E r profiles by assuming three models for it. Results show good agreement for Ohmic and small ECRH power phases but not for large power where E r > 0 indicating that other processes have to be included.

Predictive JET current ramp-up modelling using QuaLiKiz-neural-network

This work applies the coupled JINTRAC and QuaLiKiz-neural-network (QLKNN) model on the ohmic current ramp-up phase of a JET D discharge. The chosen scenario exhibits a hollow profile attributed to core impurity accumulation, which is observed to worsen with the increasing fuel ion mass from D to T. A dynamic D simulation was validated, evolving j, , , , n Be, n Ni, and n W for 7.25 s along with self-consistent equilibrium calculations, and was consequently extended to simulate a pure T plasma in a predict-first exercise. The light impurity (Be) accounted for Z eff while the heavy impurities (Ni, W) accounted for P rad. This study reveals the role of transport on the hollowing, which originates from the isotope effect on the electron-ion energy exchange affecting . This exercise successfully affirmed isotopic trends from previous H experiments and provided engineering targets used to recreate the D q-profile in T experiments, demonstrating the potential of neural network surrogates for fast routine analysis and discharge design. However, discrepancies were found between the impurity transport behaviour of QuaLiKiz and QLKNN, which lead to notable hollowing differences. Further investigation into the turbulent component of heavy impurity transport is recommended.

Quantification of X3 absorption for ITER L-mode parameters in ASDEX Upgrade

For an early H-mode access in hydrogen, ITER considers operating at 1/3 of the full field using 170 GHz X-Mode for heating at the 3rdharmonic. The optical thickness for such a heating scheme depends onTe2. It is rather low in the ohmic phase (withTeabout 1-2 keV), but reaches high single pass absorption for the strongly EC heated plasma withTeexceeding 10 keV. Launching ECRH into an ohmic plasma may trigger a boot-strap process onTeif the additional power absorption due to increasingTeexceeds the additional power losses due to increased transport (which often tends to increase with input power). In this contribution we present measurements of the X3 absorption for the parameter range relevant for ITER, i.e.ne2 1019m−3,Te2keVin order to back up theoretical estimates used for the modeling so far. In ASDEX Upgrade (AUG) such low densities cannot be reached in H-mode such that dominant heating with NBI is not an option. For moderateTe, it is also not an option to use X3 heating as main heating, due to the excessive stray radiation threatening in-vessel components. This dilemma is solved with the 2-frequency EC system of AUG. The main central heating is done with the lower frequency of 105 GHz at the 2ndharmonic and full single pass absorption. Up to 3.5 MW of ECRH are used at that frequency to varyTe. Two other gyrotrons are used at 140 GHz to probe the X3 interaction close to the plasma center with a sequence of short blips. The expected values of single pass absorption are calculated with TORBEAM and vary from 7% to 70%. Below 40% single pass absorption the non-absorbed power triggers an arc in the tile gaps of the inner heat shield which screens the thermo-couples from the incoming beam such that they cannot be used. Between 40% and 80% single pass absorption, the predictions and measurements agree within the uncertainty of the measurement, unless we have clear evidence for non-linear interactions, which are not described by TORBEAM and which are not expected in ITER, but are due to some specific experimental choices for an isolated subset of our results.

The power threshold of H-mode access in mixed hydrogen–tritium and pure tritium plasmas at JET with ITER-like wall

The heating power to access the high confinement mode (H-mode), P LH, scales approximately inversely with the isotope mass of the main ion plasma species as found in (protonic) hydrogen, deuterium and tritium plasmas in many fusion facilities over the last decades. In first dedicated L–H transition experiments at the Joint European Torus (JET) tokamak facility with the ITER-like wall (ILW), the power threshold, P LH, was studied systematically in plasmas of pure tritium and hydrogen–tritium mixtures at a magnetic field of 1.8 T and a plasma current of 1.7 MA in order to assess whether this scaling still holds in a metallic wall device. The measured power thresholds, P LH, in Ohmically heated tritium plasmas agree well with the expected isotope scaling for metallic walls and the lowest power threshold was found in Ohmic phases at low density. The measured power thresholds in ion cyclotron heated plasmas of pure tritium or hydrogen–tritium mixtures are significantly higher than the expected isotope mass scaling due to higher radiation levels. However, when the radiated power is taken into account, the ion cyclotron heated plasmas exhibit similar power thresholds as a neutral beam heated plasma, and are close to the scaling. The tritium plasmas in this study tended to higher electron heating fractions and, when heated with ion cyclotron waves, to relatively higher radiation fractions compared to other isotopes potentially impeding access to sustained H-modes.

New method for time alignment and time calibration of the TOFOR time-of-flight neutron spectrometer at JET.

The TOFOR time-of-flight (TOF) neutron spectrometer at the Joint European Torus (JET) is composed of 5 start (S1) and 32 stop (S2) scintillation detectors. Recently, the data acquisition system (DAQ) of TOFOR was upgraded to equip each of the 37 detectors with its own waveform digitizer to allow for correlated time and pulse height analysis of the acquired data. Due to varying cable lengths and different pulse processing pathways in the new DAQ system, the 160 (5 · 32) different TOF pairs of start-stop detectors must be time-aligned to enable the proper construction of a summed TOF spectrum. Given the time (energy) resolution required by the entire spectrometer system to measure different plasma neutron emission components, it is of importance to align the detector pairs to each other with sub-nanosecond precision. Previously, the alignment partially depended on using fusion neutron data from Ohmic heating phases of JET experimental pulses. The dependence on fusion neutron data in the time alignment process is, however, unsatisfactory as it involves data one would wish to include in an independent analysis for physics results. In this work, we describe a method of time-aligning the detector pairs by using gamma rays. Given the known geometry and response of TOFOR to gamma rays, the time alignment of the detector pairs is found by examining gamma events interacting in coincidence in both S1-S1 and S1-S2 detector combinations. Furthermore, a technique for separating neutron and gamma events in the different detector sets is presented. Finally, the time-aligned system is used to analyze neutron data from Ohmic phases for different plasma conditions and to estimate the Ohmic fuel ion temperature.

Z eff Measurement in Ohmic, L- and H-Mode Plasmas on HL-2A Tokamak

$Z_{\mathrm {eff}}$ is measured by two visible bremsstrahlung diagnostics on HL-2A tokamak, one of which uses a high spectral resolution spectrometer coupled with an electron multiplied CCD (temporal resolution of 100 Hz), and the other uses a filterscope system (temporal resolution of 1 MHz). The bremsstrahlung brightness measured by them coincides with each other very well. $\bar {Z}_{\mathbf {eff}}$ is systematically measured under different discharge conditions, such as divertor and limiter configurations, ohmic and auxiliary heating phases, before and after siliconization, and $L$ – $H$ transition. Under ohmic phase, $\bar {Z}_{\mathbf {eff}}$ is higher in the limiter configuration than that in the divertor configuration by a factor of ~1.5 under high-density regime. $\bar {Z}_{\mathbf {eff}}$ increases in all cases of auxiliary heating scenarios due to enhanced plasma–wall interaction, accompanied by influx increases of low ionization states of impurities such as C2+ and Fe1+ in plasma edge. Siliconization for wall conditioning is prominently effective in lowering the main impurity influxes in plasma edge and thus the impurity level, and a concentration of ~4% for carbon and ~0.2% for iron is roughly evaluated for a typical nonsiliconized plasma. A simple relationship that $Z_{\mathrm {eff}}$ is positively correlated with $\Phi _{\mathrm {in}}$ and $\tau _{\mathrm {p}}$ is verified from the L - and $H$ -mode features, where $\Phi _{\mathrm {in}}$ is the impurity influx and $\tau _{\mathrm {p}}$ the particle confinement time. A nearly constant $\bar {Z}_{\mathbf {eff}}$ evolution is observed during $L$ – $H$ and $H$ – $L$ transition and it seems that $\bar {Z}_{\mathbf {eff}}$ is insensitive to $L$ – $H$ and $H$ – $L$ transition under some $Z_{\mathrm {eff}}$ (0) threshold. The $Z_{\mathrm {eff}}$ profiles are flat/weakly hollow in plasma core region (rho $Z_{\mathrm {eff}}$ (0) and tend to be peaked when $Z_{\mathrm {eff}}$ (0) is larger than 3, which agrees with the neoclassical theory prediction.

Effect of fuel isotope mass on q-profile formation in JET hybrid plasmas

The initial current ramp phase of JET hybrid plasmas is used to optimise the target q-profile for main heating to allow access to high β and avoid MHD instabilities. Mixed protium-deuterium experiments, carried out at JET since the installation of the beryllium-tungsten wall, have shown that the q-profile evolution during this Ohmic phase varies systematically with average main ion isotope mass, indicating the need for re-optimisation for future T and mixed D-T experiments. Current diffusion modelling shows that the key factor was a reduction in electron temperature profile peaking as the hydrogenic isotope mass was increased. This was correlated with an increase in plasma radiation by metallic impurities, consistent with the increased sputtering yield by higher mass isotopes during the current ramp phase. Reduced electron temperature peaking can lead to magnetic shear reversal and the appearance of a 2/1 mode, which can lock, causing the JET massive gas injection system to be triggered to avoid an unmitigated disruption. The potential for a further reduction in electron temperature peaking in T and D-T plasmas could, therefore, result in an increased risk of disruptions. To mitigate this risk, electron temperature peaking measurements have been included in the real-time control system to allow this type of disruption to be avoided by central heating, density increase or early pulse termination. These experiments indicate the need for integrated modelling of impurity behaviour, including the plasma core, scrape-off layer and plasma wall interactions, to predict q-profile evolution in the current ramp phase and anticipate the effects of isotope changes.

Space-resolved extreme ultraviolet spectrometer for impurity diagnostics in HL-2A

Abstract A space-resolved extreme ultraviolet (EUV) spectrometer system has been developed in the HL-2A tokamak for impurity monitoring and transport studies. A charge-coupled-device (CCD) detector is mounted on a two-stage wavelength scanning mechanism for measuring the spatial profile of impurity line emissions in the wavelength range of 30−600 A. A good spatial profile of impurity line emissions is observed throughout the lower half of the HL-2A plasma along the vertical direction by installing a space-resolved slit in front of an entrance slit. The vertical observation range is ∼43 cm covering a wide plasma area from the plasma central region to the vicinity of divertor separatrix X-point. In order to make a critical analysis of the impurity transport, in particular, in the plasma edge, an accurate position calibration in the observed vertical impurity line intensity profile is necessary. For the purpose a thin metal bar is installed at the diagnositc port. Then, the vertical position of the impurity profile can be accurately determined throughout the entire wavelength range. Based on the present method, the local emissivity profile of impurity line emissions is well reconstructed with good accuracy through Abel inversion technique. Resultant CIV and CVI local emissivity profiles show a clear difference in the carbon transport between ohmic and NBI heating phases.

Electrical conduction mechanisms of metal / high-Tc superconductor (YBCO) interfaces

Current–voltage characteristics of Au/YBa2Cu3O interfaces (Au/YBCO), built on optimally-doped YBCO thin films, grown by pulsed laser deposition, were measured as a function of temperature in the 50 K to 270 K range, for two different resistance states. A non-trivial equivalent circuit model is proposed, which reveals the existence of a highly inhomogeneous scenario composed by two complex layers: one presenting both a non-linear Poole–Frenkel conduction as well as variable range hopping localization effects (probably associated with YBa2Cu3O6) mixed with a minor metallic phase, while the other is also composed by a mixture of YBCO with different oxygen contents, where a metallic ohmic phase still percolates. A microscopic description of the effects produced by the resistance switching is given, showing the evolution of carrier traps, localization effects and dielectric behavior for each state. The dielectric behavior is interpreted in terms of a Maxwell–Wagner scenario.

Analysis of plasma termination in the JET hybrid scenario

This paper analyses the final phase of hybrid scenario discharges at JET, the reduction of auxiliary heating towards finally the Ohmic phase. The here considered Ohmic phase is mostly still in the current flattop but may also be in the current ramp down. For this purpose a database is created of 54 parameters in 7 phases distributed in time of the discharge. It is found that the occurrence of a locked mode is in most cases preceded by a radiation peaking after the main heating phase either in a low power phase and/or in the Ohmic phase. To gain insight on the importance of different parameters in this process a correlation analysis to the radiation peaking in the Ohmic phase is done. The first finding is that the further away in time the analysed phases are the less the correlation is. This means in the end that a good termination scenario might also be able to terminate unhealthy plasmas safely. The second finding is that remaining impurities in the plasma after reducing the heating power in the termination phase are the most important reason for generating a locked mode which can lead to a disruption.

A theory of self-organized zonal flow with fine radial structure in tokamak

The (low frequency) zonal flow-ion temperature gradient (ITG) wave system, constructed on Braginskii's fluid model in tokamak, is shown to be a reaction-diffusion-advection system; it is derived by making use of a multiple spatiotemporal scale technique and two-dimensional (2D) ballooning theory. For real regular group velocities of ITG waves, two distinct temporal processes, sharing a very similar meso-scale radial structure, are identified in the nonlinear self-organized stage. The stationary and quasi-stationary structures reflect a particular feature of the poloidal group velocity. The equation set posed to be an initial value problem is numerically solved for JET low mode parameters; the results are presented in several figures and two movies that show the spatiotemporal evolutions as well as the spectrum analysis—frequency-wave number spectrum, auto power spectrum, and Lissajous diagram. This approach reveals that the zonal flow in tokamak is a local traveling wave. For the quasi-stationary process, the cycle of ITG wave energy is composed of two consecutive phases in distinct spatiotemporal structures: a pair of Cavitons growing and breathing slowly without long range propagation, followed by a sudden decay into many Instantons that carry negative wave energy rapidly into infinity. A spotlight onto the motion of Instantons for a given radial position reproduces a Blob-Hole temporal structure; the occurrence as well as the rapid decay of Caviton into Instantons is triggered by zero-crossing of radial group velocity. A sample of the radial profile of zonal flow contributed from 31 nonlinearly coupled rational surfaces near plasma edge is found to be very similar to that observed in the JET Ohmic phase [J. C. Hillesheim et al., Phys. Rev. Lett. 116, 165002 (2016)]. The theory predicts an interior asymmetric dipole structure associated with the zonal flow that is driven by the gradients of ITG turbulence intensity.

A Comprehensive Single-Phase, Non-Isothermal Mathematical MEA Model and Analysis of Non-Isothermal Effects

A comprehensive single-phase non-isothermal MEA model, which accounts for all applicable heat sources, viz., entropic and irreversible heating associated with electrochemical reactions, protonic and electronic ohmic heating, phase change, and heat release/absorption due to sorption/desorption of water in electrolyte is presented. Starting from a general energy transport equation, layer-specific thermal transport equations are derived. The model expands upon the isothermal model by Secanell et al. [1] and accounts for multi-step reaction kinetics for the oxygen reduction reaction (ORR) and hydrogen oxidation reaction (HOR) [2]. The model accounts for water management in two phases, viz., vapour and sorbed phase, and accounts for thermal effects on water transport such as temperature driven vapour diffusion, and thermal osmosis. The model is integrated into openFCST, a finite element based, open-source, fuel cell simulation framework, which is available for download online at http://www.openfcst.mece.ualberta.ca/.The non-isothermal model is used to estimate the maximum temperature inside the membrane electrode assembly under different operating conditions such as hot and dry, and cold and wet. At low humidity conditions, the mathematical model predicts a significant performance drop in the cell due to drying out of ionomer. The cathode catalyst layer is found to be the hottest, with temperatures rising by more than 10ºC at high current densities (Figure 1). A detailed breakdown of various heat sources inside the cell at different current densities is also provided (Figure 2). It is observed that heat of sorption, which has generally been ignored in previous non-isothermal models, is more significant than protonic ohmic heating at medium current densities, and causes a shift in temperature distribution in the in-plane direction. At high current densities, heat of sorption accounts for roughly 15% of the overall heat generated in the cell, and it is concluded that this term cannot be ignored.There is ambiguity in the literature regarding reversible heat distribution between the half-cell reactions of the ORR and HOR [3]. Based on the data by Ramousse et al. [3], the HOR should be exothermic. This is contrary to the general practice of assuming the HOR to be athermic. Using the non-isothermal model with an exothermic HOR, it is shown that the assumption causes a cell performance drop and the maximum temperature of the cell to shift from the cathode catalyst layer (CCL) to anode catalyst layer (ACL).Thermal management strongly affects water transport in the cell. Therefore, the thermal transport equation is solved with a detailed water management model including thermal-osmosis. The non-isothermal model predicts higher sorbed water movement from the ACL to CCL due to drying out of ionomer in the CCL when compared to an isothermal model (Figure 3). The contribution of thermal osmosis to sorbed water flux is as much as 15% as shown in Figure 3.The thermal conductivity of microporous layer is found to significantly influence the cell performance predictions. If reduced considerably, the model predicts self-heating of the MEA and a complete dryout of the cell (Figure 4).In summary, a detailed single-phase non-isothermal MEA model is developed. The non-isothermal fuel cell model governing equations include a thermal transport equation, a sorbed water transport equation, the mass and charge transport of all species, and multi-step kinetics for HOR and ORR electrochemical reactions. Using the model, the importance of several critical physical phenomena is analyzed.

Enhancement of edge impurity transport with ECRH in the HL-2A tokamak

Edge impurity transport is studied in electron cyclotron resonance heating (ECRH) L-mode plasmas of the HL-2A tokamak based on space-resolved vacuum ultraviolet spectroscopy with which radial profiles of impurity line emissions are measured from the core region inside the last closed flux surface (LCFS) and the edge region in the scrape-off layer, simultaneously. The radial profile of carbon emissions of C V (2271 Å: 1s2s 3S–1s2p 3P) reconstructed into the local emissivity profile is analysed with a one-dimensional impurity transport code, and the diffusion coefficient and convective velocity of impurity ions are determined in the core region of the HL-2A tokamak. The impurity source is also determined with the measured absolute emissivity profiles of C IV (1548 Å: 1s22s 2S–1s22p 2P) located at the LCFS. The ratio of C V to C IV can therefore be used as an index to characterize the core impurity transport between the LCFS and the radial region of the C V emission at a normalized radius of about ρ = 0.6. The ratio measured from ohmic discharges shows a gradual decrease with electron density. However, the ratio suddenly decreases by a factor of three when the ECRH focused in the plasma centre is switched on, suggesting a strong enhancement of the impurity transport. The analysis with the transport code indicates a change in the convective term. The convective velocity of C4+ ions changes from inward to outward direction during the ECRH phase, while an inward velocity usually exists in the ohmic phase. Possible mechanisms for the reversal of the convective velocity are discussed.

Temperature Distribution and Heat Flux on the EAST Divertor Targets in H-Mode

An infrared camera (IR) has been put into operation in the Experimental Advanced Superconducting Tokamak (EAST), which is used to measure the temperature distribution on the surface of lower divertor target plates. With a finite difference method, the heat flux onto the divertor target plates is calculated from the surface temperature profile. The high confinement mode (H-mode) with type-III edge localized modes (ELMs) has been obtained with about 1 MW lower-hybrid wave power on the EAST in the autumn experiment in 2010. The analyzed H-mode discharges were lower single null X-point diverted discharges with a density range of < ne > (1 ∼ 4) × 1019 m−3. The surface temperature of the inner target plate increases with heating power. The peak temperature on the surface of target plates is lower than 200°C with about 2.4 MW heating power. Comparison among the heat flux profiles occurring in different phases in the same discharge has been performed. It indicates that the heat flux profile obviously changes from the ohmic phase to the H-mode phase, and the full width at half maximum (FWHM) of the heat flux profile is the narrowest during the ELM-free H-phase. On the outer target plate, the peak heat flux exceeds 2 MW/m2 during the ELMy H-mode phase, whereas it is only about 0.8 MW/m2 during the ELM-free phase in the same discharge.

Electronically controlled optical beam-steering by an active phased array of metallic nanoantennas

An optical phased array of nanoantenna fabricated in a CMOS compatible silicon photonics process is presented. The optical phased array is fed by low loss silicon waveguides with integrated ohmic thermo-optic phase shifters capable of 2π phase shift with ∼ 15 mW of applied electrical power. By controlling the electrical power to the individual integrated phase shifters fixed wavelength steering of the beam emitted normal to the surface of the wafer of 8° is demonstrated for 1 × 8 phased arrays with periods of both 6 and 9 μm.

Predictive ASCOT modelling of 10Be transport in JET with the ITER-like wall

We model the transport of a beryllium (10Be) marker during a sequence of an inner-wall limited and a diverted Ohmic plasma phase in JET with the objective of identifying principal migration pathways. The 3D orbit-following code ASCOT is used for predictive analysis of an experiment during the 2011–2012 campaign on JET where three central pieces of a wall tile enriched with 10Be were installed to an inner wall guard limiter (IWGL) of the tokamak. Assuming erosion during the inner-wall limited plasma, the simulations indicate that 10Be is deposited along the IWGLs during the limiter phase which, when assuming further erosion, can lead to high deposition on the inner (high-field side) divertor during the diverted phase. In contrast, beryllium confined in the core plasma during the limiter phase is seen to be predominantly uniformly deposited during the diverted phase on the outer (low-field side) wall limiters and divertor tiles.

Residual carbon content in the initial ITER-Like Wall experiments at JET

The residual carbon content and carbon edge flux in JET have been assessed by three independent diagnostic techniques after start of plasma operation with the ITER-Like Wall (ILW) with beryllium first wall and tungsten divertor: (i) in-situ measurements with optical spectroscopy on low ionisation stages of carbon, (ii) charge-exchange recombination spectroscopy, and (iii) residual gas composition analysis in dedicated global gas balance experiments. Direct comparison experiments in L-mode discharges were carried out between references from the previously installed material configuration with plasma-facing components made of carbon-fibre composite (JET-CFC) and the JET-ILW. The temporal evolution of the C divertor flux since installation of the ILW has been studied in the ohmic phase of dedicated monitoring discharges which have been executed regularly throughout the experimental exploitation so far (60000 plasma seconds). The C flux behaviour in the divertor can be divided in three phases: initial fast drop, moderate reduction phase, and a long lasting phase with almost constant C flux. The Be flux in both divertor legs mirrors the behaviour of C. All experiments and diagnostic techniques demonstrate a strong reduction in C fluxes and C content of more than one order of magnitude with respect to JET-CFC which is in line with the reduction in long-term fuel retention due to co-deposition. There is no evidence of an increase in residual carbon in time, thus no indication that a damage of the thin tungsten coatings on CFC substrate in the divertor occurred.

ICRF specific plasma wall interactions in JET with the ITER-like wall

A variety of plasma wall interactions (PWIs) during operation of the so-called A2 ICRF antennas is observed in JET with the ITER-like wall. Amongst effects of the PWIs, the W content increase is the most significant, especially at low plasma densities. No increase of W source from the main divertor and entrance of the outer divertor during ICRF compared to NBI phases was found by means of spectroscopic and WI (400.9nm) imaging diagnostics. In contrary, the W flux there is higher during NBI. Charge exchange neutrals of hydrogen isotopes could be excluded as considerable contributors to the W source. The high W content in ICRF heated limiter discharges suggests the possibility of other W sources than the divertor alone. Dependencies of PWIs to individual ICRF antennas during q95-scans, and intensification of those for the −90° phasing, indicate a link between the PWIs and the antenna near-fields. The PWIs include heat loads and Be sputtering pattern on antenna limiters. Indications of some PWIs at the outer divertor entrance are observed which do not result in higher W flux compared to the NBI phases, but are characterized by small antenna-specific (up to 25% with respect to ohmic phases) bipolar variations of WI emission. The first TOPICA calculations show a particularity of the A2 antennas compared to the ITER antenna, due to the presence of long antenna limiters in the RF image current loop and thus high near-fields across the most part of the JET outer wall.

Effect of lower hybrid waves on turbulence and transport of particles and energy in the FTU tokamak scrape-off layer plasma

All the main features of the scrape-off layer turbulence, magnitude, frequency spectrum and perpendicular wave vector, ξt, are strongly affected by the injection of lower hybrid (LH) power into the FTU tokamak. The governing parameters are the local last closed magnetic surface values of density, ne,LCMS, and temperature, Te,LCMS. ne,LCMS determines the perpendicular wave vector of the LH waves, which is a key parameter for the multiple scattering processes, and together with Te,LCMS the collisionality that exerts a stabilizing effect on the fluctuations. This effect, still to be examined in the light of theoretical models, leads to an asymptotic value for the fluctuation relative amplitude in the ohmic phase close to 25%, and ∼10% in the LH phase, or even less, since the saturation level is not yet attained. The LH waves also can strongly raise ξt, about 3 times, and double the root mean square frequency. The transfer of momentum and energy in the mutual scattering of LH and turbulence ‘waves’ drives these changes. An increase also of the cross-correlation between temperature and electric potential fluctuations should occur in order to explain the magnitude of the fluctuation amplitude drop and the large increment of the temperature e-folding decay, by more than a factor of 2.5. Particle transport, however, does not appear to be affected to a large extent—the density e-folding decay length is almost unchanged but the power flow typical length rises by about a factor of 1.5, which is a relevant figure in view of the problem of mitigating the power loads on divertor targets in future reactors. These changes are confined mainly within the flux tube connected with the LH waves launching antenna, but start to spread significantly out of it at high plasma densities.