Scrape-off Layer Plasma Research Articles

RF optimisation of the port plug layout and performance assessment of the ITER ICRF antenna

The ITER Ion Cyclotron Range of Frequencies (ICRF) antenna's capacity to couple power to plasma is determined by the plasma Scrape-Off Layer (SOL) profile, shaping of the front strap array, layout of the Port Plug (PP) and detailed design of its RF components. The first two factors are taken into account by the Torino Polytechnic Ion Cyclotron Antenna (TOPICA) calculated strap array Scattering/Impedance 24-port (S24×24−/Z24×24−) matrices, while this paper deals with the optimisation of the PP layout and components. The RF modeling techniques are explained and used to maximise the coupled power under a set of constraints on RF quantities inside the PP. The total PP RF surface conductive and volumetric dielectric losses are calculated. The resulting S-parameters at the rear RF PP flanges are evaluated as input for the design of the pre-match, decoupling and matching network outside the PP. A discussion of the effect of errors on the PP excitation on the coupled power is also included.

A kinetic model of retarding field analyser measurements in strongly magnetized, flowing, collisional plasmas

Interpretation of retarding field analyser (RFA) measurements in flowing tokamak scrape-off layer plasmas requires a model of the perturbation introduced by the wake, or presheath, of the instrument. We extend a collisionless kinetic Mach probe theory to include a simple model of ion–ion collisions. Collisions lead to significant modifications in the downstream presheath. Depending on the ion-to-electron temperature ratio, flow speed and collisionality, the measurement obtained by averaging the apparent temperatures on each side of a bi-directional RFA is predicted to overestimate the unperturbed ion temperature, but with errors not exceeding 50% most of the time. The error increases with Mach number, and decreases with the ion-to-electron temperature ratio.

Carbon deposition at the bottom of gaps in TEXTOR experiments

Abstract Results of a new dedicated experiment addressing the problem of impurity deposition at the bottom in gaps are presented along with modelling. A test limiter with an isolated gap was exposed to the scrape-off layer plasma in TEXTOR. The exposure was accompanied by injection of 13 C-marked methane in the vicinity of the gap. Deposition at the bottom of the gap was monitored in situ with Quartz-Microbalance diagnostics. The 13 C deposition efficiency of about 2.6 × 10 −5 was measured. Post mortem analysis of resulting deposited layers performed with SIMS and EPMA techniques yields about a factor 2 smaller value corresponding to approximately 10% contribution of the gap bottom to the total 13 C deposition in the gap. This measured contribution is effectively much smaller than observed earlier in TEXTOR, taking the difference in geometry into account, and is in reasonable agreement with modelling performed with ERO and 3D-GAPS codes.

Dust–surface collisions in adhesion regime for tokamak relevant materials

Abstract Empirical studies of dust–surface collisions for metal and carbon projectiles impacting on metal targets in the velocity range from a few m/s to those in excess of 1 km/s have been carried out with the use of a modified pellet injection system. The selected projectile/target shapes, sizes and materials are mimicking the scenario of dust colliding with plasma facing components (PFCs) of a metal machine. The low velocities (10’s m/s) reported here, characteristic for dust motion in tokamak scrape-off layer plasmas, are in the range of sticking phenomenon; the critical velocity and size for bouncing off are essential and timely inputs for the dust dynamics codes and statistical models aiming to predict dust transport and redeposition on PFCs.

Integrated self-consistent analysis of NSTX performance during normal operation and disruptions

Abstract A fundamental issue in tokamak operation related to power exhaust is the understanding of heat and particle transport from core plasma into scrape-off-layer (SOL) and to plasma facing materials during plasma instabilities. We enhanced our models and upgraded HEIGHTS package using adaptive mesh refinement (AMR) to fit reactor walls arbitrary geometry and study/implementing nanoscale surface processes. A five-layer quadtree refinement scheme is used for simulation of the entire SOL plasma evolution and the extra refinement mesh (∼0.5 μm) allowed detail calculations of subsurface particle implantation, divertor hydrodynamic evolution, and erosion processes. Using NSTX actual geometry and magnetic field structure we modeled in full 3D the evolution of escaped core plasma particles starting at core border, gyration and scatterings in SOL, and deposition into plasma facing components. The characteristics of the escaped particles determined the energy source and boundary conditions for edge plasma MHD evolution and the resulting damage.

Effect of separatrix magnetic geometry on divertor behavior in DIII-D

We report on recent experiments on DIII-D that examined the effects that variations in the parallel connection length in the scrape-off layer (SOL), L||, and the radial location of the outer divertor target, RTAR, have on divertor plasma properties. Two-point modeling of the SOL plasma predicts that larger values of L|| and RTAR should lower temperature and raise density at the outer divertor target for fixed upstream separatrix density and temperature, i.e., nTAR∝[RTAR]2[L||]6/7 and TTAR∝[RTAR]−2[L||]−4/7. The dependence of nTAR and TTAR on L|| was consistent with our data, but the dependence of nTAR and TTAR on RTAR was not. The surprising result that the divertor plasma parameters did not depend on RTAR in the predicted way may be due to convected heat flux, driven by escaping neutrals, in the more open configuration of the larger RTAR cases. Modeling results using the SOLPS code support this postulate.

Observation of non-Maxwellian electron distributions in the NSTX divertor

The scrape-off layer plasma at the tokamak region is characterized by open field lines and often contains large variations in plasma properties along these field-lines. Proper characterization of local plasma conditions is critical to assessing plasma–material interaction processes occurring at the target. Langmuir probes are frequently employed in tokamak divertors but are challenging to interpretation. A kinetic interpretation for Langmuir probes in NSTX has yielded non-Maxwellian electron distributions in the divertor characterized by cool bulk populations and energetic tail populations with temperatures of 2–4 times the bulk. Spectroscopic analysis and modeling confirms the bulk plasma temperature and density which can only be obtained with the kinetic interpretation.

L-mode radiative plasma edge studies for model validation in ASDEX Upgrade and JET

The presently favoured option for reactor power handling combines metallic plasma-facing components and impurity seeding to achieve highly radiative scrape-off layer and divertor plasmas. It is uncertain whether tolerable divertor power loads will be obtained in this scenario, necessitating the development of predictive modelling tools. L-mode experiments with N2 seeding have been conducted at both ASDEX Upgrade and JET for benchmarking the critically important impurity radiation models in edge fluid codes. In both machines, particle and power loads are observed to first reduce at the inner target, and only then at the outer target. The outer divertor cools down with increasing N seeding rate, evolving from low-recycling conditions to a regime with peak temperature of 8–10eV in both devices. First SOLPS5.0 simulations of N2 seeding in ASDEX Upgrade geometry show a similar in–out asymmetry in the effect of impurity radiation when drifts are activated in the simulations.

Effect of n = 3 perturbation field amplitudes below the ELM triggering threshold on edge and SOL transport in NSTX

The pulsed application of n=3 magnetic perturbation fields with amplitudes below that which triggers ELMs results in distinct, transient responses observable on several edge and divertor diagnostics in NSTX. We refer to these responses as Sub-Threshold Edge Perturbations (STEPs). An analysis of edge measurements suggests that STEPs result in increased transport in the plasma edge and scrape-off layer, which leads to augmentation of the intrinsic strike point splitting due to error fields, i.e., an intensification of the helical divertor footprint flux pattern. These effects are much smaller in magnitude than those of triggered ELMs, and are observed for the duration of the field perturbation measured internal to the vacuum vessel. In addition, STEPs are correlated with changes to the MHD activity, along with transient reductions in the neutron production rate. Ideally the STEPs could be used to provide density control and prevent impurity accumulation, in the same manner that on-demand ELM triggering is used on NSTX, without the impulsive divertor fluxes and potential for damage to plasma facing components associated with ELMs.

Burst statistics in Alcator C-Mod SOL turbulence

Bursty fluctuations in the scrape-off layer (SOL) of Alcator C-Mod have been analyzed using gas puff imaging data. This reveals many of the same fluctuation properties as Langmuir probe measurements, including normal distributed fluctuations in the near SOL region while the far SOL plasma fluctuations are dominated by large amplitude bursts due to radial motion of blob-like structures. Conditional averaging reveals burst wave forms with a fast rise and slow decay and exponentially distributed burst amplitudes and waiting times. Based on this, a stochastic model of burst dynamics is constructed. The model predicts that fluctuation amplitudes should follow a Gamma distribution. This is shown to be a good description of the gas puff imaging data for a range of line-averaged densities.

Electron density measurements in the ITER fusion plasma

The operation of ITER requires high-quality estimates of the plasma electron density over multiple regions in the plasma for plasma evaluation, plasma control and machine protection purposes. Although the density regimes of ITER are not very different from those of existing tokamaks (1018–1021m−3), the severe conditions of the fusion plasma environment present particular challenges to implementing these density diagnostics. In this paper we present an overview of the array of ITER electron density diagnostics designed to measure over the entire ITER domain: plasma core, pedestal, edge, scrape-off layer and divertor. It will focus on the challenges faced in making these measurements, and the technical solutions of the current designs.

Derivation of stochastic differential equations for scrape-off layer plasma fluctuations from experimentally measured statistics

A stochastic differential equation for intermittent plasma density dynamics in magnetic fusion edge plasma is derived, which is consistent with the experimentally measured gamma distribution and the theoretically expected quadratic nonlinearity. The plasma density is driven by a multiplicative Wiener process and evolves on the turbulence correlation time scale, while the linear growth is quadratically damped by the fluctuation level. The sensitivity of intermittency to the nonlinear dynamics is investigated by analyzing the nonlinear Langevin representation of the beta process, which leads to a root-square nonlinearity.

Conditions for Lower Hybrid Current Drive in ITER

To control the plasma current profile represents one of the most important problems of the research of nuclear fusion energy based on the tokamak concept, as in the plasma column the necessary conditions of stability and confinement should be satisfied. This problem can be solved by using the lower hybrid current drive (LHCD) effect, which was demonstrated to occur also at reactor grade high plasma densities provided that a proper method should be utilised, as assessed on FTU (Frascati Tokamak Upgrade). This method, based on theoretical predictions confirmed by experiment, produces relatively high electron temperature at the plasma periphery and scrape-off layer (SOL), consequently reducing the broadening of the spectrum launched by the antenna produced by parasitic wave physics of the edge, namely parametric instability (PI). The new results presented here show that, for kinetic profiles now foreseen for the SOL of ITER, PI is expected to hugely broaden the antenna spectrum and prevent any penetration in the core of the coupled LH power. However, considering the FTU method and assuming higher electron temperature at the edge (which would be however reasonable for ITER) the PI-produced spectral broadening would be mitigated, and enable the penetration of the coupled LH power in the main plasma. By successful LHCD effect, the control of the plasma current profile at normalised minor radius of about 0.8 would be possible, with much higher efficiency than that obtainable by other tools. A very useful reinforce of bootstrap current effects would be thus possible by LHCD in ITER.

Simulation of plasma turbulence in scrape-off layer conditions: the GBS code, simulation results and code validation

Based on the drift-reduced Braginskii equations, the Global Braginskii Solver, GBS, is able to model the scrape-off layer (SOL) plasma turbulence in terms of the interplay between the plasma outflow from the tokamak core, the turbulent transport, and the losses at the vessel. Model equations, the GBS numerical algorithm, and GBS simulation results are described. GBS has been first developed to model turbulence in basic plasma physics devices, such as linear and simple magnetized toroidal devices, which contain some of the main elements of SOL turbulence in a simplified setting. In this paper we summarize the findings obtained from the simulation carried out in these configurations and we report the first simulations of SOL turbulence. We also discuss the validation project that has been carried out together with the GBS development.

Local effects of biased electrodes in the divertor of NSTX

The goal of this paper is to characterize the effects of small non-axisymmetric divertor plate electrodes on the local scrape-off layer plasma. Four small rectangular electrodes were installed into the outer divertor plates of NSTX. When the electrodes were located near the outer divertor strike point and biased positively, there was an increase in the nearby probe currents and probe potentials and an increase in the Li I light emission at the large major radius end of these electrodes. When an electrode located farther outward from the outer divertor strike point was biased positively, there was sometimes a significant decrease in the Li I light emission at the small major radius end of this electrode, but there were no clear effects on the nearby probes. No non-local effects were observed with the biasing of these electrodes.

Self consistent radio-frequency wave propagation and peripheral direct current plasma biasing: Simplified three dimensional non-linear treatment in the “wide sheath” asymptotic regime

A minimal two-field fluid approach is followed to describe the radio-frequency (RF) wave propagation in the bounded scrape-off layer plasma of magnetic fusion devices self-consistently with direct current (DC) biasing of this plasma. The RF and DC parts are coupled by non-linear RF and DC sheath boundary conditions at both ends of open magnetic field lines. The physical model is studied within a simplified framework featuring slow wave (SW) only and lateral walls normal to the straight confinement magnetic field. The possibility is however kept to excite the system by any realistic 2D RF field map imposed at the outer boundary of the simulation domain. The self-consistent RF + DC system is solved explicitly in the asymptotic limit when the width of the sheaths gets very large, for several configurations of the RF excitation and of the target plasma. In the case of 3D parallelepipedic geometry, semi-analytical results are proposed in terms of asymptotic waveguide eigenmodes that can easily be implemented numerically. The validity of the asymptotic treatment is discussed and is illustrated by numerical tests against a quantitative criterion expressed from the simulation parameters. Iterative improvement of the solution from the asymptotic result is also outlined. Throughout the resolution, key physical properties of the solution are presented. The radial penetration of the RF sheath voltages along lateral walls at both ends of the open magnetic field lines can be far deeper than the skin depth characteristic of the SW evanescence. This is interpreted in terms of sheath-plasma wave excitation. Therefore, the proper choice of the inner boundary location is discussed as well as the appropriate boundary conditions to apply there. The asymptotic scaling of various quantities with the amplitude of the input RF excitation is established.

Outer divertor of ASDEX Upgrade in low-density L-mode discharges in forward and reversed magnetic field: I. Comparison between measured plasma conditions and SOLPS5.0 code calculations

The scrape-off layer and divertor plasma conditions have been carefully analysed in dedicated ASDEX Upgrade experiments consisting of low-density L-mode discharges, with both forward and reversed toroidal magnetic fields and plasma currents. In the forward field, the outer divertor plasma is in a low-recycling regime with peak target temperature above 25 eV. In the reversed field with similar main plasma parameters, the target temperature is below 15 eV and the density is 5 times as high as in the forward field, indicating a higher recycling regime in the outer divertor. The SOLPS5.0 code package is used to model these divertor plasmas. Specifically, it is tested whether a combination of input assumptions exists that enables matching the code solution to all outer divertor diagnostic measurements, and whether these assumptions are compatible with constraints imposed by measurements outside the outer divertor. In the forward field, a good level of agreement with multiple outer target measurements is found with assumptions that simultaneously match the measured density and temperature profiles at the outer midplane, where the uncertainty in the radial position of the separatrix is ±0.5 cm. Similar approaches made previously for higher recycling regimes have not led to such a good consistency between all modelled and measured outer divertor parameters. In the reversed field with higher recycling in the outer divertor, a solution consistent with the outer target Langmuir probe measurements cannot be obtained, at least not without significantly compromising the match to the upstream profile measurements. Significant mismatches are observed also between the modelled and measured upstream Mach number in the forward field. These discrepancies question the global validity of the plasma solutions, and their origin is not yet clear. In part II (Aho-Mantila L. et al 2012 Nucl. Fusion 52 103007), the analysis of outer divertor conditions is complemented by local impurity migration studies, using the divertor plasma solutions presented in this paper.

Numerical investigation of edge plasma phenomena in an enhanced D-alpha discharge at Alcator C-Mod: Parallel heat flux and quasi-coherent edge oscillations

Reduced-model scrape-off layer turbulence (SOLT) simulations of an enhanced D-alpha (EDA) H-mode shot observed in the Alcator C-Mod tokamak were conducted to compare with observed variations in the scrape-off-layer (SOL) width of the parallel heat flux profile. In particular, the role of the competition between sheath- and conduction-limited parallel heat fluxes in determining that width was studied for the turbulent SOL plasma that emerged from the simulations. The SOL width decreases with increasing input power and with increasing separatrix temperature in both the experiment and the simulation, consistent with the strong temperature dependence of the parallel heat flux in balance with the perpendicular transport by turbulence and blobs. The particularly strong temperature dependence observed in the case analyzed is attributed to the fact that these simulations produce SOL plasmas which are in the conduction-limited regime for the parallel heat flux. A persistent quasi-coherent (QC) mode dominates the SOLT simulations and bears considerable resemblance to the QC mode observed in C-Mod EDA operation. The SOLT QC mode consists of nonlinearly saturated wave-fronts located just inside the separatrix that are convected poloidally by the mean flow, continuously transporting particles and energy and intermittently emitting blobs into the SOL.

Overview of the physics and engineering design of NSTX upgrade

The spherical tokamak (ST) is a leading candidate for a Fusion Nuclear Science Facility (FNSF) due to its compact size and modular configuration. The National Spherical Torus eXperiment (NSTX) is a MA-class ST facility in the US actively developing the physics basis for an ST-based FNSF. In plasma transport research, ST experiments exhibit a strong (nearly inverse) scaling of normalized confinement with collisionality, and if this trend holds at low collisionality, high fusion neutron fluences could be achievable in very compact ST devices. A major motivation for the NSTX Upgrade (NSTX-U) is to span the next factor of 3–6 reduction in collisionality. To achieve this collisionality reduction with equilibrated profiles, NSTX-U will double the toroidal field, plasma current, and NBI heating power and increase the pulse length from 1–1.5 s to 5–8 s. In the area of stability and advanced scenarios, plasmas with higher aspect ratio and elongation, high βN, and broad current profiles approaching those of an ST-based FNSF have been produced in NSTX using active control of the plasma β and advanced resistive wall mode control. High non-inductive current fractions of 70% have been sustained for many current diffusion times, and the more tangential injection of the 2nd NBI of the Upgrade is projected to increase the NBI current drive by up to a factor of 2 and support 100% non-inductive operation. More tangential NBI injection is also projected to provide non-solenoidal current ramp-up as needed for an ST-based FNSF. In boundary physics, NSTX measures an inverse relationship between the scrape-off layer heat-flux width and plasma current that could unfavourably impact next-step devices. Recently, NSTX has successfully demonstrated substantial heat-flux reduction using a snowflake divertor configuration, and this type of divertor is incorporated in the NSTX-U design. The physics and engineering design supporting NSTX Upgrade is described.

Electrostatic transport in L-mode scrape-off layer plasmas of Tore Supra tokamak. II. Transport by fluctuations

The turbulent transport of particles is investigated using rake probes at the top of the scrape-off layer (SOL) of circular ohmically heated L-mode plasmas in the Tore Supra tokamak [G. Giruzzi et al., Nucl. Fusion 49, 104010 (2009)]. Both radial and poloidal non-linear fluxes are estimated by means of two reciprocating arrays separated toroidally by 120°. The time average of the radial (poloidal) flux is positive (negative) through the whole SOL profile. The respective effective transport velocity are about 〈vr〉t≈30m s−1 and 〈vθ〉t≈−60m s−1 close to the last closed flux surface. Both components present a standard deviation about 10 times higher than their respective mean amplitude, and time-distributions highly skewed toward values of the same sign as their mean values. The existence of a nonlinear poloidal flux is associated with the local tilt of filament eddies due to electric and magnetic shear. At the last closed flux surface, where plasma filaments experience their early life, the orientation of the velocity field is consistent with structure dynamics which originate from the outboard midplane and spread along field lines toward the rest of the poloidal section. The localized tilt of the eddy structures corresponds to the effect of the magnetic shear. Further into the SOL, the orientation of the velocity field evolves along radius in agreement with a simple model of propagating filaments progressively tilted by an electric shear. The combined effects of electric and magnetic shear on the eddy tilting have potentially a crucial impact on the existence of a Reynolds stress 〈vrvθ〉t component, which is strongly poloidally asymmetric at the edge of tokamak plasmas.