Core Confinement Research Articles

The role of thermal instabilities in limiting the density in DIII-D

The mechanisms which apparently govern the maximum achievable density in several DIII-D [Luxon, Anderson, Batty et al., Plasma Physics on Controlled Nuclear Fusion Research 1986 (IAEA, Vienna, 1987), Vol. 1, p. 159] shots in which different operating procedures were used in building up the density were investigated and compared with the predictions of thermal instability theory. Core MARFEs (multifaceted asymmetric radiation from the edge) followed by a H–L (high-to-low confinement mode) transition limit the density well below the Greenwald limit in continuous gas puffed lower single-null discharges with low triangularity. Similar continuous gas puff fueled discharges with higher triangularity or with pumping in the lower divertor achieve densities at or above the Greenwald value, apparently limited by confinement degradation, without the formation of core MARFEs. Pellet fueled discharges achieve densities up to twice the Greenwald value, limited by global radiative collapse. Thermal instability theory predictions of the limiting core MARFEs, confinement degradation or global radiative collapse are in good agreement with the experimental observations for the shots examined. Evidence for an important role of neutral particles in the plasma edge in core MARFE onset and in confinement degradation was identified.

Novel data processing tools to localize turbulence and MHD activity from broadband reflectometry on ASDEX upgrade

Broadband microwave reflectometry is a diagnostic technique that provides electron density profiles with both high temporal and spatial resolution. The information extracted is the distance to the cut-off layer where the waves are reflected. Here we analyze the spectral component due to distance as well as components associated with plasma turbulence that are usually filtered out for profile evaluation. The new analysis is applied to ASDEX Upgrade regimes, with improved core confinement and H-mode edge. It enables to detect the reduction of turbulence at the edge and core, following the formation of the edge and internal transport barriers, respectively, and also the increase of turbulence associated with edge localized modes. The potentialities and limitations of this analysis to study plasma turbulence are discussed.

The physics of large and small edge localized modes

The operational boundaries for the occurrence ofvarious types of edge localized mode (ELM) instabilityin high-confinement-mode (H-mode) fusion plasmas are reviewed.The phenomenology of ELMs and the occurrence of differentELM types in parameter space are described.Favourable ELM regimes are observed in various experimentswhich combine small ELM losses withhigh edge pressure and thereby allow for good H-mode confinementeven in situations where core confinement is strongly dependenton edge parameters.Recent progress of ELM models based on ideal or resistive MHDis described and their ability to describethe various ELM regimes discussed.

Mechanism of the transport barriers formation at lowerhybrid heating in the FT-2 tokamak experiments

The possibility of controlling the transport processes in thetokamak plasma in the lower hybrid heating (LHH) experiment hasbeen demonstrated. We illustrate experimentally the observedtransport barrier formation initialized by the LHH fordifferent plasma experiment scenarios. First, it was found duringLHH. The next method to trigger improved confinement is acombination of fast current ramp-up with LHH. The mechanisms ofinternal barrier formation have been put forward to explain theobserved regime of improved core confinement. The increasedshear of the radial electric field stimulates the internal barrierformation.

Measurements and scalings of the H-mode pedestal on Alcator C-Mod

In this paper we summarize the results of recent measurements of thehigh confinement mode (H-mode) edge transport barrier in Alcator C-Mod. Anumber of new, high-resolution edge diagnostics provide measurements of plasma parameters (electron density and temperature, and impurity density)in the pedestal region. Pedestal parameters are measured both at the top edgeof the plasma and at the midplane. Very narrow (2-5 mm) electron density andtemperature and impurity density pedestals are observed. The impurity pedestalis located inside the density and temperature pedestals in the midplanemeasurements, but coincides with the ne pedestal at the top edge.Correlation of both width and height of the pedestals at the midplane withglobal plasma parameters and core confinement characteristics isobserved.At the same time there are no clear indications of scaling for the electrondensity and impurity pedestal widths measured at the top of the plasma.

Energy confinement time and heat transport in initial neutral beam heated plasmas on the large helical device

The confinement characteristics of large net-current-free plasmas heated by neutral-beam injection have been investigated in the Large Helical Device (LHD). A systematic enhancement in energy-confinement times from the scaling derived from the medium-sized heliotron/torsatron experiments have been observed, which is attributed to the edge pedestal. The core confinement is scaled with the Bohm term divided by the square root of the gyro radii. The comparative analysis using a dimensionally similar discharge in the Compact Helical System indicates gyro-Bohm dependence in the core and transport improvement in the edge region of LHD plasmas.

Scaling of Local Core Transport with Lundquist Number in the Reversed Field Pinch

We investigate the scaling of the core energy transport with the magnetic Lundquist number $S$ in the reversed field pinch. We analyze for the first time the electron temperature and thermal conductivity profiles and the dynamo magnetic fluctuations in a wide range of stationary plasmas produced in the RFX device at plasma current ranging between 0.2 and 1.1 MA. When $S$ increases we measure an improvement of core confinement associated with the strengthening of the internal electron temperature profiles gradient and with the decrease of magnetic turbulence. This shows that core energy transport is related with magnetic fluctuations and can be ameliorated in high $S$ regimes.

Stationary advanced scenarios with internal transport barrier on ASDEX Upgrade

Steady-state discharges with improved core confinement and H-mode edge with edge localized modes (ELMs) are investigated. In plasmas with an upper triangularity top close to zero an H-factor of HITER89-P = 2.7 and N = 2.2 could be maintained for 1 s and HITER89-P = 2.4 and N = 2.0 for 6 s, the latter corresponding to 40 confinement times or 2 1/2 resistive time scales for current redistribution, only limited by the duration of the possible discharge length. At a line averaged density of 4 × 1019 m-3 the central temperatures reach values of Ti = 10 keV and Te = 6.5 keV. The stationarity of the current profile is explained by magnetic reconnection driven by strong (m = 1, n = 1) fishbones, which, in the absence of sawteeth, also expel energy and impurities. Further increasing the pressure, is limited by neoclassical tearing modes. Raising the density by edge gas fuelling and the simultaneous increase of the neutral beam power, HITER89-P remained unchanged up to ne = 5.5 × 1019 m-3, accompanied by a substantial reduction of Zeff. Increasing top to 0.2, both confinement and -limit improved reaching values of HITER89-P = 3.0 and N = 2.4 at densities above ne = 5 × 1019 m-3. This resulted in the highest fusion product of nD,0Ti,0E = 0.9 × 1020 keV s m-3 so far observed in ASDEX Upgrade.

Performance near operational boundaries

The performance of ELMy H-mode operation in ASDEX Upgrade andJET is compared. Special attention is paid to variations (usuallyreductions) in this performance near the operational limits whichwill need to be approached in a next-step device. In JET it isfound that input powers substantially above the H-mode thresholdpower are required to obtain discharges with energy confinementenhancement factors at or above the usual ELMy H-mode scalings.Such a margin (as much as a factor of two in JET) is not observedin ASDEX Upgrade. It is proposed that this difference may be dueto the higher edge collisionality in ASDEX and the results arecompared to a recent theory based on interchange instabilitiesand magnetic flutter. In ASDEX Upgrade, the confinement in type IELMy discharges degrades as the density is raised due to astiffness of the temperature profiles which leads to adegradation of the core confinement. This type of stiffness isobserved in JET only at relatively high edge densities. In JET, theedge confinement degrades as the density is increased by externalgas fuelling, consistent with a constant edge pressure gradientand an edge barrier width which reduces in proportion to the edgeion poloidal Larmor radius. In both machines, H-mode performanceis limited at high density by a transition first to the type III ELMregime and then to the L-mode. The confinement penalty, relativeto good type I ELM discharges, of operating with type III ELMs isabout 25-30%. The maximum densities for operation withtype I or type III ELMs can be substantially increased byincreasing the plasma triangularity in both machines.

Core-edge studies with boundary island configurations on the W7-AS stellarator

Core, edge and scrape-off-layer plasma behaviour is studied principally under conditions of an a = 5/9 boundary island configuration - which is relevant for the upcoming W7-AS divertor campaign - but for now with ten inboard sector limiters. The major focus is on compatibility between good core confinement and attainment of high recycling at the limiter. At low input power Pin0.4 MW, operation at densities necessary to attain effective divertor action in the future invariably leads to a transition to the ELM-free H-mode accompanied by lower edge densities and increased core radiation until radiation collapse ensues. Thereby, enhancement factors in E of nearly two above the international stellarator confinement scaling are transiently achieved. The threshold density ethr, necessary to attain the H-mode increases with heating power, such that at 2 MW NBI heating power the H-mode is completely suppressed and peak densities at the limiter exceeding 1.5 × 1020 m-3 are realized. The efficacy of newly-installed control coils designed to manipulate the island geometry is tested. Their influence on the core plasma is verified. Due to geometrical effects associated with the mutual shadowing of the inboard limiters, statements regarding the influence on island physics must await the divertor configuration.

Study of a closed divertor with strong gas puffing on JFT-2M

In order to improve the particle control capability of the divertorand to demonstrate the possibility of a dense and cold divertor withhigh confinement plasmas, a closed divertor study has been carried outon the JFT-2M tokamak. When a strong gas puff is applied into thedivertor chamber, a baffling effect is revealed by enhanced radiationlocalization and high neutral pressure in the divertor chamber, lowcore fuelling and sustainment of core confinement quality. Thebaffling effect is further enhanced with E × B flow and/or current inthe scrape-off layer by applying divertor biasing. A dense and colddivertor state (nediv ≈ 4 × 1019m-3 and Tediv ≈ 4 eV), together with the improved confinement modes, can be obtained by strong gas puffing. Furthermore, the improved core confinement is not affectedsignificantly in the high density region (n̄e/neG < 0.7). TheUEDA code simulation reproduces the baffling effect and a dense andcold divertor plasma with a closed divertor structure.

Comparative studies of core and edge transport barrier dynamics of DIII-D and TFTR tokamak plasmas

The plasma dynamics of enhanced confinement regimes in the TFTR core and the DIII-D core and edge are compared in order to identify a common physics basis. Despite differences in transition timescale and location, as well as the sign of the radial electric field Er, observations suggest that E × B shear effects on turbulence induced transport play a dominant role in governing barrier dynamics in all cases. Fast confinement bifurcations are observed in the TFTR core enhanced reverse shear (ERS) regime and in the edge DIII-D H mode. Both show spontaneous Er shear layer formation prior to the confinement change and a negative Er well that persists as steep gradients form. These dynamics differ from those of DIII-D negative central shear (NCS) plasmas. There, slow transitions are observed when the applied torque from unidirectional beam injection is small, while faster development and more dramatic confinement improvements occur at higher applied torques. Unlike the H mode and ERS cases, the NCS core generally has a positive Er hill and no strong Er shear precursor. However, similarity experiments performed on TFTR indicate that ERS, L mode and NCS-like regimes can all be accessed in a continuous fashion by varying the E × B shear through changes in the applied torque at constant power. As in the DIII-D NCS case, core confinement in TFTR reverse shear plasmas improves slowly as co-rotation begins to dominate the determination of Er, no strong Er shear layer develops prior to that improvement, and the plasma possesses a positive Er hill. Reductions in transport with Er gradients of either sign are consistent with the picture of E × B shear suppression and decorrelation of turbulence. At fixed input power, intermediate levels of confinement improvement are achieved by varying the E × B shear with changes in the applied neutral beam torque. The data suggest that control over the plasma pressure profile in a reactor may be possible if an external source of E × B shear, such as might be applied with RF techniques, is used to modify the shear which otherwise occurs.

Towards steady state tokamak operation with double transport barriers

Core confinement and the overall performance of the tokamak have been considerably enhanced with the combination of two improved confinement regimes. Internal transport barriers characteristic of the optimized shear regime and an edge transport barrier of the high confinement H mode regime have been superposed in the double barrier (DB) mode. In DT discharges the DB mode has resulted in a fusion gain Q a factor of 2 higher than in conventional sawtoothing steady state ELMy H mode plasmas. The DBmode has been routinely established in the new Gas Box divertor configuration on JET. Off-axis lower hybrid current drive and edge radiation cooling through impurity seeding have been used for current and pressure profile control. This has made it possible to increase the performance up to βN = 2.4 and H89 = 2.7 and to extend flat-top phases up to a duration of three energy confinement times, approaching steady state conditions. Modelling studies of DB mode DT operation on JET predict a fusion power in the range 20-30 MW. In ITER transport code modelling, steady state operation with a fusion power output of 1.3 GW is obtained for advanced tokamak scenarios in the DB mode.

The influence of fishbones on the background plasma

Observations on ASDEX Upgrade show that fishbones can have a much strongereffect on the background plasma parameters than expected from an ideal,fast particle driven instability. They lead to a redistribution of electrontemperature and impurity densities, similar in form and magnitude tosawteeth, albeit with the plasma parameter variation occurring on asignificantly longer time-scale. Simulations of the current diffusion indischarges with favourable core confinement require fishbones to befurthermore associated also with magnetic reconnection to yield consistencybetween the conductivity and current profile information derived from ECE,bremsstrahlung, MSE and the localization of m = 1,n = 1 activity.

Approach to steady state high performance in DD and DT plasmas with optimized shear in JET

Steady state high performance with improved core confinement and sustainable plasma edge conditions has been approached on JET in a double barrier (DB) mode. The DB mode combines an internal transport barrier of the optimized shear regime with an edge transport barrier of an ELMy H mode regime. Improved confinement with an H factor HITER-89 ≈ 2 has been maintained for four energy confinement times. Ion and electron temperature profiles remain peaked in the DB mode, while the density profile is broad and similar in shape to the conventional ELMy H mode profile. The energy confinement improves across the whole plasma cross-section, and the ion heat conductivity falls to the neoclassical level in the core. Particle transport studies show that impurity accumulation can be avoided in the DB mode. In DT discharges the DB mode has attained a fusion gain of Q ≈ 0.4, producing 6.8 MW of fusion power, compared with Q ≈ 0.2 in the conventional sawtoothing ELMy H mode. The ELMs are more benign, with an amplitude an order of magnitude smaller in the DB mode. The DB mode has a high potential to improve performance in reactor relevant conditions.

ITER edge database investigations of the SOL width

Based on the ITER edge database, comparisons among the major tokamaks (ASDEX, ASDEX-Up, COMPASS-D, C-MOD, DIII-D, JET, JFT-2M and JT-60U) with respect to systematic SOL-width behavior are presented in terms of free-form regression analysis for λ n e and λ T e . Emphasis is placed on OH-L plasmas as this data is most abundant and coherent. Despite differences in size, wall-conditioning and divertor configuration, broadly similar trends are found in the regimes of low- and high-recycling and partially-detached divertor operation. In the low-recycling regime all machines appear to underlie a size scaling: λ n e (min), λ T e (min)∼const.(surface area/ I p) 0.7. Operation at high q95, high densities with respect to the Greenwald limit, or high powers leads to SOL widths considerably in excess of the values indicated by this simple scaling, and are documented in the form of machine-specific or machine-group-specific parametric expressions. For ELMy H-modes there is a general, but not universal, trend of smaller λ n e and λ T e being associated with higher core confinement. On ASDEX-Up the SOL pressure e-folding length can vary as n e I p −2 for ELMy H-modes.

Impurity enrichment and radiative enhancement using induced SOL flow in DIII-D

Experiments on DIII-D have demonstrated the efficacy of using induced scrape-off-layer (SOL) flow to preferentially enrich impurities in the divertor plasma. This SOL flow is produced through simultaneous deuterium gas injection at the midplane and divertor exhaust. Using this SOL flow, an improvement in enrichment (defined as the ratio of impurity fraction in the divertor to that in the plasma core) has been observed for all impurities in trace-level experiments (i.e., impurity level is non-perturbative), with the degree of improvement increasing with impurity atomic number. In the case of argon, exhaust gas enrichment using modest SOL flow is as high as 17. Using this induced SOL flow technique and argon injection, radiative plasmas have been produced that combine high radiation losses ( P rad/ P input > 70%), low core fuel dilution ( Z eff < 1.9), and good core confinement ( τ E > 1.0 τ E,ITER93H).

Report on the 6th European Fusion Physics Workshop

The 6th European Fusion Physics Workshop took place in December 1998 at Cadarache in France. It was hosted by the Association CEA-Euratom, Cadarache and was sponsored by the European Commission. Within an overall theme of plasma edge physics and its relation to core plasma behaviour, four topics of importance to the future development of magnetically confined plasmas were discussed in detail: the role of turbulence in edge and scrape-off layer physics, the relation between edge physics and core confinement, the influence of divertor geometry on plasma behaviour and transient and steady-state effects in plasma-wall interactions. In addition, an extended review of JET experiments on advanced modes of operation was presented. The main issues discussed in each session and the areas identified as requiring further study are summarized here.

Physics fundamentals for ITER

The design of an experimental thermonuclear reactor requires both cutting-edge technology and physics predictions precise enough to carry forward the design. The past few years of worldwide physics studies have seen great progress in understanding, innovation and integration. We will discuss this progress and the remaining issues in several key physics areas. (1) Transport and plasma confinement. A worldwide database has led to an `empirical scaling law' for tokamaks which predicts adequate confinement for the ITER fusion mission, albeit with considerable but acceptable uncertainty. The ongoing revolution in computer capabilities has given rise to new gyrofluid and gyrokinetic simulations of microphysics which may be expected in the near future to attain predictive accuracy. Important databases on H-mode characteristics and helium retention have also been assembled. (2) Divertors, heat removal and fuelling. A novel concept for heat removal - the radiative, baffled, partially detached divertor - has been designed for ITER. Extensive two-dimensional (2D) calculations have been performed and agree qualitatively with recent experiments. Preliminary studies of the interaction of this configuration with core confinement are encouraging and the success of inside pellet launch provides an attractive alternative fuelling method. (3) Macrostability. The ITER mission can be accomplished well within ideal magnetohydrodynamic (MHD) stability limits, except for internal kink modes. Comparisons with JET, as well as a theoretical model including kinetic effects, predict such sawteeth will be benign in ITER. Alternative scenarios involving delayed current penetration or off-axis current drive may be employed if required. The recent discovery of neoclassical beta limits well below ideal MHD limits poses a threat to performance. Extrapolation to reactor scale is as yet unclear. In theory such modes are controllable by current drive profile control or feedback and experiments should be forthcoming soon. Recent results on JET and TFTR have confirmed qualitative understanding of alpha particle driven toroidal Alfvén eigenmodes (TAEs). Present predictions for TAE effects in ITER are favourable, but require further work. The large stored energies in ITER have focused attention on disruption physics. Databases for thermal and current quenches, vertical displacement events (VDEs) and halo currents have enabled thermomechanical design. Some questions remain open as to the production, confinement and localization of runaway electrons in potentially unstable plasmas and mitigation strategies have been proposed. Other crucial ITER needs such as diagnostics, control and heating appear to have acceptable solutions. All this rich physics requires experimental validation by a reactor-scale plasma and care has been taken to provide sufficient flexibility for ITER to cover a wide range of scenarios.

Prospects for charge-exchange-recombination-based measurements on International Thermonuclear Experimental Reactor using a helium diagnostic neutral beam

Several important measurements in the International Thermonuclear Experimental Reactor (ITER) diagnostic mission, including the primary one of core helium ash density, are expected to be addressed using active spectroscopic techniques. These methods rely on the use of a dedicated diagnostic neutral beam (DNB) which has been optimized for the dual requirements of beam penetration and charge exchange cross section. For hydrogenic beams, this results in an optimal beam energy of ∼100 keV/amu. Signal-to-noise estimates using realistic geometries and the existing ITER profile and equilibrium data have confirmed the stringent requirements on beam quality and intensity to satisfy the stated ITER measurement precisions. In this article we consider the use of a neutral helium DNB for making active spectroscopic measurements on ITER, since helium beams offer better penetration in dense plasma for a given energy, and the prospects for given source performance may also be improved. Drawbacks include the more difficult absolute calibration of the beam density profile as well as the fundamental problem of uniquely identifying the source (fusion-based ash, beam core fuelling, or edge DNB neutralizer/source efflux) of the observed He charge-exchange recombination line in order to unambiguously characterize core helium buildup and confinement on ITER.