DIII-D National Fusion Facility

Progress, since the ITER Physics Basis publication (ITER Physics Basis Editors et al 1999 Nucl. Fusion 39 2137-2664), in understanding the processes that will determine the properties of the plasma edge and its interaction with material elements in ITER is described. Experimental areas where significant progress has taken place are energy transport in the scrape-off layer (SOL) in particular of the anomalous transport scaling, particle transport in the SOL that plays a major role in the interaction of diverted plasmas with the main-chamber material elements, edge localized mode (ELM) energy deposition on material elements and the transport mechanism for the ELM energy from the main plasma to the plasma facing components, the physics of plasma detachment and neutral dynamics including the edge density profile structure and the control of plasma particle content and He removal, the erosion of low- and high-Z materials in fusion devices, their transport to the core plasma and their migration at the plasma edge including the formation of mixed materials, the processes determining the size and location of the retention of tritium in fusion devices and methods to remove it and the processes determining the efficiency of the various fuelling methods as well as their development towards the ITER requirements. This experimental progress has been accompanied by the development of modelling tools for the physical processes at the edge plasma and plasma-materials interaction and the further validation of these models by comparing their predictions with the new experimental results. Progress in the modelling development and validation has been mostly concentrated in the following areas: refinement in the predictions for ITER with plasma edge modelling codes by inclusion of detailed geometrical features of the divertor and the introduction of physical effects, which can play a major role in determining the divertor parameters at the divertor for ITER conditions such as hydrogen radiation transport and neutral-neutral collisions, modelling of the ion orbits at the plasma edge, which can play a role in determining power deposition at the divertor target, models for plasma-materials and plasma dynamics interaction during ELMs and disruptions, models for the transport of impurities at the plasma edge to describe the core contamination by impurities and the migration of eroded materials at the edge plasma and its associated tritium retention and models for the turbulent processes that determine the anomalous transport of energy and particles across the SOL. The implications for the expected performance of the reference regimes in ITER, the operation of the ITER device and the lifetime of the plasma facing materials are discussed.

Abstract: The work of the ITPA SOL/divertor group is reviewed and implications for ITER discussed. Studies of near SOL gradients have revealed a connection to underlying turbulence models. Analysis of a multi-machine database shows that parallel conduction gradients near the separatrix scale as major radius. New SOL measurements have implicated low-field side transport as driving parallel flows to the inboard side. The high-n nature of ELMs has been elucidated and new measurements have determined that they carry ~10-20 % of the ELM energy to the far SOL with implications for ITER limiters and the upper divertor. Analysis of ELM measurements imply that the ELM continuously loses energy as it travels across the SOL – larger gaps should reduce surface loads. The predicted divertor power loads for ITER disruptions has been reduced as a result of finding that the divertor footprint broadens during the thermal quench and that the plasma can lose up to 80 % of

Author: Loarte, A. et al.; Genre: Journal Article; Issued: 2007; Keywords: 11th International Workshop on Plasma-Facing Materials and Components for Fusion Applications (PFMC-11), Greifswald, 2006-10-10 to 2006-10-12; Title: Transient Heat Loads in Current Fusion Experiments, Extrapolation to ITER and Consequences for its Operation

In this paper, a linear model for plasma current, position and shape control based on the plasma rigid motion assumption is presented and implemented in an EAST tokamak simulator. The simulator models the plasma, poloidal field (PF) coils, and power supplies, and is used to verify the control algorithm and optimize control parameters and PF coil current trajectories. Plasma position and shape control has been achieved during the last several EAST operation campaigns due to successful decoupling of plasma current, plasma position and shape. The control logic used and experimental results are described in detail. Diverted plasma shapes, including double null, upper and lower single null, and with elongation up to 2.0, triangularity in the range 0.4–0.6 and X point control accuracy of 1 cm, were successfully controlled. Smooth shape transition in the current ramp-up ensures that volt–seconds are saved and that plasma disruptions are avoided. Such control capability provides the basis for future high performance plasma operation.

Abstract To achieve low deposited power flux density to solid surfaces in magnetic fusion devices, very small values of α are required, where α is the angle between B and the surface tangent. For an oblique magnetic field, there exists in front of the solid surface a Chodura sheath (CS) (also known as the ‘magnetic pre-sheath’) of thickness several ρ i , the ion Larmor radius. The standard assumption is that the CS is additional to the Debye sheath (DS) of thickness several λ D , the Debye length. Simple fluid modelling for collisionless CS conditions gives the drop in normalized electrostatic potential across the CS as e Δ φ CS / kT e = ln(sin α ). For an electrically floating wall there is the separate constraint of ambipolar flow to the wall e Δ φ floating / kT e = 0.5 ln[(2 π m e / m i )(1 + T i / T e )], where Δ φ floating = Δ φ CS + Δ φ DS . For the case of a deuterium plasma and T i = T e , | e Δ φ floating / kT e | = 2.84. For α < 3.35°, | e Δ φ CS / kT e | exceeds 2.84 which evidently implies that the DS ceases to exist for such values of α and the entire potential drop would then occur across the CS. New analysis of the CS provides solutions for a number of quantities of practical importance, which improve on the solutions presently in use in models and edge impurity codes. Compared with the latter, the results of the present analysis indicate that (i) the E -field directed towards the solid surface is stronger and (ii) the plasma density drops more rapidly approaching the solid surface. The effect of (i) is to increase the probability of prompt local deposition of sputtered particles, while (ii) has the opposite effect.

Results from a series of dedicated experiments measuring the effect of particle and energy pulses from Type-I edge localized modes (ELMs) in the DIII-D scrape-off layer (SOL) and divertor are compared with a simple model of ELM propagation in the boundary plasma. The simple model asserts that the propagation of ELM particle and energy perturbations is dominated by ion parallel convection along SOL field lines and the recovery from the ELM perturbation is determined by recycling physics. Timescales associated with the initial changes of boundary plasma parameters are expected to be on the order of the ion transit time from the outer midplane, where the ELM instability is initiated, to the divertor targets. To test the model, the ion convection velocity is changed in the experiment by varying the plasma density. At moderate to high density, ne/nGr = 0.5–0.8, the delays in the response of the boundary plasma to the midplane ELM pulses, the density dependence of those delays and other observations are consistent with the model. However, at the lowest densities, ne/nGr∼0.35, small delays between the responses in the two divertors, and changes in the response of the pedestal thermal energy to ELM events, indicate that additional factors including electron conduction in the SOL, the pre-ELM condition of the divertor plasma, and the ratio of ELM instability duration to SOL transit time, may be playing a role. The results show that understanding the response of the SOL and divertor plasmas to ELMs, for various pre-ELM conditions, is just as important for predicting the effect of ELM pulses on the target surfaces of future devices as is predicting the characteristics of the ELM perturbation of the core plasma.

Scaling to larger tokamaks of high confinement plasmas with radiating edges, induced by impurities, is being studied through internationally collaborative experiments on JET. In campaigns till the end of 2000, three different regimes have been explored. A small number of limiter L-mode discharges seeded with neon have most closely repeated the approach used on TEXTOR-94, but different collisionality and particle transport in JET impede central peaking of the density associated with improved confinement. Divertor L-modes at intermediate density, again with neon injection, have pursued transiently enhanced states found on DIII-D. Confinement up to H-mode quality, together with radiation fractions of approximate to40%, have briefly been obtained, though central Z(eff) quickly increases. Most effectively, neon and argon seeding of higher density ELMy H-modes formed mainly at low triangularity on the septum of the MkIIGB divertor, resembling a pumped-limiter arrangement, have been examined. Good confinement has been sustained at densities close to the Greenwald level in 'afterpuff' (AP) phases following the end of main gas fuelling, for little change of central Z(eff) but up to approximate to60% radiation. Outstanding normalized properties up to H-97 = 0.99 at f(Gwd) = 0.94 have thus been achieved, above the conventional H-mode density limit for diverted plasmas. Stationarity of states has also been extended to many energy confinement times by including low, extra gas inputs in the 'AP', suggestive of an optimized fuelling scheme. Further development in 2001 is reported separately in [1]. Accompanying ELMs are generally reduced in frequency though not evidently in size, electron pedestal pressure being almost unchanged from unseeded behaviour. There are indications of the most favourable impurity species scaling with plasma parameters, performance, radiation and its concentration within a mantle all increasing with argon compared to neon in JET. These benefits in terms of integrated properties are just as required for long burning pulses in ITER, supporting its use of a radiating mantle to assist not only power exhaust but performance too. Impurity-seeded H-modes can therefore contribute directly to next-step scenario development.

Recent progress towards obtaining high density and high confinement in JET as required for the ITER reference scenario at Q = 10 is summarized. Plasmas with simultaneous confinement H-98(y.2) = 1 and densities up to n/n(Gw) similar to 1 are now routinely obtained. This has been possible (i) by using plasmas at high (delta similar to 0.5) and medium (delta similar to 0.3-0.4) triangularity with sufficient heating power to maintain Type I ELMs, (ii) with impurity seeded plasmas at high (delta similar to 0.5) and low (delta less than or equal to 0.2) triangularity, (iii) with an optimized pellet injection sequence, maintaining the energy confinement and raising the density, and (iv) by carefully tuning the gas puff rate leading to plasmas with peaked density profiles and good confinement at long time scales. These high performance discharges exhibit Type I ELMs, with a new and more favourable behaviour observed at high densities, requiring further studies. Techniques for a possible mitigation of these ELMs are discussed, and first promising results are obtained with impurity seeding in discharges at high triangularity. Scaling studies using the new data of this year show a strong dependence of confinement on upper triangularity, density and proximity to the Greenwald limit. Observed MHD instabilities and methods to avoid these in high density and high confinement plasmas are discussed.

An overview is given of recent advances toward the realization of high density, high confinement plasmas with radiating mantles in limiter and divertor tokamaks worldwide. Radiatively improved mode discharges on the Torus Experiment for Technology Oriented Research 94 (TEXTOR-94) [Proceedings of the 16th IEEE Symposium on Fusion Engineering, 1995 (Institute for Electrical and Electronics Engineers, Piscataway, NJ, 1995), p. 470] have recently been obtained at trans-Greenwald densities (up to n̄/nGW=1.4) with high confinement mode free of edge localized modes (ELM-free H-mode) confinement quality. Experiments in DIII-D [J. Luxon et al., Proceedings of the 11th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Association, Vienna, 1987), Vol. 1, p. 159] divertor plasmas with a low confinement mode edge have confirmed the dramatic changes in confinement observed with impurity seeding on TEXTOR-94. Recent experiment with impurity seeding on the Joint European Torus [Rebut et al., Fusion Eng. Des. 22, 7 (1993)], and the Japanese Atomic Energy Research Institute Tokamak 60 Upgrade [Horiike et al., Fusion Eng. Des. 16, 285 (1991); Hosogane et al., Proceedings of the 16th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion Research Montreal, 1987 (International Atomic Energy Association, Vienna, 1987), Vol. 3, p. 555] have extended high confinement in ELMy H-mode plasmas to higher densities. Finally, recent progress in the physical understanding of radiating mantle plasmas is summarized.

Dumortier, P.; Andrew, P.; Bonheure, G.; Budny, I. L.; Buttery, R.; Charlet, M.; Coffey, I. H.; de Baar, M.; de Vries, P. C.; Eich, T.; Hillis, D.; Ingesson, L. C.; Liu, C.; Maddison, G.; Jachmich, S.; Jackson, G.; Kallenbach, A.; Koslowski, H. R.; Lawson, K. D.; Messiaen, A. M.; Monier-Garbet, P.; Murakami, M.; Nave, M. F. F.; Ongena, J.; Parail, V.; Piuatti, M. E.; Rapp, J.; Sartori, F.; Stamp, M.; Strachan, J.; Suttrop, W.; Telesca, V. M.; Tokar, A. D.; Unterberg, B.; Valisa, M.; von Hellermann, M.; Weyssow, B.

ELMy H-Mode experiments at JET since 2000 have focussed on the steady state and simultaneous realization of the ITER QDT=10 requirements in the normalized parameters for density, confinement and beta. Steady state phases (~ 6s or ~ 15τ<sub>E</sub>) in discharges satisfying these requirements have been obtained by (i) increasing the triangularity to the ITER reference value (δ ~ 0.5) and in plasmas at low δ ~ 0.2 by seeding of Ar. Impurity seeding in high delta discharges increases the radiation level to that needed for ITER, and further increases density and confinement of unseeded reference discharges. An optimised HFS pellet injection sequence is another means to increase density and confinement. Density peaking, which would increase further ITER's performance, has been obtained with pellet injection, impurity seeding in low delta discharges and in unseeded ELMy H-Mode discharges with carefully tuned gas fuelling. Promising evidence for a reduction of the heat load caused by ELMs in high density discharges, further enhanced by impurity seeding, will be discussed. Destabilization of NTMs can limit the plasma performance and methods to avoid these will be summarized.

Termination of the runaway electron current generated during plasma disruptions is found in JT-60U during simulated vertical plasma displacement events where the safety factor at the plasma surface qs decreases. For all discharges with runaway electron generation, the runaway current disappears for qs⩾2 with the appearance of spikes in the magnetic fluctuations. The growth rate of the spikes in the magnetic fluctuations decreases by an order of magnitude during the termination of runaway current. Corresponding to the loss of runaway electrons by magnetic fluctuations, heat flux pulses are measured at the inner divertor plates, which indicates interaction of the wall with the runaway electrons. The halo current during runaway termination is small and increases after runaway termination with a dominant toroidal mode of n = 1.

A technique of fitting a modified hyperbolic tangent to the edge profiles has improved the localization of plasma edge parameters. Non-dimensional edge parameters are broadly consistent with several theories of the L-H transition that use edge gradients in their formulation of a critical threshold parameter. The ion ∇ B drift direction has only a small effect on the edge plasma conditions measured near the plasma midplane but a large effect on the divertor plasma. The dramatic change of power threshold with the ion ∇ B drift direction implies that phenomena in the divertor region may be critical for the L-H transition.

Improving confinement and β limits simultaneously in long pulse ELMy H mode discharges is investigated. The product βN H98y serves as a useful figure of merit for performance, where βN ≡ β/(I/aB) and H98y is the ratio of the thermal confinement time to the most recent ELMy H mode confinement scaling established by the ITER confinement database working group. In discharges with q0 ≈ 1 (no sawteeth) and discharges with qmin > 1.5 and negative central magnetic shear, βN ≈ 2.9 and H98y ≈ 1.4 are sustained for up to 2 s. Although peaked profiles are observed, steep internal transport barriers are not present. Further increases in βN in these discharges are limited by neoclassical tearing modes (NTMs) in the positive shear region. In another recently developed regime, βN ≈ 3.8 and H98y ≈ 2 have been sustained during large infrequent ELMs in non-sawtoothing discharges with q0 ≈ 1. This level of performance is similar to that obtained in ELM free regimes such as VH mode. The limitation on βN and pulse length in these discharges is also the onset of NTMs.

In the DIII-D electron heating and current drive installation, up to six gyrotron microwave generators in the 1-MW class at pulse lengths up to 5 s have been operated simultaneously. The frequency for all the gyrotrons is 110 GHz, corresponding to the second harmonic of the electron gyrofrequency at 2 T. The peak generated power has been >4 MW with peak injected power slightly greater than 3 MW. The radio frequency (rf) generators are located remotely and are connected to the tokamak by up to 100 m of evacuated circular corrugated waveguide carrying the HE1,1 mode with overall transmission efficiency, including coupling to the waveguide, of up to 75%. Ancillary equipment for polarization control, beam switching, power monitoring, control of launch direction, and system protection has been developed.The system has been used to support a wide variety of physics experiments, including control of magnetohydrodynamic modes, current density profile modifications, basic plasma heating and current drive, transport studies, and rf-assisted start-up. The gyrotron complex is being upgraded by the acquisition of additional tubes with 5- to 10-s pulse length capability.

Recent progress towards an increased understanding of the physical processes in the divertor and scrape-off layer (SOL) plasmas in DIII-D has been made possible by a combination of new diagnostics, improved computational models and changes in divertor geometry. The work focused primarily on ELMing H mode discharges. The physics of partially detached divertor plasmas, with divertor heat flux reduction by divertor radiation enhancement using D2 puffing, was studied in two dimensions, and a model of the heat and particle transport was developed that includes conduction, convection, ionization, recombination and flows. Plasma and impurity particle flows were measured with Mach probes and spectroscopy and compared with the UEDGE model. The model now includes self-consistent calculations of carbon impurities. Impurity radiation was increased in the divertor and SOL with `puff and pump' techniques using SOL D2 puffing, divertor cryopumping and argon puffing. The important physical processes in plasma-wall interactions were examined with a DiMES (divertor material evaluation system) probe, plasma characterization near the divertor plate and the REDEP code. Experiments comparing single null plasma operation in baffled and open divertors demonstrated a change in the edge plasma profiles. These results are consistent with a reduction in the core ionization source calculated with UEDGE. Divertor particle control in ELMing H mode with pumping and baffling resulted in a reduction in H mode core densities to ne/nGr ≈ 0.25 (with nGr the Greenwald density). Divertor particle exhaust and heat flux were studied as the plasma shape was varied from a lower single null to a balanced double null, and finally to an upper single null.

The results of recent experimental and theoretical studies concerning the effects of plasma shape and current and pressure profiles on edge instabilities in DIII-D are presented. Magnetic oscillations with toroidal mode number n ≈ 2-9 and a fast growth time γ-1 = 20-150 μs are often observed prior to the first giant type I ELM in discharges with moderate squareness. High n ideal second ballooning stability access encourages edge instabilities by facilitating the buildup of the edge pressure gradient and bootstrap current density, which destabilize the intermediate to low n modes. Analysis suggests that discharges with large edge pressure gradient and bootstrap current density are more unstable to n > 1 modes. Calculations and experimental results show that ELM amplitude and frequency can be varied by controlling access to the second ballooning stability regime at the edge through variation of the squareness of the discharge shape. A new method is proposed to control edge instabilities by reducing access to the second ballooning stability regime at the edge using high order local perturbation of the plasma shape in the outboard bad curvature region.

Impurity seeding in both the Joint European Torus (JET) and DIII-D tokamaks has produced L-mode discharges with confinement enhancements comparable to H-mode and a near doubling of the core ion temperature when compared,to similar unseeded discharges. Although Z(eff) increases with the neon injection, the total neutron rate is as high, or higher, than reference discharges and the calculated thermal neutron rate increases dramatically in both devices. Modelling with the gyrokinetic simulation code shows a reduction in low k turbulence growth rates with neon injection decreasing to less than the E x B shearing rate, consistent with stabilization of ion temperature gradient modes in both JET and DIII-D. Reductions in ion thermal diffusivity are also observed with impurity seeding. Neoclassical m/n = 3/2 tearing modes limit the duration of best performance in DIII-D with neon injection, while n = 1 and n = 2 magnetohydrodynamic modes limit the performance in JET.

ITER will explore a plasma parameter envelope currently not available in tokamaks. This will require a set of diagnostics that can follow this envelope. To implement these diagnostics in a reliable and robust way requires development of current techniques in many areas to make them applicable to ITER: they need to be operable in the ITER environment and satisfy the physics and engineering requirements. In some cases, the exploitation of new techniques will be required. While much work has been carried out in this area, significant further work remains to bring the system to implementation.

Two approaches to achieving long timescale stabilization of the ideal kink mode with a real, finite conductivity wall are considered: plasma rotation and active feedback control. DIII-D experiments have demonstrated stabilization of the resistive wall mode (RWM) by sustaining β greater than the no-wall limit for up to 200 ms, much longer than the wall penetration time of a few milliseconds. These plasmas are typically terminated by an m = 3, n = 1 mode as the plasma rotation slows below a few kilohertz. Recent temperature profile data show an ideal MHD mode structure, as expected for the RWM at β above the no-wall limit. The critical rotation rate for stabilization is in qualitative agreement with recent theories for dissipative stabilization in the absence of magnetic islands. However, drag by small amplitude RWMs or damping of stable RWMs may contribute to an observed slowing of rotation at high β, rendering rotational stabilization more difficult. An initial open loop active control experiment, using non-axisymmetric external coils and a new array of saddle loop detectors, has yielded encouraging results indicating a delayed onset of the RWM.