A. Kallenbach

Chapter 4: Power and particle control

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.

Tungsten: an option for divertor and main chamber plasma facing components in future fusion devices

The tungsten programme in ASDEX Upgrade is pursued towards a full high-Z device. The spectroscopic diagnostic of W has been extended and refined and the cooling factor of W has been re-evaluated. The W coated surfaces now represent a fraction of 65% of all plasma facing components (24.8 m(2)). The only two major components that are not yet coated are the strikepoint region of the lower divertor as well as the limiters at the low field side. While extending the W surfaces, the W concentration and the discharge behaviour have changed gradually pointing to critical issues when operating with a W wall: anomalous transport in the plasma centre should not be too low, otherwise neoclassical accumulation can occur. One very successful remedy is the addition of central RF heating at the 20-30% level. Regimes with low ELM activity show increased impurity concentration over the whole plasma radius. These discharges can be cured by increasing the ELM frequency through pellet ELM pacemaking or by higher heating power. Moderate gas puffing also mitigates the impurity influx and penetration, however, at the expense of lower confinement. The erosion yield at the low field side guard limiter can be as high as 10(-3) and fast particle losses from NBI were identified to contribute a significant part to the W sputtering. Discharges run in the upper W coated divertor do not show higher W concentrations than comparable discharges in the lower C based divertor. According to impurity transport calculations no strong high-Z accumulation is expected for the ITER standard scenario as long as the anomalous transport is at least as high as the neoclassical one.

Scaling of the tokamak near the scrape-off layer H-mode power width and implications for ITER

A multi-machine database for the H-mode scrape-off layer power fall-off length, λq in JET, DIII-D, ASDEX Upgrade, C-Mod, NSTX and MAST has been assembled under the auspices of the International Tokamak Physics Activity. Regression inside the database finds that the most important scaling parameter is the poloidal magnetic field (or equivalently the plasma current), with λq decreasing linearly with increasing Bpol. For the conventional aspect ratio tokamaks, the regression finds , yielding λq,ITER 1 mm for the baseline inductive H-mode burning plasma scenario at Ip = 15 MA. The experimental divertor target heat flux profile data, from which λq is derived, also yield a divertor power spreading factor (S) which, together with λq, allows an integral power decay length on the target to be estimated. There are no differences in the λq scaling obtained from all-metal or carbon dominated machines and the inclusion of spherical tokamaks has no significant influence on the regression parameters. Comparison of the measured λq with the values expected from a recently published heuristic drift based model shows satisfactory agreement for all tokamaks.

Divertor power load feedback with nitrogen seeding in ASDEX Upgrade

Feedback control of the divertor power load by means of nitrogen seeding has been developed into a routine operational tool in the all-tungsten clad ASDEX Upgrade tokamak. For heating powers above about 12?MW, its use has become inevitable to protect the divertor tungsten coating under boronized conditions. The use of nitrogen seeding is accompanied by improved energy confinement due to higher core plasma temperatures, which more than compensates the negative effect of plasma dilution by nitrogen on the neutron rate. This paper describes the technical details of the feedback controller. A simple model for its underlying physics allows the prediction of its behaviour and the optimization of the feedback gain coefficients used. Storage and release of nitrogen in tungsten surfaces were found to have substantial impact on the behaviour of the seeded plasma, resulting in increased nitrogen consumption with unloaded walls and a latency of nitrogen release over several discharges after its injection. Nitrogen is released from tungsten plasma facing components with moderate surface temperature in a sputtering-like process; therefore no uncontrolled excursions of the nitrogen wall release are observed. Overall, very stable operation of the high-Z tokamak is possible with nitrogen seeding, where core radiative losses are avoided due to its low atomic charge Z and a high ELM frequency is maintained.

ELM pace making and mitigation by pellet injection in ASDEX upgrade

In ASDEX Upgrade, experimental efforts aim to establish pace making and mitigation of type-I edge localized modes (ELMs) in high confinement mode (H-mode) discharges. Injection of small size cryogenic deuterium pellets (~(1.4?mm)2 ? 0.2?mm ? 2.5 ? 1019?D) at rates up to 83?Hz imposed persisting ELM control without significant fuelling, enabling for investigations well inside the type-I ELM regime. The approach turned out to meet all required operational features. ELM pace making was realized with the driving frequency ranging from 1 to 2.8 times the intrinsic ELM frequency, the upper boundary set by hardware limits. ELM frequency enhancement by pellet pace making causes much less confinement reduction than by engineering means like heating, gas bleeding or plasma shaping. Confinement reduction is observed in contrast to the typical for engineering parameters. Matched discharges showed triggered ELMs ameliorated with respect to intrinsic counterparts while their frequency was increased. No significant differences were found in the ELM dynamics with the available spatial and temporal resolution. By breaking the close correlation of ELM frequency and plasma parameters, pace making allows the establishment of fELM as a free parameter giving enhanced operational headroom for tailoring H-mode scenarios with acceptable ELMs. Use was made of the pellet pace making tool in several successful applications in different scenarios. It seems that further reduction of the pellet mass could be possible, eventually resulting in less confinement reduction as well.

Impurity seeding for tokamak power exhaust: from present devices via ITER to DEMO

A future fusion reactor is expected to have all-metal plasma facing materials (PFMs) to ensure low erosion rates, low tritium retention and stability against high neutron fluences. As a consequence, intrinsic radiation losses in the plasma edge and divertor are low in comparison to devices with carbon PFMs. To avoid localized overheating in the divertor, intrinsic low-Z and medium-Z impurities have to be inserted into the plasma to convert a major part of the power flux into radiation and to facilitate partial divertor detachment. For burning plasma conditions in ITER, which operates not far above the L–H threshold power, a high divertor radiation level will be mandatory to avoid thermal overload of divertor components. Moreover, in a prototype reactor, DEMO, a high main plasma radiation level will be required in addition for dissipation of the much higher alpha heating power. For divertor plasma conditions in present day tokamaks and in ITER, nitrogen appears most suitable regarding its radiative characteristics. If elevated main chamber radiation is desired as well, argon is the best candidate for the simultaneous enhancement of core and divertor radiation, provided sufficient divertor compression can be obtained. The parameter Psep/R, the power flux through the separatrix normalized by the major radius, is suggested as a suitable scaling (for a given electron density) for the extrapolation of present day divertor conditions to larger devices. The scaling for main chamber radiation from small to large devices has a higher, more favourable dependence of about Prad,main/R2. Krypton provides the smallest fuel dilution for DEMO conditions, but has a more centrally peaked radiation profile compared to argon. For investigation of the different effects of main chamber and divertor radiation and for optimization of their distribution, a double radiative feedback system has been implemented in ASDEX Upgrade (AUG). About half the ITER/DEMO values of Psep/R have been achieved so far, and close to DEMO values of Prad,main/R2, albeit at lower Psep/R. Further increase of this parameter may be achieved by increasing the neutral pressure or improving the divertor geometry.

Plasma?surface interaction, scrape-off layer and divertor physics: implications for ITER

Recent research in scrape-off layer (SOL) and divertor physics is reviewed; new and existing data from a variety of experiments have been used to make cross-experiment comparisons with implications for further research and ITER. Studies of the region near the separatrix have addressed the relationship of profiles to turbulence as well as the scaling of the parallel power flow. Enhanced low-field side radial transport is implicated as driving parallel flows to the inboard side. The medium-n nature of edge localized modes (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. The predicted divertor power loads for ITER disruptions are reduced while those to main chamber plasma facing components (PFCs) increase. Disruption mitigation through massive gas puffing is successful at reducing PFC heat loads. New estimates of ITER tritium retention have shown tile sides to play a significant role; tritium cleanup may be necessary every few days to weeks. ITERs use of mixed materials gives rise to a reduction of surface melting temperatures and chemical sputtering. Advances in modelling of the ITER divertor and flows have enhanced the capability to match experimental data and predict ITER performance.

Type-I ELM substructure on the divertor target plates in ASDEX Upgrade

In the ASDEX Upgrade tokamak, the power deposition structures on the divertor target plates during type-I edge localized modes (ELMs) have been investigated by infrared thermography. In addition to the axisymmetric strike line, several poloidally displaced stripes are resolved, identifying an ELM as a composite of several subevents. This pattern is interpreted as being a signature of the helical perturbations in the low field side edge during the non-linear ELM evolution. Based on this observation, the ELM related magnetic perturbation in the midplane can be derived from the target load pattern. In the start phase of an ELM collapse, average toroidal mode numbers around n ≈ 3–5 are found evolving to values of n ≈ 12–14 during the ELM power deposition maximum. Further information about the non-linear evolution of the ELM mode structure is obtained from statistical analyses of the spatial distribution, heat flux amplitudes and number of single stripes.

Assessment of compatibility of ICRF antenna operation with full W wall in ASDEX Upgrade

The compatibility of ICRF (ion cyclotron range of frequencies) antenna operation with high-Z plasma facing components is assessed in ASDEX Upgrade (AUG) with its tungsten (W) first wall.The mechanism of ICRF-related W sputtering was studied by various diagnostics including the local spectroscopic measurements of W sputtering yield YW on antenna limiters. Modification of one antenna with triangular shields, which cover the locations where long magnetic field lines pass only one out of two (0π)-phased antenna straps, did not influence the locally measured YW values markedly. In the experiments with antennas powered individually, poloidal profiles of YW on limiters of powered antennas show high YW close to the equatorial plane and at the very edge of the antenna top. The YW-profile on an unpowered antenna limiter peaks at the location projecting to the top of the powered antenna.An interpretation of the YW measurements is presented, assuming a direct link between the W sputtering and the sheath driving RF voltages deduced from parallel electric near-field (E||) calculations and this suggests a strong E|| at the antenna limiters. However, uncertainties are too large to describe the YW poloidal profiles.In order to reduce ICRF-related rise in W concentration CW, an operational approach and an approach based on calculations of parallel electric fields with new antenna designs are considered. In the operation, a noticeable reduction in YW and CW in the plasma during ICRF operation with W wall can be achieved by (a) increasing plasma–antenna clearance; (b) strong gas puffing; (c) decreasing the intrinsic light impurity content (mainly oxygen and carbon in AUG). In calculations, which take into account a realistic antenna geometry, the high E|| fields at the antenna limiters are reduced in several ways: (a) by extending the antenna box and the surrounding structures parallel to the magnetic field; (b) by increasing the average strap–box distance, e.g. by increasing the number of toroidally distributed straps; (c) by a better balance of (0π)-phased contributions to RF image currents.

Characterization of pedestal parameters and edge localized mode energy losses in the Joint European Torus and predictions for the International Thermonuclear Experimental Reactor

This paper presents the experimental characterization of pedestal parameters, edge localized mode (ELM) energy, and particle losses from the main plasma and the corresponding ELM energy fluxes on plasma facing components for a series of dedicated experiments in the Joint European Torus (JET). From these experiments, it is demonstrated that the simple hypothesis relating the peeling-ballooning linear instability to ELM energy losses is not valid. Contrary to previous observations at lower triangularities, small energy losses at low collisionality have been obtained in regimes at high plasma triangularity and q95∼4.5, indicating that the edge plasma magnetohydrodynamic stability is linked with the transport mechanisms that lead to the loss of energy by conduction during type I ELMs. Measurements of the ELM energy fluxes on the divertor target show that their time scale is linked to the ion transport along the field and the formation of a high energy sheath, in agreement with kinetic modeling of ELMs. Higher...