C. F. Maggi

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.

Plasma detachment in JET Mark I divertor experiments

The experimental characteristics of divertor detachment in the JET tokamak with the Mark?I pumped divertor are presented for ohmic, L?mode and ELMy H?mode experiments with the main emphasis on discharges with deuterium fuelling only. The range over which divertor detachment is observed for the various regimes, as well as the influence of divertor configuration, direction of the toroidal field, divertor target material and active pumping on detachment, will be described. The observed detachment characteristics, such as the existence of a considerable electron pressure drop along the field lines in the scrape-off layer (SOL), and the compatibility of the decrease in plasma flux to the divertor plate with the observed increase of neutral pressure and D? emission from the divertor region, will be examined in the light of existing results from analytical and numerical models for plasma detachment. Finally, a method to evaluate the degree of detachment and the window of detachment is proposed, and all the observations of the JET Mark?I divertor experiments are summarized in the light of this new quantitative definition of divertor detachment.

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.

Non-boronized compared with boronized operation of ASDEX Upgrade with full-tungsten plasma facing components

After completion of the tungsten coating of all plasma facing components, ASDEX Upgrade has been operated without boronization for 1 1/2 experimental campaigns. This has allowed the study of fuel retention under conditions of relatively low D co-deposition with low-Z impurities as well as the operational space of a full-tungsten device for the unfavourable condition of a relatively high intrinsic impurity level. Restrictions in operation were caused by the central accumulation of tungsten in combination with density peaking, resulting in H?L backtransitions induced by too low separatrix power flux. Most important control parameters have been found to be the central heating power, as delivered predominantly by ECRH, and the ELM frequency, most easily controlled by gas puffing. Generally, ELMs exhibit a positive impact, with the effect of impurity flushing out of the pedestal region overbalancing the ELM-induced W source. The restrictions of plasma operation in the unboronized W machine occurred predominantly under low or medium power conditions. Under medium-high power conditions, stable operation with virtually no difference between boronized and unboronized discharges was achieved. Due to the reduced intrinsic radiation with boronization and the limited power handling capability of VPS coated divertor tiles (?10?MW?m?2), boronized operation at high heating powers was possible only with radiative cooling. To enable this, a previously developed feedback system using (thermo-)electric current measurements as approximate sensor for the divertor power flux was introduced into the standard AUG operation. To avoid the problems with reduced ELM frequency due to core plasma radiation, nitrogen was selected as radiating species since its radiative characteristic peaks at lower electron temperatures in comparison with Ne and Ar, favouring SOL and divertor radiative losses. Nitrogen seeding resulted not only in the desired divertor power load reduction but also in improved energy confinement, as well as in smaller ELMs.

Impurity behaviour in the ASDEX Upgrade divertor tokamak with large area tungsten walls

At the central column of ASDEX Upgrade, an area of 5.5 m2 of graphite tiles was replaced by tungsten-coated tiles representing about two-thirds of the total area of the central column. No negative influence on the plasma performance was found, except for internal transport barrier limiter discharges. The tungsten influx ΓW stayed below the detection limit only during direct plasma wall contact or for reduced clearance in divertor discharges spectroscopic evidence for ΓW could be found. From these observations a penetration factor of the order of 1% and effective sputtering yields of about 10-3 could be derived, pointing to a strong contribution by light intrinsic impurities to the total \mbox{W-sputtering}. The tungsten concentrations ranged from below 10-6 up to a few times 10-5. Generally, in discharges with increased density peaking, a tendency for increased central tungsten concentrations or even accumulation was observed. Central heating (mostly) by ECRH led to a strong reduction of the central impurity content, accompanied by a very benign reduction of the energy confinement. The observations suggest that the W-source strength plays only an inferior role for the central W-content compared to the transport, since in the discharges with increased W-concentration neither an increase in the W-influx nor a change in the edge parameters was observed. In contrast, there is strong experimental evidence, that the central impurity concentration can be controlled externally by central heating.

ELM frequency control by continuous small pellet injection in ASDEX Upgrade

Injection of cryogenic deuterium pellets has been successfully applied in ASDEX Upgrade for external edge localized mode (ELM) frequency control in type-I ELMy H-mode discharge scenarios. A pellet velocity of 560 m s−1 and a size of about 6 × 1019 D-atoms was selected for technical reasons, although even lower masses were found sufficient to trigger ELMs. A moderate repetition rate close to 20 Hz was chosen to avoid over-fuelling of the core plasma. Pellet sequences of up to 4 s duration were injected into discharges close to the L–H threshold, intrinsically developing large compound ELMs at a rate of 3 Hz. With pellet injection, these large ELMs were completely replaced by smaller type-I ELMs at the much higher pellet frequency, accompanied by a slight increase of density and even of stored energy. This external ELM control could be repeatedly switched on and off by just interrupting the pellet train. ELMs were triggered in less than 200 µs after pellet arrival at the plasma edge, at which time only a fraction of the pellet has been ablated, forming a rather localized, three-dimensional plasmoid, which drives the edge unstable well before the deposited mass is spread toroidally. The pellet controlled case has also been compared with a discharge at a somewhat lower density, but with otherwise rather similar data, developing spontaneous 20 Hz type-I ELMs. Despite the different trigger mechanisms, the general ELM features turn out to be qualitatively similar, possibly because of the similarity of the two cases in terms of ELM relevant parameters. The scaling with background plasma, heating power, pellet launch parameters, etc over a larger range still remains to be investigated.

Tungsten transport in JET H-mode plasmas in hybrid scenario, experimental observations and modelling

The behaviour of tungsten in the core of hybrid scenario plasmas in JET with the ITER-like wall is analysed and modelled with a combination of neoclassical and gyrokinetic codes. In these discharges, good confinement conditions can be maintained only for the first 2?3?s of the high power phase. Later W accumulation is regularly observed, often accompanied by the onset of magneto-hydrodynamical activity, in particular neoclassical tearing modes (NTMs), both of which have detrimental effects on the global energy confinement. The dynamics of the accumulation process is examined, taking into consideration the concurrent evolution of the background plasma profiles, and the possible onset of NTMs. Two time slices of a representative discharge, before and during the accumulation process, are analysed with two independent methods, in order to reconstruct the W density distribution over the poloidal cross-section. The same time slices are modelled, computing both neoclassical and turbulent transport components and consistently including the impact of centrifugal effects, which can be significant in these plasmas, and strongly enhance W neoclassical transport. The modelling closely reproduces the observations and identifies inward neoclassical convection due to the density peaking of the bulk plasma in the central region as the main cause of the accumulation. The change in W neoclassical convection is directly produced by the transient behaviour of the main plasma density profile, which is hollow in the central region in the initial part of the high power phase of the discharge, but which develops a significant density peaking very close to the magnetic axis in the later phase. The analysis of a large set of discharges provides clear indications that this effect is generic in this scenario. The unfavourable impact of the onset of NTMs on the W behaviour, observed in several discharges, is suggested to be a consequence of a detrimental combination of the effects of neoclassical transport and of the appearance of an island.

H-mode threshold and confinement in helium and deuterium in ASDEX Upgrade

In 2008, experiments have been carried out in ASDEX Upgrade to compare H-mode power threshold and confinement time in helium and deuterium. A scan in magnetic field and a wide density variation indicate that the threshold power in the two gases is very similar. The density dependence of the threshold exhibits a clear minimum. Confinement in helium is about 30% lower than in deuterium, mainly due to the reduction in the ion density caused by Z = 2 in helium.

Compatibility of ITER Scenarios with full Tungsten Wall in ASDEX Upgrade

The transition of ASDEX Upgrade (AUG) from a graphite device to a full tungsten device is demonstrated with a reduction by an order of magnitude in both the carbon deposition and deuterium retention. The tungsten source is dominated by sputtering from intrinsic light impurities, and the tungsten influxes from the outboard limiters are the main source for the plasma. In H-mode discharges, central heating (neutral beams, ECRH) is used to increase turbulent outward transport avoiding tungsten accumulation. ICRH can only be used after boronization as its application otherwise results in large W influxes due to light impurities accelerated by electrical fields at the ICRH antennas. ELMs are important in reducing the inward transport of tungsten in the H-mode edge barrier and are controlled by gas puffing. Even without boronization, stationary, ITER baseline H-modes (confinement enhancement factor from ITER 98(y, 2) scaling H98 ~ 1, normalized beta ?N ~ 2), with W concentrations below 3 ? 10?5 were routinely achieved up to 1.2?MA plasma current.The compatibility of high performance improved H-modes with unboronized W wall was demonstrated, achieving H98 = 1.1 and ?N up to 2.6 at modest triangularities ? ? 0.3 as required for advanced scenarios in ITER. With boronization the light impurities and the radiated power fraction especially in the divertor were reduced and the divertor plasma was actively cooled by N2 seeding. N2 seeding does not only protect the divertor tiles but also considerably improves the performance of improved H-mode discharges. The energy confinement increased to H98-factors of 1.25 (?N ~ 2.7) and thereby exceeded the best values in a carbon-dominated AUG machine under similar conditions. Recent investigations show that this improvement is due to higher temperatures rather than to peaking of the electron density profile.Further ITER discharge scenario tests include the demonstration of ECRF assisted low voltage plasma start-up and current rise to q95 = 3 at toroidal electric fields below 0.3?V?m?1, to achieve a ITER compatible range of plasma internal inductance of 0.71?0.97. The results reported here strongly support tungsten as a first wall material solution.

Tokamak operation with high-Z plasma facing components

Plasma operation with high-Z plasma facing components is investigated with regard to sputtering, core impurity contamination and scenario restrictions. A simple model based on dimensionless quantities for fuel and high-Z ion sources and transport to describe the high-Z concentration in the plasma core is introduced. The impurity release and further transport is factorized into the sputtering yield, the relative pedestal penetration probability and a core confinement enhancement factor. Since there are quite large uncertainties, in particular, in the sputtering source and the edge transport of high-Z impurities, very different scenarios covering a wide parameter range are taken into account in order to resolve the experimental trends. Sputtering of tungsten by charge exchange neutrals in the energy range 0.5–2 keV is comparable to the effect of impurity ion sputtering, while the impact of thermal fuel ions is negligible. Fast ions produced by neutral beam injection as well as sheath acceleration during ICR heating may cause considerable high-Z sources if the limiters on the lowfield side have high-Z surfaces. The critical behaviour of the central high-Z concentration in some experimental scenarios could be attributed to edge and core transport parameters in the concentration model. The improved H-mode with off-central heating turns out to be the most critical one, since a hot edge is combined with peaked density profiles. (Some figures in this article are in colour only in the electronic version)