Overview of ASDEX Upgrade results

Transcription

1 INSTITUTE OF PHYSICS PUBLISHING and INTERNATIONAL ATOMIC ENERGY AGENCY NUCLEAR FUSION Nucl. Fusion () 7 8 PII: S9-()68- Overview of ASDEX Upgrade results H. Zohm, C. Angioni, R. Arslanbekov, C. Atanasiu, G. Becker, W. Becker, K. Behler, K. Behringer, A. Bergmann, R. Bilato, V. Bobkov, D. Bolshukhin, T. Bolzonella, K. Borrass, M. Brambilla, F. Braun, A. Buhler, A. Carlson, G.D. Conway, D.P. Coster, R. Drube,R.Dux, S. Egorov, T. Eich, K. Engelhardt, H.-U. Fahrbach, U. Fantz, H. Faugel, K.H. Finken 6, M. Foley 7, P. Franzen, J.C. Fuchs, J. Gafert, K.B. Fournier 8, G. Gantenbein 9, O. Gehre, A. Geier, J. Gernhardt, T. Goodman, O. Gruber, A. Gude,S.Günter, G. Haas, D. Hartmann, B. Heger, B. Heinemann, A. Herrmann, J. Hobirk, F. Hofmeister, H. Hohenöcker, L.D. Horton, V. Igochine, A. Jacchia, M. Jakobi,F.Jenko, A. Kallenbach, O. Kardaun, M. Kaufmann, A. Keller, A. Kendl, J.-W. Kim,K.Kirov, R. Kochergov, H. Kollotzek, W. Kraus, K. Krieger, T. Kurki-Suonio, B. Kurzan, P.T. Lang, C. Lasnier 8, P. Lauber, M. Laux, A.W. Leonard, F. Leuterer, A. Lohs, A. Lorenz, R. Lorenzini, C. Maggi, H. Maier, K. Mank, M.-E. Manso, P. Mantica, M. Maraschek, E. Martines, K.-F. Mast, P. McCarthy 7, D. Meisel, H. Meister,F.Meo, P. Merkel, R. Merkel, D. Merkl, V. Mertens, F. Monaco,A.Mück, H.W. Müller,M.Münich, H. Murmann, Y.-S. Na,G.Neu,R.Neu, J. Neuhauser, F. Nguyen, D. Nishijima, Y. Nishimura, J.-M. Noterdaeme, I. Nunes, G. Pautasso, A.G. Peeters, G. Pereverzev, S.D. Pinches,E.Poli, M. Proschek 6, R. Pugno, E. Quigley 7, G. Raupp, M. Reich, T. Ribeiro, R. Riedl, V. Rohde, J. Roth, F. Ryter, S. Saarelma, W. Sandmann, A. Savtchkov 6, O. Sauter, S. Schade, H.-B. Schilling, W. Schneider, G. Schramm, E. Schwarz, J. Schweinzer, S. Schweizer, B.D. Scott, U. Seidel, F. Serra, S. Sesnic, C. Sihler, A. Silva, A.C.C. Sips, E. Speth,A.Stäbler, K.-H. Steuer, J. Stober, B. Streibl, E. Strumberger, W. Suttrop, A. Tabasso, A. Tanga, G. Tardini, C. Tichmann, W. Treutterer, M. Troppmann, H. Urano, P. Varela, O. Vollmer, D. Wagner, U. Wenzel, F. Wesner, E. Westerhof 7,R.Wolf, E. Wolfrum,E.Würsching, S.-W. Yoon,Q.Yu, D. Zasche, T. Zehetbauer and H.-P. Zehrfeld Max-Planck-Institut für Plasmaphysik, EURATOM Association, Garching, Germany Institute of Atomic Physics, EURATOM Association, Romania Consorzio RFX, EURATOM Association, Padova, Italy Plasma Physics Department, Technical University, St Petersburg, CIS, Russia University of Augsburg, Germany 6 Institut für Plasmaphysik, EURATOM Association, FZ Jülich, Germany 7 Physics Department, University College Cork, Association EURATOM-DCU, UK 8 Lawrence Livermore National Laboratory, Livermore, USA 9 Institut für Plasmaforschung, Stuttgart University, Germany CRPP, Ecole Polytechnique Federale de Lausanne, EURATOM Association, Switzerland IFP Milano, EURATOM Association, Italy Helsinki University of Technology, EURATOM Association, Finland General Atomics, San Diego, USA Centro de Fusao Nuclear, IST Lisbon, EURATOM Association, Portugal CEA Cadarache, EURATOM Association, France 6 Institute for Applied Physics, EURATOM Association, Vienna, Austria 7 FOM Rijnhuizen, EURATOM Association, The Netherlands zohm@ipp.mpg.de Received October, accepted for publication May Published December Online at stacks.iop.org/nf//7 9-//7+$. IAEA, Vienna Printed in the UK 7

2 Overview of ASDEX Upgrade results Abstract Recent results from the ASDEX Upgrade experimental campaigns and are presented. An improved understanding of energy and particle transport emerges in terms of a critical gradient model for the temperature gradients. Coupling this to particle diffusion explains most of the observed behaviour of the density profiles, in particular, the finding that strong central heating reduces the tendency for density profile peaking. Internal transport barriers (ITBs) with electron and ion temperatures in excess of kev (but not simultaneously) have been achieved. By shaping the plasma, a regime with small type II edge localized modes (ELMs) has been established. Here, the maximum power deposited on the target plates was greatly reduced at constant average power. Also, an increase of the ELM frequency by injection of shallow pellets was demonstrated. ELM free operation is possible in the quiescent H-mode regime previously found in DIII-D which has also been established on ASDEX Upgrade. Regarding stability, a regime with benign neoclassical tearing modes (NTMs) was found. During electron cyclotron current drive (ECCD) stabilization of NTMs, β N could be increased well above the usual onset level without a reappearance of the NTM. Electron cyclotron resonance heating and ECCD have also been used to control the sawtooth repetition frequency at a moderate fraction of the total heating power. The inner wall of the ASDEX Upgrade vessel has increasingly been covered with tungsten without causing detrimental effects on the plasma performance. Regarding scenario integration, a scenario with a large fraction of noninductively driven current ( %), but without ITB has been established. It combines improved confinement (τ E /τ ITER98.) and stability (β N.) at high Greenwald fraction (n e /n GW.8) in steady state and with type II ELMy edge and would offer the possibility for long pulses with high fusion power at reduced current in ITER. PACS numbers:..fa,..rk,..tn,..wq. Introduction The ASDEX Upgrade tokamak programme [] focuses on the preparation of ITER. This includes the investigation of the physics of the baseline scenario, the edge localized mode (ELMy) H-mode, as well as the preparation of advanced scenarii with performance exceeding that of the ELMy H-mode and a major fraction of noninductively driven current. The three underlying physics areas studied are transport of energy and particles, operational boundaries due to MHD instabilities and exhaust of energy and particles by a poloidal divertor. In addition, a number of technological issues, mostly related to the heating, fuelling and exhaust systems are resolved. In particular, a strong emphasis is laid on the study of W as a wall material. In this paper, we summarize the progress made with these studies in the and experimental campaigns. The results described confirm the viability of the conventional scenario for ITER. The progress made with advanced scenarii shows that also a significant improvement of performance with respect to the standard scenario may be possible in ITER.. Upgrade of technical systems In and, several upgrades of the ASDEX Upgrade tokamak and its auxiliary systems have been completed. A major change was the redirection of one of the two neutral beam injection (NBI) boxes to provide off-axis deposition with fairly tangential beams []. Figure (a) shows the geometry. With MW at 9 kev, it was possible to show that a significant current could be driven by NBI. Figure (b) shows a comparison of two discharges: in #88, off-axis beams were used whereas in #887, on-axis beams were applied. Both discharges have similar T e, n e and Z eff profiles, so that the electrical resistance is equal. However, with off-axis NBI, the consumption of OH flux is reduced, as can be seen by the difference in OH current ramp needed to sustain the plasma current. While the magnitude of the driven current is in reasonable agreement with the theoretically expected value, the expected change of the current profile is not confirmed by the MSE measurements. Such an effect could be due to an anomalous radial diffusion of the fast NBI particles. Further studies will have to clarify the precise nature of this effect. After a refurbishment of the ion cyclotron resonance heating (ICRH) generator tubes and with the use of the db couplers that minimize the effect of reflected ICRH power onto the generators, reliable ICRH operation even in type I ELMy H-mode was achieved with power levels up to 7. MW (9% of the generator power) at the antennae. Ion cyclotron resonance frequency (ICRF) power proved to be a viable tool to study transport physics since it complements NBI heating by providing central heating without particle fuelling or direct momentum input (see section ). Also, we have validated ion cyclotron heating at the resonance frequency using a two-species (D and H) plasma in which a fast wave excites an ion Bernstein wave in the centre by mode conversion []. The outer divertor leg of the ASDEX Upgrade divertor was rebuilt to accommodate all triangularities with good pumping and power handling. This opened up the operational space at higher triangularity δ up to δ =., which had a positive impact on confinement and high densities as well as power exhaust with small (type II) ELMs (see section ). Also, more and more tiles on the inner heat shield were covered successively with a W-coating in order to advance towards a C-free interior of the machine (see section ). Improvements in the tokamak power supply led to an increase in pulse length; now, s I p flattop are possible, exceeding the current redistribution time and therefore providing the possibility to study steady state issues even on this timescale. In the coming campaign, even higher δ.6 will be possible. 7

3 H. Zohm et al 6 kv 9 kv CD #88: off-axis NBI #887: on-axis NBI [MA].8 [MW] [ka] I p P NBI same stored energy I OH - Time (s) Figure. Geometry of the ASDEX Upgrade NBI system (left) and comparison of two discharges with on-axis and off-axis NBI. The case with off-axis NBI has lower OH flux consumption, proving the higher NBI driven current in this case. T e [kev]. ECH ECH. /. 6 / 9. / 79. /.... ECH ECH ρ tor χ PB / e /Te or χ HP / e /Te [m / /(s kev )] PB Exp. HP PB HP Model κ =. qλ =. I p = 8 ka 6 gradt e /T e [m - ] Figure. T e profiles from a series of experiments with constant heating power, but varying power fraction between two different locations (left). The variation of local heat flux provides a series of points for analysis of χ e from power balance (PB) and from modulation experiments (HP) well described by the simple model from equation (). The analysis is done at ρ =... Transport studies.. Energy transport in conventional scenarii A major finding in the last years was the observation of stiff T e and T i profiles, i.e. a fixed ratio between central temperature and edge temperature. This was explained by critical gradient models, that predict the onset of turbulence (ITG, ETG, TEM) below a critical value of the temperature gradient length L T = T/ T []. Using the flexible ASDEX Upgrade electron cyclotron resonance heating (ECRH) system, it was verified that an electron heat conductivity of the generic form ( ( ) ) χ e = χ e + qλte / L Te L Te crit ( ( ) ) H () L Te L Te crit where H is the Heaviside function, q the local safety factor and λ a free parameter, explains the experimental observation very well. This heuristic model is based on the idea that above a critical gradient, transport increases strongly. Also, a gyro- Bohm dependence of anomalous transport was assumed. The ad hoc assumption of a q-dependence is needed to explain the experimentally observed dependence on plasma current as well as to reproduce the radial variation of transport. The T / -dependence was found to show good agreement with the experimental data by varying T at constant q []. This model also resolves the issue of different values of χ e from power balance and heat pulse analysis. Figure shows the results from an experiment where the location of the ECRH deposition was varied over the minor radius, keeping the total power constant. This allows us to vary the local heat flux by up to an order of magnitude with unchanged edge T e. In these experiments, the ECRH power was also modulated to infer the heat pulse χ e. The parameters (/L Te ) crit and λ in equation () were fitted to the power balance data (χ e is very small and set to zero in the fit). This set of parameters also describes very well the results for the heat pulse χ e,hp which can be derived 7

4 Overview of ASDEX Upgrade results from equation () by χ e,hp = χ e + ( χ e / T e ) T e [6]. It should be noted that the factor q in equation () leads to a good agreement with the experiment for fixed λ even when I p is varied. These experiments demonstrate that the T e profiles are moderately stiff: for finite central heating, the experimentally measured /L Te is in general above (/L Te ) crit, i.e. /L Te is usually situated somewhat above the marginal point for turbulence onset [7], in good agreement with the recent modelling using the Weiland ITG/TEM model [8]. As shown in figure, in the off-axis case the T e profile is very close to the threshold in the inner part. This allows us to measure directly the threshold which is in good agreement with that derived for TEM whereas that for ETG is higher by a factor of about... Particle transport Since the temperature profiles in general exhibit the stiff behaviour described above, a given pedestal temperature implies a certain central temperature. However, n profiles are in general not stiff and thus density peaking is a way to improve confinement and has recently attracted attention. It was shown that on ASDEX Upgrade, density peaking occurs on a timescale much longer than the energy confinement time. This can well be described by an inward pinch of the order of the neoclassical Ware pinch in combination with a fixed ratio of the particle diffusion coefficient D to the heat conductivity χ [9]. In fact, this model also explains the experimental finding that the density peaking observed with off-axis heating vanishes when the heating power is applied in the centre. Figure shows a comparison of two discharges where the ICRH deposition is varied and a reference NBI discharge. As can be seen, central deposition leads to a much flatter n e profile, consistent with the interpretation that the central heating power drives the turbulence stronger there (as is expected from the critical gradient model) and thus, through D χ, also the particle diffusion increases and dominates over the inward pinch. This mechanism may also explain the so-called ECRH-pumpout, i.e. strongly increased particle transport in the presence of ECRH []. Central heating by ICRH and ECRH has also successfully been applied to avoid impurity accumulation in the centre 9 ( m ) n e q ICRH on-axis ICRH off-axis NBI (linear scale) ρ tor [, ]: under circumstances with reduced central heating, impurity accumulation can be observed in ASDEX Upgrade discharges. This can be explained by neoclassical particle transport since a peaked main ion profile can lead to much stronger peaked impurity profiles. A quantitative analysis shows good agreement between neoclassical predictions and experimental results for ASDEX Upgrade []. If, under these circumstances, some central heating is applied, impurity accumulation can be completely suppressed. For example, the W-concentration in the centre can be reduced by two orders of magnitude with only.8 MW of ECRH in a discharge heated with MW of NBI... H-mode barrier and pedestal physics The transport studies described above give a quantitative understanding of the temperature and density profiles for given pedestal values. However, the understanding of the H-mode pedestal itself is less developed. Therefore, a major effort was put into precisely determining of the pedestal parameters, including measurements of T e and n e with a special Thomson scattering geometry that gives very high resolution in the edge. It had been found before that p in the pedestal region is roughly consistent with the ideal ballooning limit. However, this is a limit to p, but does not discriminate between T and n. Recently, we have found that T e and n e profiles in the pedestal region are closely coupled by the relation η e = d(log T e )/d(log n e ) =, i.e. T e = c n e where c is spatially constant [], but varies with plasma parameters such as I p,b t or the neutral density, etc []. Figure shows this relation for a set of ASDEX Upgrade discharges. For known c, this relation, together with the ballooning constraint, would determine the profiles in the pedestal region. Concerning the L H transition, a series of experiments was conducted in ASDEX Upgrade and JET in order to determine whether the L H power threshold is invariant if expressed in dimensionless variables. This was clearly confirmed, ruling out a dominant role of atomic physics, which would introduce a different dependence in the physics of the L H transition [6]. Finally, during a campaign with reversed I p (i.e. counter NBI), it was possible to reproduce the completely ELM-free T e (ev) q ICRH on-axis ICRH off-axis NBI (logarithmic scale) ρ tor Figure. T e and n e profiles for discharges heated with on-axis ICRH, off-axis ICRH and NBI with off-axis deposition. Since T e profiles are similar, the centrally heated case has higher central χ e and thus also higher D, leading to enhanced central particle transport. 7

5 H. Zohm et al T e (ev) η P,max 9 n (m- e ) 9 9 n (m- e,sep ) Figure. Relation between edge T e and n e showing a relationship T e n e (left, points of same colours relate to individual shots), i.e. η e = as shown in the right figure. quiescent H-mode (QH-mode) regime previously found in the DIII-D tokamak [7]. Crucial elements to access the regime were counter NBI, low density and a high clearance (i.e. large plasma-wall distance) equilibrium. Figure shows a typical example with an ELM-free period of more than s in steady state [8]. Consistent with the DIII-D results, an edge harmonic oscillation (EHO) was observed instead of the ELM activity. Details on this mode can be found in section.. However, due to the very low n e needed to access this regime, Z eff was quite high in spite of the addition of ICRH which prevented accumulation of impurities. Further experimental work has to clarify the relevance of this regime to reactor-type conditions... Internal transport barriers Recent studies on transport barriers have led to record values of T e and T i in ASDEX Upgrade. Based on existing scenarii [], a major focus was on establishing reliable operational scenarii to reproduce internal transport barriers (ITBs). Concerning ion ITBs, we changed the scenario from an inner wall or upper single null scenario to the more usual lower single null operation. The previous scenarii had been chosen because under these conditions, the discharge stayed in L-mode, which was necessary to avoid disruptions due to external kinks that occurred when an ITB discharge transited to H-mode. In the recent campaign, it was found that ITBs could be produced with an H-mode edge without disruptions when the onset of the NBI heating was slightly delayed. A detailed analysis shows that under these circumstances, the target density was lower, which seems to be a decisive parameter for ITB formation in ASDEX Upgrade [9]. Also, the q-profile has more time to relax and may be less inverted than previously; this would be a possible explanation for the absence of the external kink in the new scenario. Unfortunately, due to technical problems, no MSE measurements were available in the recent campaign to validate this hypothesis. In ITB discharges with H-mode edge, the ITB phase is accompanied by an ELM-free phase. In this phase, the stored energy increases almost linearly without an apparent MHD limit to the central pressure until the first ELM happens. In fact, the stored energy achieved before the first ELM is linearly proportional to the heating power applied. The first ELM after this phase usually also affects the plasma centre and thereby terminates the ITB. The duration of the ELM free phase is almost independent of heating power and the edge p saturates early on in this phase; a possible explanation is that the edge p is limited by a more benign MHD instability and only after the build up of sufficient bootstrap current in the edge does the first ELM occur. Figure 6 shows an example in which β N is reached together with very good confinement (H 89 = ). Concerning ITB physics, a clear reduction of turbulence within the ITB and at its foot point has been found using reflectometry, although the turbulence level does not go down to the diagnostic noise level at low frequencies, indicating that not all modes are suppressed []. A further point of interest is the observation that the foot point of the ITB is usually located just outside the shear reversal zone and the steep gradient zone then extends across the shear reversal zone [9]. This is of particular importance concerning the prospects for stationarity of an ITB discharge with large bootstrap current, since a location of the foot point inside the shear reversal zone would point to a tendency for the ITB to shrink in time. While the results mentioned so far mainly concern pronounced ion ITBs, we have also made progress in establishing electron ITBs with ctr-electron cyclotron current drive (ECCD). Previously, it had been found that ctr-eccd could lead to strong electron ITBs, but the ITBs had a tendency to collapse. By using higher I p and n e (i.e. a smaller fraction of driven current), it has been possible to sustain the duration of the ITB phase and reach T e values in excess of kev [9].. MHD stability.. Edge localized modes A major programmatic point on ASDEX Upgrade is the preparation of H-mode scenarii with tolerable ELMs, since extrapolations for the transient power loads onto the ITER divertor due to regular type I ELMs give rise to concerns. One 7

6 Overview of ASDEX Upgrade results Figure. Typical QH mode discharge in ASDEX Upgrade with an ELM-free phase of more than s. Note that between.8 and. s, no ECE data are available due to signal cut-off (- ---). 8 Plasma current ( ka) Heating power (MW) Stored energy ( kj) Density (9 ) β H N Time (s) 6 Confinement factor H 89 Normalized pressure β N q 9 H α light from divertor (a.u.) Time (s) Figure 6. ITB discharge with H-mode edge reaching transiently β N = and H 89 =. The first ELM ends the high performance phase. way to accomplish this is the operation with type II ELMs []. The operational space for type II ELMs in ASDEX Upgrade is characterized by high δ, closeness to double null and high density. Recent studies indicate that closeness to double null is required, whereas the constraint of high δ, which is to some extent coupled to the closeness to double null in ASDEX Upgrade, has recently been reduced to δ.. Also, the type II operational space is characterized by somewhat elevated q 9 with respect to the ITER value, but has been extended down to q 9 =. when operated close to double null []. MHD stability analysis using carefully reconstructed equilibria including an estimate of the edge current has shown that in the type II ELM configuration, the peeling ballooning stability for lower mode numbers is improved and also, the width of the eigenfunction for given n decreases due to the higher edge shear [,]. Both effects have a tendency to lead to smaller ELMs, consistent with the observation of type II ELMs under these conditions. Another regime with small ELMs is the EDA-regime found on ALCATOR C-Mod []. Experiments to establish dimensionless similar discharges between ASDEX Upgrade and C-Mod have been performed; however, they are complicated by the fact that the H-mode is not accessible in the dimensionless similar ASDEX Upgrade discharge and, at the higher heating power required in ASDEX Upgrade, ELM-free phases or type I ELMs occur [6]. This will require further studies, in particular because, as mentioned above, in JET-ASDEX Upgrade similarity experiments, the L H threshold has been found to scale in dimensionless similar discharges. Experiments were also performed to increase the type I ELM frequency by injection of shallow pellets that have little effect on n e, but instantaneously trigger a type I ELM. It could be shown that, by injecting pellets at a frequency higher than the natural type I ELM frequency, the ELM frequency could be significantly increased [7]. An example is shown in figure 7. Since the ELM energy loss associated with these ELMs is consistent with the scaling derived from variation of the natural ELM frequency (ν ELM δw const. at constant P heat ), this may open a route to mitigation of the peak heat loads for given heating power. As mentioned above, the ELM-free (QH)-mode regime was established on ASDEX Upgrade. The EHO that was found to be characteristic for this regime in DIII-D is also observed in ASDEX Upgrade during the ELM-free phase. Figure 8 shows an example where up to the th harmonic can be seen in the spectrum of a Mirnov coil. A feature previously not seen is shown on the right-hand side of figure 8. Here, data acquired 7

7 H. Zohm et al from magnetic coils with a very high frequency response up to MHz and sampled at MHz are shown together with the envelope of the EHO. Note the strongly nonsinusoidal character of the radial magnetic field oscillation which points to a poloidally localized structure, consistent with the existence of many harmonics. As can be seen, each cycle of the EHO coincides with a high-frequency burst at khz. The burst frequency actually decreases in time and therefore points to a fast particle drive of these oscillations. It is at present not clear, how this high frequency oscillation is linked to the EHO, but its high frequency seems to be related to the EHO frequency [8]. However, a role of fast particles in the EHO physics might explain why it is found only at low density and with ctr-nbi, where first orbit losses are much higher than in the co-case. The fast magnetic measurements shown above also showed interesting new features about MHD connected to ELMs [8]. While type III ELMs did not exhibit any other..9.. n ( e m -) W MHD (MJ) D outer div α..8.6 Time [s] Hz Hz # # Figure 7. Comparison of two discharges with same external control parameters. While # shows long ELM free phases with large ELM events in between, injection of shallow pellets at Hz in # provokes ELMs at this frequency. significant precursor activity than the already well-known 7 khz precursor, type I ELMs, which show pronounced low-frequency precursors usually only with ctr-nbi, exhibit a previously unresolved high frequency activity. This is shown in figure 9, where two examples from co- and ctr-nbi heated H-mode discharges are shown. A high frequency oscillation can clearly be seen for ms before the actual ELM. The frequency is different between co- and ctr-nbi, pointing towards a contribution of both ω e and v E B due to the beams, which add up for ctr-nbi, but have to be subtracted for co-nbi. In fact, the same local m number of 6 is found in both cases. Note that this is a local m measured in the midplane outside, and the actual m may be much higher since the local poloidal wavelength is highest on the midplane outside. It should be noted that this oscillation precedes type I ELMs, but actually does not always grow explosively before ELMs, so that it may be related to the saturation of edge pressure before an ELM occurs, but not necessarily to the MHD mode(s) responsible for the ELM itself... Neoclassical tearing modes ASDEX Upgrade continues to study ways to avoid, remove or mitigate the effects of neoclassical tearing modes (NTMs). Previous experiments demonstrated the possibility to stabilize NTMs with DC co-eccd, but without full recovery of confinement and β if the NBI power was kept constant [9]. Now, with higher NBI power, β N could be significantly increased above the NTM onset level without reappearance of the mode []. In addition, the confinement properties of the discharge recover to the value before NTM onset. However, this may be due to the use of the off-axis NBI sources which have a tendency to produce peaked density profiles [] and therefore counteract the confinement degradation by ECRH. An example is shown in figure. At even higher β N, the NTM reappeared in spite of the ECCD; however, this may f [khz] ASDEX Upgrade #6 n=8 n=9n= n=7 n=6 n= n= n= n= n= quiescent phase n= FB FB FB ELMy H-mode time [s] [a.u.] Radial magnetic field pickup coil (f > khz) ASDEX Upgrade #6 - - Radial magnetic field pick-up coil (filtered and integrated) - - Soft X-ray intensity (peripheral chord) H-alpha intensity (outer divertor) time [s] [a.u.] [a.u.] [W / m ] Figure 8. EHO in QH-mode. The low frequency mode shows harmonics up to n = (left part, fishbone activity and its harmonics are also seen, marked FB ). The right part shows that this is correlated with high frequency bursts transporting particles across the separatrix. 76

8 Overview of ASDEX Upgrade results f [MHz] co-inj., MW NI f [MHz] ctr-inj., MW NI.6. magnetic coil D α # time (s) magnetic coil.6 D α # time (s) Figure 9. High frequency magnetic oscillations prior to type I ELMs with co-nbi (left) and ctr-nbi (right). NBI Power (MW) ECRH Power (MW*) #989 n= amplitude( a.u) Toroidal Field Normalized Beta. Neutron Rate (e/s). H-Factor ITER98(p,y) time (s) Figure. NTM stabilization by DC ECCD on the high field side. The phases in which the (, ) NTM exists are marked by the shaded areas. With increased NBI power, β N can be raised significantly above the onset value and confinement recovers. be due to a mismatch between deposition and island due to the higher Shafranov shift and calls for feedback controlled deposition which will become possible in the near future, when a fast steerable ECRH launcher will be installed in ASDEX Upgrade. An interesting finding concerning NTMs was the discovery of the so-called frequently interrupted regime (FIR) NTMs, in which, for sufficiently high β,an(m, n) NTM never reaches its saturated island width because it is periodically reduced due to interaction with a (m +,n+) mode [,]. In this regime, the confinement reduction due to the NTM is significantly reduced and it may be a possible operating regime with tolerable NTMs. It was found that in discharges with peaked n e profile, the NTM onset β is unusually low, which may be explained by the fact that n contributes more strongly to the bootstrap current than T []. This effect has now been included in the NTM onset scalings by using ( T n+ n T)/ p rather than β N in the evaluation of the perturbed bootstrap that drives the mode. Also, experiments were carried out where the heating power was ramped down in order to determine β when the mode is turned off [, ]. This scaling is shown in figure. It can j bs /j pl [a.u.] 8 6 NTM onset NTM decay...8. ρ* Figure. Scaling of the perturbed bootstrap current at NTM onset and switch off versus ρ. Both curves show a linear dependence. be seen that a marked hysteresis exists between the onset and the switch-off, but the previously found ρ -dependence still holds... Sawtooth tailoring with ECRH Sawteeth play a key role in determining the central confinement of energy and particles [6]. On the one hand, a loss of sawteeth leads to improved confinement, on the other hand, it may lead to intolerably high central impurity concentrations. 77

9 H. Zohm et al τ / τ ST with ECCD ST w/o ECCD 7 6 HFS co-eccd inversion radius complete stabilization LFS τ / τ ST with ECCD ST w/o ECCD 7 6 HFS inversion radius complete stabilization ctr-eccd LFS ECR deposition in ρ ECR deposition in ρ Figure. Ratio of sawtooth repetition period with and without ECCD as a function of the ECCD deposition (negative values of ρ indicate deposition on the HFS). Complete stabilization can be achieved by co-eccd outside the inversion radius or central ctr-eccd. Therefore, a method to tailor the sawtooth amplitude and repetition frequency is highly desirable. This has special importance in the presence of fast particles, that act on sawtooth stability themselves. Experiments were conducted in ASDEX Upgrade in which.8 MW of ECRH power was applied to an H-mode discharge heated with MW of NBI. The location of the ECRH deposition was varied by slow ramps of the toroidal field. The results are summarized in figure. It can be seen that co-eccd leads to an increase in the sawtooth period when deposited just outside the sawtooth inversion radius. With 8 kw of ECRH/ECCD, it is actually possible to completely stabilize the sawteeth under these circumstances. For central deposition, a reduction in sawtooth period occurs. Similar behaviour is observed with pure ECRH, which is consistent with a local decrease of resistivity due to ECRH that also generates co-current. Conversely, for ctr-eccd, central deposition stabilizes the sawteeth, clearly demonstrating the importance of the local current generated by ECCD. These findings are consistent with the hypothesis that a decrease in current density gradient at the q = surface tends to stabilize sawteeth. Note that, contrary to stabilization of sawteeth by fast particles, the amplitude does not increase here. This method has already been used to create sawtoothfree discharges for impurity transport investigations; future studies will also examine the effect of sawtooth tailoring on NTM onset by affecting the seed island size... Disruptions Disruptions continue to be an important point of investigation on ASDEX Upgrade. A major concern for ITER are the forces connected with halo currents induced by the movement of a vertically unstable plasma during a disruption. On JET, it had been found that a large asymmetric radial force component can occur during disruptions that can considerably shift the vacuum vessel horizontally. On ASDEX Upgrade, careful measurements were performed to see if this also occurs, but the lateral forces were found to be only a minor fraction of the vertical forces occuring during a disruption [7]. Thus, it has not been possible yet to establish a connection between lateral force and the toroidal asymmetry in halo current that is usually measured in ASDEX Upgrade. Experiments on disruption mitigation focused on the application of an impurity gas puff at times atmospheric pressure as an alternative to the previously used killer pellet. From figure, it can be seen that with particles of Ne injected into a disruption, the forces were reduced up to roughly a factor of due to the faster I p decay and the smaller vertical displacement which reduce the halo currents. Thus, Ne gas injection is an attractive alternative to a killer pellet.. Exhaust and plasma wall interaction.. Operation with Divertor IIb As mentioned above, the outer leg of the ASDEX Upgrade divertor was redesigned to accommodate higher δ plasmas with good pumping and power handling capability. The design principle that had led to the previous Divertor II design was kept: vertical targets with close baffling ensure high divertor density and thus enable strong radiative losses from this region, mainly due to C radiation. Consequently, it was found that the power handling capability of Divertor IIb is as good as that of its predecessor [8]: although a somewhat higher heat flux to the target plates is observed in Divertor IIb under otherwise similar main plasma conditions, this can be attributed to the changed geometry that leads to a reduction of the wetted surface due to a steeper field line intersection angle. In fact, by correcting these effects and calculating the parallel heat flux, we arrive at very similar values for Divertor II and Divertor IIb. Figure shows the geometry and a comparison of heat fluxes in both divertors. Another important property of the divertor is its pumping capability. Comparison between the old and new configuration show only a modest change in the He compression, i.e. the favourable properties of Divertor II were preserved [8]. The new Divertor IIb exhibits the same favourable properties as its predecessor, but allows for more flexibility in plasma shaping which has become increasingly important for many aspects. 78

10 Overview of ASDEX Upgrade results no gas m.8.7 x particles. x particles I p x particles z curr strain gauge ms # a.u. ms #6 ms #8 ms # time (s) time (s) time (s).. time (s) Figure. Disruption forces for disruptions in which Ne gas was puffed to mitigate the loads. With particles, the forces are greatly reduced. power density (MW/m ) 6 #. s # 9. s (shifted on new strike-point position) transition module heat flux not calculated parallel heat flux (MW/m ) 8 6 DIV IIb DIV II #. s # 9. s..... length across surface (m) ρ pol Figure. Geometry of Divertor II (grey) and Divetor IIb (black) showing the better fit of high δ equilibria into Divertor IIb. The right figures show power density and parallel heat flux under similar conditions (ELMy H-mode with MW of NBI) for both divertors... Tungsten as first wall material Recent results on the co-deposition of T in C plasma facing components as well as the observation of T-enriched C-flakes in the louvers of the JET divertor have drawn increasing attention to alternative wall materials. In 996, ASDEX Upgrade had been successfully operated with a W-divertor. However, C as the main impurity prevailed, pointing towards a significant role of the C-covered inner heat shield as an impurity source. Thus, we started to successively cover the inner heat shield tiles with W. Operation with up to 7. of the 8. m of the inner heat shield covered with W has, under routine operation conditions, not shown any detrimental effect. In particular, the W-concentration rarely exceeded, which is the critical number for ITER. Conditions in which this number was exceeded were either related to direct contact of the plasma with the inner wall or conditions in which the electron density showed a strong tendency to peak, such as with low central heating. As has been mentioned above, in these cases, addition of central heating led to a large reduction of the central W-concentration []. Erosion of W from the inner heat shield showed a pronounced asymmetry over the tiles that led to the conclusion that it is mainly due to ion bombardment (i.e. D and impurities) which, due to the guiding of the particles by the magnetic field, can lead to shadowing of certain areas [9]. This was even found when the plasma was relatively far away from the inner wall, which demonstrates that plasma reaches far across the separatrix, a fact that also manifests in the observation of two different decay lengths of the SOL density []. This may be due to the bursty nature of SOL transport which can expel particles further than would be expected from ordinary diffusion. Figure shows a view of the inner heat shield covered with W-coated tiles. Also shown are traces from an experiment where the plasma was moved towards the central column and then away from it. From the observed increase in W-influx and the concomitant increase in the W-concentration in the main plasma, a penetration probability for W eroded from the inner wall to enter the main plasma can be derived []. This number is found to be of the order of %, consistent with modelling using the DIVIMP []. The analysis of migration 79

11 H. Zohm et al _ n ( /m ) e... T e (kev) 8 Γ W ( at./m s) 999 doubletiles divertorbaffle inner gap (cm) t (sec) -6 c ( ) W.... t (sec) Figure. Inner wall with W-coated tiles (left) and traces from an experiment where the plasma was brought near the inner wall to determine the relation between W-influx and W-content. The position scan does not influence global plasma parameters such as n e, T e. and redeposition patterns is less clear due to the fact that the main W-erosion actually comes from the plasma ramp-down phase at the end of the discharge when the hot plasma column leans on the inner wall []. It should be pointed out that C is still present in these discharge and significantly contributes to the radiation; radiation control in a C-free environment remains a challenge to be addressed in the future, when full W coverage is implemented. 6. Scenario integration The physics elements described above are all important steps towards an integrated scenario for ITER. However, the integration itself, i.e. the combination of good confinement, high stability and benign power exhaust at good pumping capability, becomes more and more important, since it is clear that some of the solutions for individual points are not mutually compatible. Therefore, the development of integrated scenarii is a major focus of the ASDEX Upgrade experimental programme. For the conventional H-mode scenario, the remaining issues are the NTM stability (see section.) and the mitigation of ELM effects on the divertor as addressed in section.. However, recent developments aim at improving the performance above the standard H-mode in view of steady state, i.e. noninductively driven tokamak discharges, the so-called advanced tokamak (AT) line. The route of development here is to maximize the fraction of bootstrap current, which requires high β pol and to reduce the total current which requires enhanced confinement, expressed, e.g. as H 89 = τ E /τ E,ITER89. Figure 6 shows the commonly used figure of merit, β N H 89, for various ASDEX Upgrade scenarii. As can be seen, ITB scenarii reach very high values, but only for short duration due to the stability limitations discussed in section.. In the so-called improved H-mode and the high β N H-mode, high performance has been achieved in stationary state over many confinement times. In fact, although the highest neutron rate in ASDEX Upgrade is reached in ITB discharges, the highest Q equiv is actually achieved in the improved H-mode []. The improved H-mode is characterized by flat central shear and the absence of sawteeth while fishbones maintain the stationarity of the q-profile []. The confinement improvement mainly comes from density peaking while the temperature profiles are still similar to usual H-mode profiles, but with a relatively high pedestal temperature as boundary condition. A major drawback of this scenario was the limitation to relatively low densities n e /n GW.. As noted before, operation at higher δ allows us to obtain good confinement at high fractions of n GW. Thus, by using higher δ, it was possible to extend the performance of the improved H-mode regime to n e /n GW.8.9. Again, these discharges do not have sawteeth and the n e profile is slightly peaked. Actually, higher δ also proved to be beneficial to lower q 9 in the improved H-mode regime, which was not possible before due to the onset of sawteeth with lower q 9. The degree of peaking can be controlled by central ICRH heating as described in section.. This regime also shows improved stability properties with respect to the improved H-mode regime (with the main limitation being NTM onset) and is called high β N -regime [, ]. It can be obtained in a stationary state at values up to β N =. [6]. This may partly be explained by to the relatively high collisionality, but stability also profits from shaping: increased shaping is expected to affect the stability index, and it allows higher β N for given β pol, which is the actual quantity limited by NTMs. The beneficial effect of δ is also clearly documented in β N H 89 n/n GW in the right graph of figure 6, where at low δ.., the value of β N H 89 n/n GW does not exceed, but can be in excess of. for high δ.., which reflects the above mentioned possibility to achieve good confinement at high Greenwald fraction. Note that the region of intermediate δ.. was not studied in detail so far, and is thus not fully populated in the figure. Figure 7 shows an example of a high β N discharge. In addition to the favourable performance in steady state, it can also be seen that the conditions for this scenario, i.e. high δ and closeness to double null in combination with high n e are also compatible with type II ELMs, which can be seen from the plots of deposited power on the target plates: although both frames are obtained in phases with equal heating power, the 8

12 Overview of ASDEX Upgrade results Figure 6. Performance parameter β N H 89 for various high performance scenarii in ASDEX Upgrade (left). The right figure demonstrates the beneficial effect of increasing δ in achieving higher densities at high performance..-. s.9-.9 s β N D α H9-P n e /ne GW 6 7 Time (s) # Figure 7. High β N -discharge with type II ELMy edge. The power flux to the divertor plates as shown by the inserts exhibits type I peaks early in the discharge (left), but is continuous later on in the type II phase (right). distinct peaks due to type I ELMs exceed MW m whereas in the later phase with type II ELMs, the heat flux is continuous and below MW m. For ITER, this scenario therefore shows main advantages compared to the usual H-mode scenario, i.e. the possibility to run at lower I p and higher bootstrap fraction at similar fusion power (note that P fus /P loss only depends on the ratio H/q), but reduced peak heat flux on the divertor, and therefore to extend the duration of the discharge. However, the bootstrap fraction is still smaller than in an ITB discharge with same β. Therefore, ITB discharges are still the prime candidate for steady state tokamak operation, but the high β N scenario may be considered an intermediate step on this route. It is therefore of great importance to understand if the main ingredients, i.e. the density peaking, improved NTM stability and type II ELM edge, scale in a consistent way to ITER. This will be the subject of further work. 7. Summary and conclusions Recent ASDEX Upgrade results have enabled progress in the understanding of transport, stability and exhaust. An improved understanding of heat and particle transport in the plasma core has been gained using a stiff temperature model and coupling of heat and particle transport coefficients. Concerning stability, active control strategies for both ELMs and NTMs start to emerge as well as regimes in which the detrimental effects of these instabilities are mitigated. It has also been shown that W can be considered a serious candidate for a suitable first wall material in reactor-grade devices. A scenario has been demonstrated in which many of the ITER requirements are simultaneously met or even exceeded. Future work will focus on improved understanding of underlying physics as well as integration of the different physics elements to a consistent scenario. The future hardware upgrades, i.e. an extension of the ECRH system to higher power and longer pulse length at variable frequency and an increasing W coverage of the plasma facing components will enable further progress in this direction. References [] Gruber O. et al Overview of ASDEX Upgrade results Nucl. Fusion 69 8

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