Overview of ASDEX Upgrade results

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1 Overview of ASDEX Upgrade results O.Gruber,R.Arslanbekov,C.Atanasiu a,a.bard,g.becker,w.becker, M. Beckmann, K. Behler, K. Behringer, A. Bergmann, R. Bilato, D. Bolshukin, K. Borrass, H.-S. Bosch, B. Braams b, M. Brambilla, R. Brandenburg c, F. Braun, H. Brinkschulte, R. Brückner, B. Brüsehaber, K. Büchl, A. Buhler, H. Bürbaumer c,a.carlson,m.ciric,g.conway,d.p.coster,c.dorn,r.drube, R. Dux, S. Egorov d, W. Engelhardt, H.-U. Fahrbach, U. Fantz e,h.faugel, M. Foley f,p.franzen,p.fu g, J.C. Fuchs, J. Gafert, G. Gantenbein h,o.gehre, A. Geier, J. Gernhardt, E. Gubanka, A. Gude, S. Günter, G. Haas, D. Hartmann, B. Heinemann, A. Herrmann, J. Hobirk, F. Hofmeister, H. Hohenöcker, L. Horton, L. Hu g,d.jacobi,m.jakobi,f.jenko,a.kallenbach,o.kardaun, M. Kaufmann, A. Kendl, J.-W. Kim, K. Kirov, R. Kochergov, H. Kollotzek, W. Kraus, K. Krieger, B. Kurzan, G. Kyriakakis i, K. Lackner, P.T. Lang, R.S. Lang, M. Laux, L. Lengyel, F. Leuterer, A. Lorenz, H. Maier, K. Mank, M.-E. Manso j,m.maraschek,k.-f.mast,p.j.mccarthy f,d.meisel,h.meister, F. Meo, R. Merkel, V. Mertens, J.P. Meskat h,r.monk,h.w.müller, M. Münich, H. Murmann, G. Neu, R. Neu, J. Neuhauser, J.-M. Noterdaeme, I. Nunes j, G. Pautasso, A.G. Peeters, G. Pereverzev, S. Pinches, E. Poli, R. Pugno, G. Raupp, T. Ribeiro j,r.riedl,s.riondato,v.rohde,h.röhr, J. Roth, F. Ryter, H. Salzmann, W. Sandmann, S. Sarelma, S. Schade, H.-B. Schilling, D. Schlögl, K. Schmidtmann, R. Schneider, W. Schneider, G. Schramm, J. Schweinzer, S. Schweizer, B.D. Scott, U. Seidel, F. Serra j, S. Sesnic, C. Sihler, A. Silva j, A. 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, N. Tsois i,w.ullrich,m.ulrich,p.varela j, O. Vollmer, U. Wenzel, F. Wesner, R. Wolf, E. Wolfrum, R. Wunderlich, N. Xantopoulos i, Q. Yu, M. Zarrabian, D. Zasche, T. Zehetbauer, H.-P. Zehrfeld, A. Zeiler, H. Zohm Max-Planck-Institut für Plasmaphysik, Euratom IPP Association, Garching, Germany a Institute of Atomic Physics, Bucharest, Romania b Courant Institute, New York University, New York, N.Y., United States of America c Technical University of Vienna, Vienna, Austria d Efremov Institute, St. Petersburg, Russian Federation e University of Augsburg, Augsburg, Germany f University College, Cork, Republic of Ireland g Academia Sinica, Hefei, China h Institut für Plasma Forschung, University of Stuttgart, Stuttgart, Germany i NSCR Demokritos, Athens, Greece j Centro de Fusão Nuclear, Lisbon, Portugal Nuclear Fusion, Vol. 41, No. 1 c 1, IAEA, Vienna 1369

2 O. Gruber et al. Abstract. Ion and electron temperature profiles in conventional L and H mode on ASDEX Upgrade are generally stiff and limited by a critical temperature gradient length T/T as given by ion temperature gradient (ITG) driven turbulence. ECRH experiments indicate that electron temperature (T e) profiles are also stiff, as predicted by electron temperature gradient turbulence with streamers. Accordingly, the core and edge temperatures are proportional to each other and the plasma energy is proportional to the pedestal pressure for fixed density profiles. Density profiles are not stiff, and confinement improves with density peaking. Medium triangularity shapes (δ <.45) show strongly improved confinement up to the Greenwald density n GW and therefore higher β values, owing to increasing pedestal pressure, and H mode density operation extends above n GW.Densityprofile peaking at n GW was achieved with controlled gas puffing rates, and first results from a new high field side pellet launcher allowing higher pellet velocities are promising. At these high densities, small type II ELMs provide good confinement with low divertor power loading. In advanced scenarios the highest performance was achieved in the improved H mode with H L-89P β N 7. atδ =.3 for five confinement times, limited by neoclassical tearing modes (NTMs) at low central magnetic shear (q min 1). The T profiles are still governed by ITG and trapped electron mode (TEM) turbulence, and confinement is improved by density peaking connected with low magnetic shear. Ion internal transport barrier (ITB) discharges mostly with reversed shear (q min > 1) and L mode edge achieved H L-89P.1 and are limited to β N 1.7 by internal and external ideal MHD modes. Turbulence driven transport is suppressed, in agreement with the E B shear flow paradigm, and core transport coefficients are at the neoclassical ion transport level, where the latter was established by Monte Carlo simulations. Reactor relevant ion and electron ITBs with T e T i 1 kev were achieved by combining ion and electron heating with NBI and ECRH, respectively. In low current discharges full non-inductive current drive was achieved in an integrated high performance H mode scenario with n e = n GW,highβ p =3.1, β N =.8 andh L-89P =1.8, which developed ITBs with q min 1. Central co-eccd at low densities allows a high current drive fraction of >8%, while counter-eccd leads to negative central shear and formation of an electron ITB with T e() > 1 kev. MHD phenomena, especially fishbones, contribute to achieving quasi-stationary advanced discharge conditions and trigger ITBs, which is attributed to poloidal E B shearing driven by redistribution of resonant fast particles. But MHD instabilities also limit the operational regime of conventional (NTMs) and advanced (double tearing, infernal and external kink modes) scenarios. The onset β N for NTM is proportional to the normalized gyroradius ρ. Complete NTM stabilization was demonstrated at β N =.5 using ECCD at the island position with 1% of the total heating power. MHD limits are expected to be extended using current profile control by off-axis current drive from more tangential NBI combined with ECCD and wall stabilization. Presently, the ASDEX Upgrade divertor is being adapted to optimal performance at higher δ s and tungsten covering of the first wall is being extended on the basis of the positive experience with tungsten on divertor and heat shield tiles. 1. Introduction The ASDEX Upgrade non-circular tokamak programme has been largely focused on: (i) Confinement and performance related physics in the ITER baseline scenario, the ELMY H mode near operational limits, with close interaction between edge and core plasma; (ii) Investigation of scenarios and physics of advanced tokamak plasma concepts with internal transport barriers (ITBs) leading to enhanced performance and possibly stationary operation; (iii) MHD stability and active stabilization of β limiting instabilities as well as avoidance and mitigation of disruptions; (iv) Edge and divertor physics in these high power, high confinement regimes, with the aim of identifying and optimizing power exhaust and particle control (ash removal). The similarity of ASDEX Upgrade to ITER in respect of poloidal field coil system and divertor configuration makes it particularly suited to testing control strategies for shape and plasma performance. NBI was available with injection energies of up to 1 kev for deuterium (with MW), which allowed studies of the influence of heat deposition on transport and of fast particle effects. Minority heating with the ICRH system (up to 5.7 MW coupled) was used for central heating at high densities, and to study the influence of deep particle refuelling and toroidal rotation by substituting NBI. 137 Nuclear Fusion, Vol. 41, No. 1 (1)

3 Article: Overview of ASDEX Upgrade results The ECRH system (four gyrotrons with a coupled power of 1.4 MW for s) allowed pure electron heating, transport studies and feedback stabilization of neoclassical tearing modes (NTMs). Provisions for current drive and for active control of current profiles in advanced scenarios were available with nearly perpendicular NBI and ECRF using steerable mirrors (up to 35 ka driven on-axis). Experiments using on-axis fast wave or off-axis mode conversion current drive with the ICRF system have been started. The present version of the closed divertor II, LYRA (vertical target plates including a roof baffle between them, cryopump), which is rather similar to the ITER-FEAT reference design, allows a strong reduction of heat flux to the target plates and is capable of handling heating powers of up to MW or P/R of 1 MW/m. The observed strongly reduced, distributed power flux to the surrounding structures during both ELM and ELM free phases is in agreement with B-Eirene simulations [1]. The study of reactor compatible materials was continued by covering parts of the low field side (LFS) heat shield with tungsten to reduce carbon influx []. High shaping capability with elongations κ 1.8 and triangularities δ of up to.45 at the separatrix allowed study of the influence of plasma shape on performance and operational limits. Such equilibria have been run with the outer strike point being on the top of the roof baffle or on the outer vertical target to obtain higher pumping speeds. Additionally, the similarity in cross-section to other divertor tokamaks is important in determining size scalings for core, edge and divertor physics. This collaborative work, including extrapolation to ITER parameters, has continued and was even enhanced in the new JET operation under the European Fusion Development Agreement (EFDA). The promising method of refuelling by pellet injection yielded first results with a new high field side (HFS) pellet launcher capable of injection speeds of up to 1 m/s. The following sections report on recent progress made in ASDEX Upgrade experiments since the last IAEA conference [3]. First, ion and electron temperature profile stiffness (Section [4, 5]) and high density H mode operation (confinement, density profile peaking, HFS pellet injection and tolerable type II ELMs) (Section 3 [4 6]) are described. Cold pulse and impurity transport studies are not reported here [7, 8]. The next sections are devoted to advanced scenarios such as improved H mode, ion ITBs and reactor relevant combined ion and electron ITBs (Section 4 [4, 5, 9, 1]), and fully non-inductively driven discharges at high β p andwitheccdexhibiting electron ITBs (Section 5 [9, 1]). Finally, stabilization of NTMs by ECCD (Section 6 [11, 1]), MHD in advanced scenarios (Section 7 [1]) and operation with high Z wall coatings (Section 8 []) are presented.. Temperature profile stiffness Understanding of transport in the conventional H mode scenario is still an active field of research because it determines the prediction and design of next step devices. As reported earlier [3, 13, 14], ion temperature (T i ) profiles in conventional H mode discharges of ASDEX Upgrade are generally stiff : temperature profiles are limited by a maximum value of T/T, the inverse temperature gradient length L T (Fig. 1). At medium and high densities the electron temperatures (T e ) show the same behaviour, since they are closely coupled to T i and heated as well by NBI. Turbulence driven by the ion temperature gradient (ITG) is believed to be the main origin of turbulence causing the anomalous transport through the ion channel [15 17]. It leads to limitation of T i at a critical value of T/T as observed. The transport properties of temperature gradient driven turbulence can be expressed for the heat diffusivity by two additive terms. One term represents transport without temperature gradient driven turbulence, for instance neoclassical transport for ions, while the other represents the transport by temperature gradient driven turbulence, which is influenced by magnetic and rotational shear, effective ion mass and density gradient length. This term is equal to zero for T/T < ( T/T) c but increases strongly when T/T > ( T/T) c and may saturate. Then transport is high and keeps the profiles close to ( T/T) c. Gyro-Bohm scaling χ gb T 3/ of this term causes an increase of stiffness with temperature. These considerations are not necessarily valid inside the sawtooth inversion radius, where profiles may be limited by the MHD activity, or at the edge, where the temperatures may be too low. The profile stiffness reflects a strong coupling between core and H mode pedestal temperatures or, equivalently, the fact that the profiles plotted on a logarithmic scale are shifted vertically according to their edge temperature (Fig. 1). Nuclear Fusion, Vol. 41, No. 1 (1) 1371

4 O. Gruber et al #1761 T i (applied at r/a =.8) and the density profiles used in the simulations are taken from the experiment... Electron temperature profile stiffness with strong electron heating T i [kev] ρ pol Figure 1. HmodeT i profiles are stiff during a density ramp-up (lines) and remain fixed for on- and off-axis heat deposition (symbols, same heating power) with temperature gradient length T/ T R/5..1. Temperature profile stiffness in NBI heated H modes Profile stiffness in NBI heated H modes was confirmed by comparing the influence of rather different heat deposition profiles using 6 and 93 kev neutral beams at high plasma densities ( n e 1 m 3 ): measured temperature profiles are unaffected. The reaction of the transport coefficients to these different heating power profiles is as expected. For the deeper deposition with 93 kev beams, χ eff in the core increases by a factor of almost in relation to the more off-axis deposited power with 6 kev beams. At these high densities, owing to the strong electron ion coupling, only an effective heat diffusivity can be derived from power balance. Obviously, the heat conductivity adjusts itself to maintain the observed stiff temperature profiles and ceases to be a useful description of transport behaviour. In addition to these studies, a database including profile data from about 3 H mode discharges with NBI [5, 17] indicates that T i and T e profiles are quite stiff at medium and high densities. Simulations of these discharges using three models based on ITG and including trapped electron mode physics and E B shear flow IFS/PPPL [15], GLF3 [18] and Weiland [19] agree well with these data and confirm the ITG turbulence as the driving mechanism. The boundary condition for the temperature Recent calculations [] indicate that electron temperature gradient (ETG) turbulence is able to drive a large electron heat transport because large cells, streamers, can develop. The properties of temperature gradient driven turbulence for ions and electrons are qualitatively similar, as has been confirmed in ASDEX Upgrade discharges with T e >T i using ECRH in steady state and power modulation experiments [4]. Central heating using either.8 or 1.6 MW of ECRH power at different plasma currents resulted in similar shapes of the electron temperature profiles, which are shifted according to their edge temperatures in the logarithmic scale of Fig.. The shaded area indicates a region where the slope of the profiles, i.e. T/T, is the same. In the centre the slope changes and is different for each profile. This region extends beyond the inversion radius of the sawteeth. Obviously, the heat flux to the electron channel is not high enough to reach ( T/T) c. Towards the edge the profiles gradually deviate from the constant slope, which seems to occur for T e values below 6 ev. In the stiff region the value of T/T does not depend on I p, which implies that the well known improvement of confinement (and temperature increase) with I p is provided by higher edge temperatures. Electron heat diffusivities of these discharges from power balance analyses normalized to the gyro- Bohm scaling T 3/ exhibit the expected properties, namely small χ/t 3/ values for T e /T e < 7m 1 and a strong increase above this threshold [4]. The normalized diffusivities level off at different values as the stiffness increases with higher temperatures. This is also in qualitative agreement with ETG calculations, indicating that the appearance of streamers requires a finite value of temperature []. The results from a similar discharge in hydrogen indicate that the electron transport is not sensitive to the ion mass. Complementary to these experiments with central ECRH, discharges with off-axis ECRH indicate that transport indeed decreases in the region between the plasma axis and ECRH deposition, consistent with the decrease of T/T to below the critical value [4]. These steady state analyses were supplemented by ECRH power modulation experiments confirming 137 Nuclear Fusion, Vol. 41, No. 1 (1)

5 Article: Overview of ASDEX Upgrade results 6 T e [kev] 1.1 not stiff:q e too low 1.6 MW 1 MA OH 1 MA 1.6 MW.6 MA OH.6 MA not stiff: T e too low stiff region #13558 # ρ tor Figure. Stiff T e profiles for pure ohmic heating and additional central ECRH (1.6 MW) at two plasma currents. the electron temperature stiffness. These investigations further revealed that changes in transport for off-axis ECRH cases are caused by a change of diffusivity and cannot be attributed to convection [4]. In the stiff region, the heat pulses propagating outwards and inwards yield different values for the derived χ HP by virtue of the fact that heat pulses tend to increase or decrease T/T locally, depending on their direction of propagation. The χ HP values are above the χ s derived from power balance (χ PB ), and their ratio χ HP /χ PB strongly increases with temperature, depending again on the direction of the pulses. This ratio characterizes the stiffness of the profile. Accordingly, differences in the heat pulse propagation between deuterium and hydrogen discharges at equal discharge parameters, but at different resulting temperatures, can be reconciled by the temperature dependence of the stiffness. 3. High density H mode operation 3.1. Influence of non-stiff density profiles and triangularity on energy confinement To achieve high fusion yield by means of a high central density and low power load in the divertor with sufficient SOL density, a fusion reactor has to operate at high densities close to the Greenwald density n GW. But confinement degrades at high line averaged densities n e towards n GW, while it improves with n e at low densities. This is a consequence of temperature profile stiffness (T core T ped ), the nearly fixed pedestal pressure of the edge H mode barrier limited by MHD stability and increasing with triangularity, and the fact that density profiles tend not to be stiff. At low densities more peaked density profiles are observed, while increasing the density by gas puffing leads to density profile broadening, and the edge density strongly increases in relation to the line averaged density [3]. Correspondingly, the stored energy even decreases with increasing n e and stiff T profiles, in contrast to the ITER confinement time scalings containing a positive contribution from n e. For a given shape of the density profile, the plasma energy is expected to be proportional to the pedestal pressure. This is indeed confirmed for n e <.8n GW by 3 representative discharges, where a strong correlation between electron pedestal pressure p e,ped (at ρ =.8) and thermal plasma energy W th is seen. A linear regression applied to this data set yields: W th p.75 e,ped P h.14 with a low RMSE value of about 1% [4]. This representation provides a unification of discharges with different ELM types, I p s and triangularities (.1 <δ<.4), which only act on confinement via the edge pressure, namely as p e,ped Ip 1.4 P h. δ.3. The dependence upon I p does not quite reach Ip as expected from ballooning mode stability, which might be due to changes of other quantities in the edge region influencing stability, e.g. magnetic shear. An overview of the confinement data is provided in Fig. 3 for τ E normalized by the IPB98(y) ELMy H mode scaling, the H H-98P factor. The positive density dependence in IPB98(y) accentuates the confinement degradation at high densities by an even stronger decrease of the H factor. The benefits resulting from more highly triangular plasma shapes are first manifested in enhanced confinement with H H-98P 1 at the Greenwald density due to higher pedestal pressures. Then back-transition into the L mode at high densities is shifted to densities above n GW, thus substantially extending the operation regime of the H mode, but it also exhibits a large scatter at a given value of n e /n GW. This scatter can be attributed to the non-stiff density profiles, which, Nuclear Fusion, Vol. 41, No. 1 (1) 1373

6 O. Gruber et al. τ / τ(iterh-98p) δ =.35 δ =.9 δ =.18 ITER FEAT n / n Figure 3. Normalized thermal energy confinement versus normalized line averaged density n e at different triangularities δ. besides being more peaked at low densities than at higher ones, can vary according to the discharge scenario. This is taken into account in a regression by including the central density n e () and the density in the scrape-off layer averaged over 6 cm starting from the separatrix outwards, n e,sol, yielding the following expression: e GW τ E,th I 1. p P.56 h δ. n e ().3 n.17 e,sol (RMSE: 9%)[4]. This expression reflects the good confinement for peaked density profiles and confinement degradation with gas puffing represented by n e,sol.itisalsoin agreement with the W th and p e,ped scalings given before. The low RMSE is provided by the fact that this variable set takes the physics background into account. 3.. Density profile peaking with constant gas puffing Often in experiments the density is ramped up quickly to the anticipated value by strong gas puffing, and the high neutral gas flux causes high edge densities, with their negative effect on global confinement.with soft densitycontrolorlowexternalgas flux, the line averaged density slowly increases on electron density (1 m ) MW NBI ASDEX Upgrade # MW ICRH.5 MW NBI.5 MW ICRH 5 MW NBI 7.5 MW NBI normalized poloidal flux radius Figure 4. Comparison of electron density profile peaking in H mode at the Greenwald density with moderate gas puffing using different heat deposition scenarios. a timescale of a second (many energy confinement times) and can exceed the Greenwald density [4]. After an initial decrease the confinement time itself again increases and H H98-P = 1 is restored. The density profile shape broadens first because of the increasing edge density, but on the longer timescale the density profile significantly peaks, as shown in Fig. 4 for 5 MW NBI. In this phase the pedestal and SOL regions, which are determined by the particle source, remain unchanged, as also do the stiff temperature profiles. As explained above, the nonstiff density profiles together with stiff temperature profiles are the reason for the decrease of confinement observed at the beginning of the gas pulse and the confinement enhancement during the peaking phase. Finally, good confinement is lost owing to high central radiation caused by loss of sawteeth leading to accumulation of heavy impurities in the centre. Before accumulation Z eff is below 1.3 in the core. The operational window for the density profile peaking seems to be limited by a correlation between heat and particle transport. The heat flux was changed in the experiments by increasing the heating power or its deposition profile by combining NBI (slightly hollow heating profile) and ICRF minority heating (central deposition). This causes the region with density peaking to shrink with increasing heating power, and the density profiles even become flat when central heating is increased (Fig. 4). In the latter case the particle source with NBI is not changed and therefore cannot be the reason for the loss of density peaking. The increase of central power deposition in both cases, with its increase in transport 1374 Nuclear Fusion, Vol. 41, No. 1 (1)

7 Article: Overview of ASDEX Upgrade results at limited T/T, is responsible for the density flattening. In summary, an inward particle pinch is involved (Ware pinch seems to be sufficient), which is cancelled by increasing outward diffusion at higher central heating. These observations are supported by applying ECRH in various types of discharge which react to this additional heating with flatter density profiles. One may speculate that ECRH increases the electron heat flux and thus the ETG turbulence level. This density pump-out was observed earlier in other devices with ECRH [1]. n e 4 m/s 5 ms 1 19 m Advanced HFS pellet injection With injection of frozen hydrogen pellets from the magnetic LFS, the fuelling efficiency in strongly heated plasmas was found to be poor. In contrast, pellet injection from the HFS, first performed on ASDEX Upgrade, was found to give significantly higher efficiency, resulting in H mode discharges well above the Greenwald density with high confinement properties []. This is due to the fast radial drift in the toroidal field gradient of the high β plasmoid that forms around the ablating pellet, which transfers the pellet particles towards the plasma centre [3]. In both injection scenarios performance is limited by prompt convective particle and energy losses resulting from pellet induced ELM bursts within 1 ms after injection. This problem should be overcome by applying higher launch velocities to deposit the pellet beyond the ELMing region at the edge. However, strongly curved guiding tubes with turning points of the curvature and the fragile nature of cryogenic deuterium have up to now imposed a speed limit of 4 m/s. After a systematic study we have designed a new HFS pellet injection system using a looping geometry, where the pellets are ejected away from the tokamak device by the accelerating centrifuge and then guided back to the torus HFS by a roughly elliptical track [4]. In first proof of principle experiments with a preliminary looping system, roughly 5% of the pellets with a velocity of 56 m/s could be injected into the plasma with only a small mass loss [6]. At higher velocities of up to 1 m/s, only fractions of the pellet reached the plasma. At a velocity of 56 m/s, a deeper penetration (an increase from.165 to.15 m) and particle deposition towards the centre were found, mitigating the fast convective losses connected with the pellet ELMs (Fig. 5), but the pellet induced ELMs during the first 15 ms are also 56 m/s Figure 5. Density decay after HFS pellet injection at two velocities and ELM activity (D α in the divertor). The initial density decay time increases from 48 (v = 4 m/s) to 58 ms at higher velocity (average from several traces). The ELM activity in the first 15 ms after pellet injection is reduced at the higher pellet speed. alleviated. Further optimization of the looping system should enable us to define the demands of pellet injection systems for future large scale devices Tolerable type II ELMs close to the Greenwald density H modes with type I ELMs exhibit the best confinement performance, but the concomitant large power load during the ELM activity may not be acceptable in a reactor. Type III ELMs show almost no additional heat load on targets, but they entail confinement degradation. Small high frequency (or grassy) ELMs of type II were recently observed in DIII-D [5] and JT-6U [6], whereas the enhanced D α regime was reported from Alcator C-Mod [7]. These regimes were obtained for q 95 > 4.5 and δ>.4, but only at medium densities. On ASDEX Upgrade, type II ELMs were identified in a comparable operation region close to a double null configuration, but in contrast to the earlier experiments at high densities at the Greenwald density [4]. This new result is quite attractive for a future reactor. In addition the confinement in this type II ELM regime is almost as good as with type I ELMs and is significantly higher than for type III ELMs, providing τ E /τ H98-P.95. The Nuclear Fusion, Vol. 41, No. 1 (1) 1375

8 O. Gruber et al. O I O O I I type I ELMs type II ELMs H α outer divertor 1 V time (s) type III ELMs H α outer divertor H α outer divertor time (s) time (s) 1 V MW/m Figure 6. Heat fluxes to the outer (o) and inner (i) strike point regions from infrared thermography for type I, II and III ELMs. The height of each bar corresponds to a poloidal length of 7 cm along the target surface. The ELM activity is indicated by the H α traces. thermography data of the divertor target plates show that the peak power load is strongly reduced with type II ELMs, reaching at most MW/m almost without peaks on the outer plate, in comparison with 5MW/m for type I ELMs (Fig. 6). The power flux on the inner plates is always very low under these conditions. Magnetic measurements indicate that type II ELMs often have a precursor at 15 3 khz, indicating that they are indeed MHD events. Their signatures are different from those of type III ELMs, and they provide better confinement because they preserve the high edge pressure. 4. Advanced tokamak discharges Advanced tokamak scenarios aim at an operating regime with improved confinement and stability properties (high normalized beta β N = β/(i p /ab t ) > 3), thus allowing a reduced size tokamak reactor at lower plasma current to be operated in steady state [8 3]. ITBs, which are characterized by large pressure gradients in the region of reduced 1 V plasma turbulence, intrinsically drive bootstrap currents, which can constitute a significant part of the plasma current. Steady state operation requires that the inductively driven plasma current be completely replaced by external non-inductively driven currents (in addition to the bootstrap current), which, however, should be kept low to minimize the recirculating energy. On ASDEX Upgrade we have used nearly perpendicular NBI and ECCD for on-axis current drive up to now. Two advanced scenarios were investigated by modifying the current density profile by means of early heating in the current ramp. Both involve peaked profiles of density and toroidal rotation velocity. The one gives improved core confinement in combination with an H mode edge barrier and a flat central q profile, offering stationary, inductively driven improved H mode operation [3 34]. This scenario may allow ignition even in the ITER-FEAT device. Secondly, ITBs with mainly reversed magnetic shear and q min values above 1.5 exhibit, at least in combination with an L mode edge, steep pressure gradients in the barrier region and therefore bootstrap currents exceeding 5% of the plasma current [3, 34 36]. A combination of ITB and H mode edge is needed for true steady state, non-inductively driven tokamak operation combining improved performance and high bootstrap current fraction f bs = I bs /I p β N q 95 A,athighq95 4 and β N > Improved H mode The improved H mode regime is obtained with low preheating in the current ramp-up and subsequent power increase causing a transition into the H mode. A stationary state is obtained for up to 4 energy confinement times and several internal current diffusion times and is limited only by the available pulse length [3, 3, 33]. Continuously reconnecting (1,1) fishbones clamp the current profile in the core and lead to stationary central zero shear with q min 1 (see Section 7.1) [33 35]. The density in this scenario was extended up to slightly below half of the Greenwald density to integrate power exhaust and He ash removal by appropriate divertor conditions. With gas fuelling, the threshold heating power for ITB formation had to be raised by 5%. A density increase caused by improved core particle confinement at more triangular plasma shapes does not change the ITB onset conditions and allows higher β N values up to.6 without destabilizing NTMs. The confinement increases to H H98-P =1.7 orh L89-P = 3, this being largely 1376 Nuclear Fusion, Vol. 41, No. 1 (1)

9 Article: Overview of ASDEX Upgrade results 1 [kev] central T i T i at r/a = T at r/a=.8 [kev] i 6 4 central T e Ti behaviour T e at r/a = T at r/a=.8 [kev] e 1 hysteresis for gas puff? central density / density at r/a = H factor ITER98-P n e at r/a=.8 [1 m ] n [119m-3 e ] Figure 7. Comparison of temperature and density profiles and H H-98P factor of standard (blue symbols) and improved (red symbols) H mode (density scan with.3 < n e/n GW <.6 at I p =1MA,P NBI =5MW,δ =.). due to the increasing pedestal pressure with higher δ [3, 34]. The more effective impurity screening at higher edge densities reduces Z eff from 3 (at.35n GW ) to in the plasma centre. No temporal accumulation of impurities is observed, but impurities show centrally peaked densities consistent with neoclassical predictions. First measurements point to sufficient He exhaust in this scenario. Detailed studies have shown that the confinement physics in this regime can be largely unified with that of the standard H mode [5]. In Fig. 7 the data of a density scan.3 < n e /n GW <.6 withthesamei p (1 MA), the same NBI heating power (5 MW) and low triangularity δ =. are shown. The improved H mode shots have in general a lower line averaged density and higher central as well as edge ion temperature. In all discharges the ion temperature profile is stiff by virtue of the fact that core (at r = and r/a =.4) and pedestal (at r/a =.8) temperatures are proportional to each other. The electron temperature deviates from the linear dependence at lower density (with higher temperatures) where both electron ion coupling and heat flux from NBI into the electron channel are reduced. The density profiles strongly peak at n e below m 3,this being correlated with the flattening of the central electron temperature. These data sets allow the following conclusions. First, the high ion core temperatures of the improved H mode can be explained by T stiffness alone. The roughly constant pressure at the pedestal top at fixed δ allows higher ion edge temperature at lower edge densities, and owing to T stiffness the core temperatures also increase, but the temperature gradient is still limited by profile stiffness, as is also shown in. In conclusion, the improved Hmodecanbedefinedintermsofthepeakeddensity profiles. Moreover, the confinement scaling given Fig. 8(a). The higher temperatures would not lead to an increase in the stored energy if the density profiles were stiff as well. The density peaking in the improved H mode at lower line averaged densities, however, leads to an increase in the plasma energy of up to 35%. This already takes into account the ion dilution by increasing Z eff at lower densities. The H factor compared with the H mode scaling ITER98(y,1), shown in Fig. 7, increases even up to 1.6, owing to the density dependence in the scaling law τ H-98P n.41 e in Section 3.1, τ E,th I 1. p P.56 h δ. n e ().3 n.17 e,sol, describes the good confinement of improved H mode as well. Both T i and T e profiles in the region.4 < r/a <.8 can successfully be modelled with the ITG/trapped electron mode (TEM) models, confirming the above conclusions. A prerequisite for the density peaking may be due to the zero magnetic shear in the centre. Substitution of ICRF minority heating for NBI heating leads to gradual suppression of fishbone activity and the reappearance of sawteeth. As long as sawtooth periods are very long compared with the energy confinement time, the central q profile remains flat and density peaking together with the plasma energy are kept high. Increasing ICRF power finally leads to higher sawtooth activity accompanied by the disappearance of strong density peaking and a high H factor. Stabilization of ITG turbulence by E r B shear flows does not seem to be so important, since toroidal plasma rotation, the main driver for the E r field (see also Section 4.), already decreases with the addition of 5% ICRH power. Nuclear Fusion, Vol. 41, No. 1 (1) 1377

10 O. Gruber et al. Ion temperature [kev] standard H mode L mode ITB (a) standard H mode scaled improved H mode L mode.5 1 Normalized effective radius Central ion temperature [kev] ITB with L mode edge (b) H mode L mode Ion temperature r =.6 a [kev] Figure 8. (a) Ion temperature profiles for different confinement modes. (b) Comparison of temperature profile peaking for different confinement modes. 4.. Ion ITBs with strong ion heating In ASDEX Upgrade we obtain ITBs for ions (strong ion heating with NBI), for electrons (strong electron heating with ECRF in current drive mode, see Section 5.) and reactor relevant ion and electron ITBs with T e T i 1 kev. Most of the ion ITBs are obtained by significant heating in the current ramp-up, predominantly ion heating by NBI of up to 7.5 MW. The H mode transition has to be avoided, as the edge bootstrap current, associated with the H mode, in combination with the large pressure gradients of the ITB results in the destabilization of externalkinksassoonastheedgeq becomes rational (see Section 7). The q profiles range from mainly negative shear with q min > 1.5 to sometimes monotonic shape with a broad zero shear central region with q min 1. Without external current drive, however, the negative central shear cannot be sustained for more than about 1 s. These ion ITBs exhibit strong internal pressure gradients which are characterized by an ITG length much smaller than that of the H or L mode plasmas governed by ITG turbulence and central ion temperatures in excess of 1 kev [5, 9]. This is shown in Fig. 8, which compares T i profiles and collects central and edge T i values of different confinement regimes. The large scatter in the ITB data suggests non-stiff behaviour or different confinement states. In the negative central shear regime, H mode was until recently prevented by limiter contact. This scenario has now been extended by upper single null (SNtop) configurations with an increased H mode threshold due to unfavourable ion B drift allowing higher power levels [9]. The T i and q profiles (from MSE measurements in combination with the equilibrium code CLISTE) of such a discharge at 5. and7.5mwofnbiarepresentedinfig.9.with increasing heating power the temperatures rise and an expanding radius of the ITB associated with an outward shift of the q min radius is obtained (H H-98P reaches 1.). The q min radius is always well aligned with the foot of the transport barrier (Figs 9 and 1). Shortly before the disruptive termination of these high power SNtop discharges an H mode transition occurs. The often disruptive termination of these high power SNtop discharges is caused by coupling of low (m, n) ideal internal modes (driven by large pressure gradients) to an external kink when q 95 approaches 4. This tendency is enhanced by H mode transition, since the associated edge bootstrap current favours external kinks (as mentioned before). Since in ASDEX Upgrade the conducting walls are far away from the plasma surface, the theoretical stability limit was calculated without a conducting wall with the GATO code [1], yielding a critical β N =1.7 in agreement with the experimental limit. With strong ion ITBs the turbulence in the core is 1378 Nuclear Fusion, Vol. 41, No. 1 (1)

11 Article: Overview of ASDEX Upgrade results T (kev) 15 1 T i # s 4 v (km/s) 3 v tor # s 1 1 n (m-3 e ) 15 T i # MW (flat-top) T (ECE) e 1 n e T [kev] 1 T (Th.sc.) e.5 1 ρ tor.5 1 ρ tor 5 5 MW (ramp-up) 8 q 6 4 l i = ρ tor 5 χ [m /s] AUG #171 χ i,nc χ i χ e Monte Carlo.5 1 ρ tor Figure 9. Temperature, density, toroidal rotation velocity and q profiles of an ion ITB L mode discharge. Core heat diffusivities are reduced to the ion neoclassical transport level, which is calculated from Monte Carlo simulations (1 MA,.5 T, 5 MW NBI). Standard neoclassical and potato orbit corrected [3, 31] predictions are given by dotted curves. suppressed, as is confirmed by reflectometry signals. A clear reduction of the turbulence is observed when the position of the reflection layer falls together with the foot of the developing transport barrier. Suppression of temperature gradient driven turbulence by the sheared radial electric field E r is plausible, since the E r B shearing rate greatly exceeds the linear ITG growth rates γ max,itg obtained from the GLF3 model [37] in the centre, ω E B = (RB p /B t ) (E r /RB p )/ r >γ max,itg (B p and B t are the poloidal and toroidal magnetic fields). Within the error bars ω E B is close to the growth rate at the foot of the barrier (r/a =.6, Fig. 1) [5]. The radial electric field E r = r p/(en e ) v t B p + v p B t (v p and v t are the poloidal and toroidal rotation velocities), is mainly determined by the toroidal q 5 Rates [1 Hz] T e (ECE) 5 MW (ramp-up) ρ tor AUG #13149 L mode ITB ρ tor 7.5 MW (flat-top) γ max ω ExB Figure 1. Temperature and q profiles during an ion ITB L mode discharge (1 MA,.5 T). ω E B shearing rate and linear ITG growth rate γ max are compared at the 7.5 MW level in flat-top. velocity, since the contributions of the poloidal rotation and the pressure gradient at the ITB itself are smaller. Here, shearing rates from the measured poloidal rotation are in general larger than those predicted by neoclassical theory, but the error bars on these measurements are rather large, too. Therefore neoclassical theory can be used to estimate the contribution from poloidal rotation. We have found that the Shafranov shift plays an important role in reducing the linear growth rates to levels where they are below the E B shearing rate [5]. The central Shafranov shift is much larger than in normal Nuclear Fusion, Vol. 41, No. 1 (1) 1379

12 O. Gruber et al. discharges because both the pressure and the safety factor are large on-axis with negative shear. For E B suppressed turbulence and strong ion ITBs, the ion heat transport is expected to be reduced to the neoclassical level. In fact the analysed ion heat diffusivities are even below those of standard neoclassical theory (Fig. 9). Standard neoclassical theory breaks down close to the axis because the radial size of the orbits is no longer small compared with the minor radius of the flux surface. These fat banana orbits are especially large in reversed shear discharges, and consequently the region in which standard theory breaks down is enlarged. As recent theories [38, 39] describing the influence of the large orbits on the neoclassical ion heat flux have yielded highly contradictory results, Monte Carlo simulations were performed to model the collisions, which consist of pitch angle scattering and a correction for momentum conservation [5]. Experimental and simulated heat conductivities show reasonable agreement inside and at the steep gradient region around r/a =.5 (Fig. 9) Reactor relevant combined ion and electron ITBs In a fusion reactor, where electrons are predominantly heated by fast particles, ITBs must be compatible with T e being equal to or even exceeding T i. Combining NBI and ECRH/ECCD during the current ramp-up, ITBs for both electrons and ions with T e T i 1 kev have been achieved [36, 4]. With NBI only an ion ITB with central T i 1 kev is formed (Fig. 11). Especially when central counter- ECCD is applied in a similar discharge, an electron ITB also occurs; this is manifested as a strong rise of the electron temperature in Fig. 11, with almost no change in the peaked profiles of ion temperature, density and toroidal rotation velocity. Despite the fivefold increase of the heat flux into the electrons in the plasma centre, neither the electron nor the ion thermal conductivity increases [3, 36]. The ITG growth rates obtained from the GLF3 model [37] for the two discharges are again below the poloidal shearing rates. Here the destabilizing effect for the ITG from the rising ratio of electron to ion temperature is largely compensated by the stabilizing Shafranov shift, which is larger in the case with ECRH because of the higher electron pressure in the centre. Since the ITB is located inside ρ =.4 and covers a small fraction of the plasma volume only, the maximum H L-89P is still about 1.8. q NBI only q min q T (kev)tite # Time (s) NBI + ctr-eccd q min T i q T e ECCD # Time (s) Figure 11. Comparison of reversed magnetic shear L mode discharges with NBI heating ( ion ITB) and with NBI together with central counter-eccd ( ion and electron ITB). Time traces of central and minimum q (q, q min), central temperatures and Mirnov signals for n =1modes(I p =1MA,B t =.45 T). 5. Fully non-inductively driven discharge scenarios As mentioned earlier, stationary tokamak operation requires the maintenance of the plasma current via a high fraction of internal diffusion driven bootstrap current supplemented by external noninductively driven currents. We have started to investigate current drive scenarios with on-axis current drive using nearly perpendicular NBI and ECCD at a level of I p = 4 ka (q 95 =9) High β p > 3 discharges with dominant bootstrap current drive at the Greenwald density To achieve a high bootstrap current, the poloidal β is maximized by operating at a low plasma current [41] of 4 ka and a high NBI power of 1 MW. Fully non-inductively driven H mode discharges at high β p > 3 and densities at the Greenwald density have been achieved. As shown in Fig. 1, these discharges start from an ohmic plasma and the NBI powerisrampedupinstepsof.5mwto1mw. H mode transition already occurs when the first 138 Nuclear Fusion, Vol. 41, No. 1 (1)

13 Article: Overview of ASDEX Upgrade results neutral beam source is switched on. The density rise initially is facilitated by edge gas fuelling, while above 7.5 MW NBI fuelling alone is sufficient to maintain the density at or above n GW. The current in the ohmic transformer is kept constant after reaching the maximum NBI power of 1 MW. It is obvious here that an ion ITB at ρ.5 is triggered by the redistribution of fast particles owing to fishbones (see Section 7.) [9]. This is confirmed by the reduction of the core ion heat diffusivity as calculated using the transport code ASTRA. Despite the fourfold increase of the heating power, χ i is lowest in the high power phase except in the very centre of the plasma, where a small temperature plateau is formed. The minimum value at the position of the ITB is.5 m /s and is well below the neoclassical values with and without finite orbit width corrections (see Section 4. [38, 39]). The appearance of this ITB coincides with the extinction of the sawtooth activity and as with the improved confinement H mode (Section 4.1) reconnecting (1,1) fishbones remain in the plasma, indicating that a stationary current profile with a q 1 surface is present. This agrees with the current or q profiles inferred from MSE measurements by means of the CLISTE equilibrium code, showing a broad minimum of q 1 in the centre. It agrees reasonably well also with current diffusion simulations using ASTRA and assuming neoclassical conductivity (Fig. 1). The large bootstrap current fraction reaches f BS =.56 ± 1% (error estimated from 1% error in T e and 1.5 <Z eff < 3) and tends to cause negative central shear in the initially monotonic q profile. The neutral beam current drives the rest of the plasma current. The plasma current slightly decreases owing to the flux loss by the somewhat decreasing vertical field (β decreases slightly) and the central fishbone activity. 5.. ECCD discharges and electron ITB (T e 15 kev) with reversed shear For the investigation of electron heating and current drive, ECCD has been applied in L mode plasmas with low current and density [9]. A low current permits a large fraction of the ohmic current to be replaced by the externally driven current, while the low density optimizes the current drive efficiency, which scales with T e /n e. Starting from an ohmic plasma at n e m 3, the ECR waves of three gyrotrons (1. MW) were centrally launched in either co- or counter-current direction, Figure 1. Time evolution of a fully non-inductively driven H mode discharge with integrated high performance with respect to β, n e/n GW and H factors. The OH transformer current is kept constant from.5 s onwards. Temperature and q profiles at.6 s show an ion ITB. while the remaining.4 MW was used for heating. With the onset of ECRH, T e rises sharply in both cases, but the temperature profiles are distinctly different. While with co-eccd the temperature gradient in the plasma core is nearly constant (T e () 1keV),anelectronITBdevelops in the counter-eccd discharge in the presence of a reversed magnetic shear inside ρ t.5. The electron temperatures go up to 15 kev in the negative shear region, but a non-thermal part may be included in the ECE measurements (ions remain cold below 1 kev). Again the position of the barrier foot is at the q min radius. For electron ITBs, reversed shear seems to be a prerequisite [36, 4], in agreement with the Nuclear Fusion, Vol. 41, No. 1 (1) 1381

14 O. Gruber et al. theory of ETG driven turbulence and its possible stabilization []. This difference is also reflected in the MHD behaviour. For co-eccd, continuous sawtooth activity suggests the existence of a q = 1 surface throughout the discharge, in agreement with ASTRA simulations and MSE measurements. Sawteeth disappear in the case of counter-eccd. Instead, continuously irregular (,1) modes with a high growth rate cause T e collapse. These modes are located near the maximum of the temperature gradient, suggesting a pressure driven ideal mode, which might be avoided by a smoother deposition profile and lower electron pressure gradients. As two q = surfaces are present in the plasma [9], the mode may be of double ideal kink structure. Deposition, propagation and absorption of the electron cyclotron were calculated with the TOR- BEAM code [43], predicting in both cases a driven current of about 3 ka. This is confirmed by current diffusion simulations predicting for the co-eccd case a current drive fraction of f ECCD =.8 and a nearly fully non-inductive drive (f BS =.1) [44]. For the counter-eccd case the strong MHD modes and huge current density gradients did not permit a quantitative assessment. 6. Stabilization of NTMs by ECCD NTMs are the main β limiting MHD events in conventional H mode scenarios with positive shear in ASDEX Upgrade. Their onset β N scales proportionally to normalized gyroradius ρ without significant dependence on collisionality ν over a wide range of ν, as already reported [3, 45]. The database supporting this conclusion has been broadened by including NTMs occurring during pellet fuelled discharges at substantially higher ν,aswellasbycomparison with NTM data from the old ASDEX tokamak [11]. The data points now span more than an order of magnitude variation in collisionality. This onset behaviour is consistent with the stabilizing terms of the polarization current model and, if the heat flux limit for parallel heat flow in the islands is taken into account, may also be described with the χ /χ model in the generalized Rutherford equation describing the evolution of the island width. In a density range approaching the Greenwald limit, NTMs can be suppressed by increasing ν [46]. This threshold collisionality was found to be well described again by the expression given from the polarization current model. pert. magn. field [a.u.] #933 b 3/ b /1.4 time (s) w/a / / t / τ Figure 13. Temporal sequence of (3,) and (,1) NTMs shown by the perturbed magnetic field. After the (,1) mode grows to a sufficient level, the (3,) mode vanishes, in agreement with theoretical modelling (right panel). In ASDEX Upgrade, NTMs may occur in a sequence of modes that descends from close to q =1 to a certain outermost surface determined by the local onset conditions. Usually, the final mode is a (3,) or, at lower densities, a (,1) NTM [11, 1]. An important observation has been that NTMs of different helicities never coexist in a stationary state. Figure 13 shows the change from a (3,) to a (,1) mode, which was modelled with a non-linear cylindrical tearing mode code [11, 1], also shown in Fig. 13. Stabilization of NTMs by using ECRH or ECCD is of great importance for next step experiments, which already suffer at lower β N s from NTMs because of the onset scaling given above. On ASDEX Upgrade, we inject up to 1. MW of ECRH (14 GHz, second harmonic X mode absorption on the HFS) into the plasma [47, 48]. By changing the injection angle, we vary the amount of driven current, while feedforward scans in B t of 5 1% within 1 s, i.e. slow compared with the island growth time, are used to fine tune the radial position of the absorption location. With about 1% of the total heating power, corresponding to 15 ka of helical current driven within the island, a (3,) NTM at β N =.5 can be completely stabilized by co-eccd (Fig. 14). The experimental results are in qualitative agreement with the generalized Rutherford equation and in quantitative agreement with a two dimensional nonlinear cylindrical tearing mode code [49], the main stabilizing effect being the local helical current generated by ECCD. There is no significant difference between AC and DC stabilization. This is due to the fact that the fast electrons generated by ECCD quickly equilibrate along the field lines so that power deposited with finite width near the X point does not generate substantial current within the island. Pure ECRH is much less effective in stabilizing the mode. R 138 Nuclear Fusion, Vol. 41, No. 1 (1)

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