overview of asdex upgrade results - ucla physics & …jenko/papers/nfpap.pdf · m. kaufmann, a....

Overview of ASDEX Upgrade results

O. Gruber, R. Arslanbekov, C. Atanasiua, A. Bard, G. Becker, W. Becker,M. Beckmann, K. Behler, K. Behringer, A. Bergmann, R. Bilato, D. Bolshukin,K. Borrass, H.-S. Bosch, B. Braamsb, M. Brambilla, R. Brandenburgc,F. Braun, H. Brinkschulte, R. Bruckner, B. Brusehaber, K. Buchl, A. Buhler,H. Burbaumerc, A. Carlson, M. Ciric, G. Conway, D.P. Coster, C. Dorn, R. Drube,R. Dux, S. Egorovd, W. Engelhardt, H.-U. Fahrbach, U. Fantze, H. Faugel,M. Foleyf , P. Franzen, P. Fug, J.C. Fuchs, J. Gafert, G. Gantenbeinh, O. Gehre,A. Geier, J. Gernhardt, E. Gubanka, A. Gude, S. Gunter, G. Haas, D. Hartmann,B. Heinemann, A. Herrmann, J. Hobirk, F. Hofmeister, H. Hohenocker,L. Horton, L. Hug, 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. Kyriakakisi, K. Lackner, P.T. Lang,R.S. Lang, M. Laux, L. Lengyel, F. Leuterer, A. Lorenz, H. Maier, K. Mank,M.-E. Mansoj, M. Maraschek, K.-F. Mast, P.J. McCarthyf , D. Meisel, H. Meister,F. Meo, R. Merkel, V. Mertens, J.P. Meskath, R. Monk, H.W. Muller, M. Munich,H. Murmann, G. Neu, R. Neu, J. Neuhauser, J.-M. Noterdaeme, I. Nunesj,G. Pautasso, A.G. Peeters, G. Pereverzev, S. Pinches, E. Poli, R. Pugno,G. Raupp, T. Ribeiroj, R. Riedl, S. Riondato, V. Rohde, H. Rohr, J. Roth,F. Ryter, H. Salzmann, W. Sandmann, S. Sarelma, S. Schade, H.-B. Schilling,D. Schlogl, K. Schmidtmann, R. Schneider, W. Schneider, G. Schramm,J. Schweinzer, S. Schweizer, B.D. Scott, U. Seidel, F. Serraj, S. Sesnic, C. Sihler,A. Silvaj, A. Sips, E. Speth, A. Stabler, K.-H. Steuer, J. Stober, B. Streibl,E. Strumberger, W. Suttrop, A. Tabasso, A. Tanga, G. Tardini, C. Tichmann,W. Treutterer, M. Troppmann, N. Tsoisi, W. Ullrich, M. Ulrich, P. Varelaj,O. Vollmer, U. Wenzel, F. Wesner, R. Wolf, E. Wolfrum, R. Wunderlich,N. Xantopoulosi, Q. Yu, M. Zarrabian, D. Zasche, T. Zehetbauer, H.-P. Zehrfeld,A. Zeiler, H. ZohmMax-Planck-Institut fur Plasmaphysik, Euratom–IPP Association,Garching, Germany

a Institute of Atomic Physics, Bucharest, Romaniab Courant Institute, New York University, New York, N.Y., United States of Americac Technical University of Vienna, Vienna, Austriad Efremov Institute, St. Petersburg, Russian Federatione University of Augsburg, Augsburg, Germanyf University College, Cork, Republic of Irelandg Academia Sinica, Hefei, Chinah Institut fur Plasma Forschung, University of Stuttgart, Stuttgart, Germanyi NSCR Demokritos, Athens, Greecej Centro de Fusao Nuclear, Lisbon, Portugal

Nuclear Fusion, Vol. 41, No. 10 c©2001, IAEA, Vienna 1369

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 tem-

perature gradient (ITG) driven turbulence. ECRH experiments indicate that electron temperature

(Te) 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 < 0.45) show strongly

improved confinement up to the Greenwald density nGW and therefore higher β values, owing to

increasing pedestal pressure, and H mode density operation extends above nGW . Density profile

peaking at nGW 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 HL-89P βN ≈ 7.2 at δ = 0.3 for

five confinement times, limited by neoclassical tearing modes (NTMs) at low central magnetic shear

(qmin ≈ 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 (qmin > 1) and L mode edge —

achieved HL-89P ≤ 2.1 and are limited to βN ≤ 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 Te ≈ Ti ≈ 10 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 ne = nGW , high βp = 3.1, βN = 2.8 and HL-89P = 1.8, which developed ITBs with

qmin ≈ 1. Central co-ECCD at low densities allows a high current drive fraction of >80%, while

counter-ECCD leads to negative central shear and formation of an electron ITB with Te(0) > 12 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 βN

for NTM is proportional to the normalized gyroradius ρ∗. Complete NTM stabilization was demon-

strated at βN = 2.5 using ECCD at the island position with 10% 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 pro-gramme has been largely focused on:

(i) Confinement and performance related physicsin the ITER baseline scenario, the ELMYH mode near operational limits, with closeinteraction between edge and core plasma;

(ii) Investigation of scenarios and physics ofadvanced tokamak plasma concepts with inter-nal transport barriers (ITBs) leading toenhanced performance and possibly stationaryoperation;

(iii) MHD stability and active stabilization of β

limiting instabilities as well as avoidance andmitigation of disruptions;

(iv) Edge and divertor physics in these high power,high confinement regimes, with the aim ofidentifying and optimizing power exhaust andparticle control (ash removal).

The similarity of ASDEX Upgrade to ITER inrespect of poloidal field coil system and divertorconfiguration makes it particularly suited to test-ing control strategies for shape and plasma perfor-mance. NBI was available with injection energiesof up to 100 keV for deuterium (with 20 MW),which allowed studies of the influence of heat deposi-tion on transport and of fast particle effects. Minor-ity heating with the ICRH system (up to 5.7 MWcoupled) was used for central heating at high den-sities, and to study the influence of deep particlerefuelling and toroidal rotation by substituting NBI.

1370 Nuclear Fusion, Vol. 41, No. 10 (2001)

Article: Overview of ASDEX Upgrade results

The ECRH system (four gyrotrons with a coupledpower of 1.4 MW for 2 s) allowed pure electron heat-ing, transport studies and feedback stabilization ofneoclassical tearing modes (NTMs). Provisions forcurrent drive and for active control of current pro-files in advanced scenarios were available with nearlyperpendicular NBI and ECRF using steerable mir-rors (up to 350 kA driven on-axis). Experimentsusing on-axis fast wave or off-axis mode conver-sion current drive with the ICRF system have beenstarted.

The present version of the closed divertor II,LYRA (vertical target plates including a roof baf-fle between them, cryopump), which is rather sim-ilar to the ITER-FEAT reference design, allows astrong reduction of heat flux to the target platesand is capable of handling heating powers of upto 20 MW or P/R of 12 MW/m. The observedstrongly reduced, distributed power flux to the sur-rounding structures during both ELM and ELM freephases is in agreement with B2-Eirene simulations[1]. The study of reactor compatible materials wascontinued by covering parts of the low field side(LFS) heat shield with tungsten to reduce carboninflux [2].

High shaping capability with elongations κ ≤ 1.8and triangularities δ of up to 0.45 at the separatrixallowed study of the influence of plasma shape onperformance and operational limits. Such equilibriahave been run with the outer strike point being onthe top of the roof baffle or on the outer vertical tar-get to obtain higher pumping speeds. Additionally,the similarity in cross-section to other divertor toka-maks is important in determining size scalings forcore, edge and divertor physics. This collaborativework, including extrapolation to ITER parameters,has continued and was even enhanced in the newJET operation under the European Fusion Develop-ment Agreement (EFDA). The promising method ofrefuelling by pellet injection yielded first results witha new high field side (HFS) pellet launcher capableof injection speeds of up to 1000 m/s.

The following sections report on recent progressmade in ASDEX Upgrade experiments since the lastIAEA conference [3]. First, ion and electron temper-ature profile stiffness (Section 2 [4, 5]) and high den-sity H mode operation (confinement, density profilepeaking, HFS pellet injection and tolerable type IIELMs) (Section 3 [4–6]) are described. Cold pulseand impurity transport studies are not reported here[7, 8]. The next sections are devoted to advancedscenarios such as improved H mode, ion ITBs and

reactor relevant combined ion and electron ITBs(Section 4 [4, 5, 9, 10]), and fully non-inductivelydriven discharges at high βp and with ECCD exhibit-ing electron ITBs (Section 5 [9, 10]). Finally, stabi-lization of NTMs by ECCD (Section 6 [11, 12]), MHDin advanced scenarios (Section 7 [10]) and opera-tion with high Z wall coatings (Section 8 [2]) arepresented.

2. Temperature profile stiffness

Understanding of transport in the conventionalH mode scenario is still an active field of researchbecause it determines the prediction and design ofnext step devices. As reported earlier [3, 13, 14], iontemperature (Ti) profiles in conventional H mode dis-charges of ASDEX Upgrade are generally ‘stiff’: tem-perature profiles are limited by a maximum valueof ∇T/T , the inverse temperature gradient lengthLT (Fig. 1). At medium and high densities the elec-tron temperatures (Te) show the same behaviour,since they are closely coupled to Ti and heated aswell by NBI. Turbulence driven by the ion temper-ature gradient (ITG) is believed to be the main ori-gin of turbulence causing the anomalous transportthrough the ion channel [15–17]. It leads to limita-tion of Ti at a critical value of ∇T/T as observed.The transport properties of temperature gradientdriven turbulence can be expressed for the heat dif-fusivity by two additive terms. One term representstransport without temperature gradient driven tur-bulence, for instance neoclassical transport for ions,while the other represents the transport by tempera-ture gradient driven turbulence, which is influencedby magnetic and rotational shear, effective ion massand density gradient length. This term is equal tozero for ∇T/T < (∇T/T )c but increases stronglywhen ∇T/T > (∇T/T )c and may saturate. Thentransport is high and keeps the profiles close to(∇T/T )c. Gyro-Bohm scaling χgB ∝ T 3/2 of thisterm causes an increase of stiffness with tempera-ture. These considerations are not necessarily validinside the sawtooth inversion radius, where profilesmay be limited by the MHD activity, or at the edge,where the temperatures may be too low. The pro-file stiffness reflects a strong coupling between coreand H mode pedestal temperatures or, equivalently,the fact that the profiles plotted on a logarithmicscale are shifted vertically according to their edgetemperature (Fig. 1).

Nuclear Fusion, Vol. 41, No. 10 (2001) 1371

O. Gruber et al.

0.2

0.4

0.6

0.81.0

3.0

Ti [

keV

]

0 0.2 0.4 0.6 0.8 1ρ

po l

5.0#10761

Ti

Figure 1. H mode Ti profiles are stiff during a density

ramp-up (lines) and remain fixed for on- and off-axis heat

deposition (symbols, same heating power) with temper-

ature gradient length T/∇T ≈ R/5.

2.1. Temperature profile stiffnessin NBI heated H modes

Profile stiffness in NBI heated H modes was con-firmed by comparing the influence of rather differentheat deposition profiles using 60 and 93 keV neutralbeams at high plasma densities (ne ≈ 1020 m−3):measured temperature profiles are unaffected. Thereaction of the transport coefficients to these differ-ent heating power profiles is as expected. For thedeeper deposition with 93 keV beams, χeff in thecore increases by a factor of almost 2 in relationto the more off-axis deposited power with 60 keVbeams. At these high densities, owing to the strongelectron–ion coupling, only an effective heat diffusiv-ity can be derived from power balance. Obviously,the heat conductivity adjusts itself to maintain theobserved stiff temperature profiles and ceases to bea useful description of transport behaviour.

In addition to these studies, a database includ-ing profile data from about 30 H mode dischargeswith NBI [5, 17] indicates that Ti and Te profilesare quite stiff at medium and high densities. Simu-lations of these discharges using three models basedon ITG and including trapped electron mode physicsand E×B shear flow — IFS/PPPL [15], GLF23 [18]and Weiland [19] — agree well with these data andconfirm the ITG turbulence as the driving mecha-nism. The boundary condition for the temperature

(applied at r/a = 0.8) and the density profiles usedin the simulations are taken from the experiment.

2.2. Electron temperature profile stiffnesswith strong electron heating

Recent calculations [20] indicate that electrontemperature gradient (ETG) turbulence is able todrive a large electron heat transport because largecells, ‘streamers’, can develop. The properties of tem-perature gradient driven turbulence for ions andelectrons are qualitatively similar, as has been con-firmed in ASDEX Upgrade discharges with Te > Ti

using ECRH in steady state and power modulationexperiments [4].

Central heating using either 0.8 or 1.6 MW ofECRH power at different plasma currents resulted insimilar shapes of the electron temperature profiles,which are shifted according to their edge tempera-tures in the logarithmic scale of Fig. 2. The shadedarea indicates a region where the slope of the pro-files, i.e. ∇T/T , is the same. In the centre the slopechanges and is different for each profile. This regionextends beyond the inversion radius of the sawteeth.Obviously, the heat flux to the electron channel is nothigh enough to reach (∇T/T )c. Towards the edge theprofiles gradually deviate from the constant slope,which seems to occur for Te values below 600 eV. Inthe stiff region the value of∇T/T does not depend onIp, which implies that the well known improvementof confinement (and temperature increase) with Ip isprovided by higher edge temperatures.

Electron heat diffusivities of these discharges frompower balance analyses — normalized to the gyro-Bohm scaling T 3/2 — exhibit the expected prop-erties, namely small χ/T 3/2 values for ∇Te/Te <

7 m−1 and a strong increase above this threshold[4]. The normalized diffusivities level off at differ-ent values as the stiffness increases with higher tem-peratures. This is also in qualitative agreement withETG calculations, indicating that the appearance ofstreamers requires a finite value of temperature [20].The results from a similar discharge in hydrogen indi-cate that the electron transport is not sensitive to theion mass. Complementary to these experiments withcentral ECRH, discharges with off-axis ECRH indi-cate that transport indeed decreases in the regionbetween the plasma axis and ECRH deposition, con-sistent with the decrease of ∇T/T to below thecritical value [4].

These steady state analyses were supplementedby ECRH power modulation experiments confirming



0.1

1

0 0.2 0.4 0.6 0.8 1

1.6 MW 1 MAOH 1 MA1.6 MW 0.6 MAOH 0.6 MA

Te

[keV

]

ρtor

6#13558#13654

not stiff: T e too low

not stiff:q e too low

stiff region

Figure 2. Stiff Te profiles for pure ohmic heating

and additional central ECRH (1.6 MW) at two plasma

currents.

the electron temperature stiffness. These investiga-tions further revealed that changes in transport foroff-axis ECRH cases are caused by a change of diffu-sivity and cannot be attributed to convection [4]. Inthe stiff region, the heat pulses propagating outwardsand inwards yield different values for the derivedχHP by virtue of the fact that heat pulses tendto increase or decrease ∇T/T locally, depending ontheir direction of propagation. The χHP values areabove the χ’s derived from power balance (χPB ), andtheir ratio χHP/χPB strongly increases with temper-ature, depending again on the direction of the pulses.This ratio characterizes the stiffness of the profile.Accordingly, differences in the heat pulse propaga-tion between deuterium and hydrogen discharges atequal discharge parameters, but at different resultingtemperatures, can be reconciled by the temperaturedependence of the stiffness.

3. High density H mode operation

3.1. Influence of non-stiff densityprofiles and triangularityon energy confinement

To achieve high fusion yield by means of a highcentral density and low power load in the diver-tor with sufficient SOL density, a fusion reactor hasto operate at high densities close to the Greenwald

density nGW . But confinement degrades at highline averaged densities ne towards nGW , while itimproves with ne at low densities. This is a con-sequence of temperature profile stiffness (Tcore ∝Tped), the nearly fixed pedestal pressure of theedge H mode barrier limited by MHD stability andincreasing with triangularity, and the fact that den-sity profiles tend not to be stiff. At low densities morepeaked density profiles are observed, while increas-ing the density by gas puffing leads to density profilebroadening, and the edge density strongly increasesin relation to the line averaged density [3]. Corre-spondingly, the stored energy even decreases withincreasing ne and stiff T profiles, in contrast tothe ITER confinement time scalings containing apositive contribution from ne.

For a given shape of the density profile, the plasmaenergy is expected to be proportional to the pedestalpressure. This is indeed confirmed for ne < 0.8nGW

by 300 representative discharges, where a strong cor-relation between electron pedestal pressure pe,ped (atρ = 0.8) and thermal plasma energy Wth is seen. Alinear regression applied to this data set yields:

Wth ∝ p0.75e,pedP 0.14

h

with a low RMSE value of about 10% [4].This representation provides a unification of dis-

charges with different ELM types, Ip’s and triangu-larities (0.1 < δ < 0.4), which only act on confine-ment via the edge pressure, namely as

pe,ped ∝ I1.4p P 0.2

h δ0.3.

The dependence upon Ip does not quite reach I2p

as expected from ballooning mode stability, whichmight be due to changes of other quantities inthe edge region influencing stability, e.g. magneticshear.

An overview of the confinement data is pro-vided in Fig. 3 for τE normalized by the IPB98(y)ELMy H mode scaling, the HH-98P factor. The pos-itive density dependence in IPB98(y) accentuatesthe confinement degradation at high densities by aneven stronger decrease of the H factor. The ben-efits resulting from more highly triangular plasmashapes are first manifested in enhanced confinementwith HH-98P ≈ 1 at the Greenwald density due tohigher pedestal pressures. Then back-transition intothe L mode at high densities is shifted to densitiesabove nGW , thus substantially extending the opera-tion regime of the H mode, but it also exhibits a largescatter at a given value of ne/nGW . This scatter canbe attributed to the non-stiff density profiles, which,


O. Gruber et al.

n / n

τ / τ

(ITER

H-9

8P)

δ = 0.29δ = 0.18

0.0 0.2 0.4 0.6 0.8 1.0 1.20.0

0.5

1.0

1.5

2.0

δ = 0.35

e GW

ITER FEAT

Figure 3. Normalized thermal energy confinement ver-

sus normalized line averaged density ne at different

triangularities δ.

besides being more peaked at low densities than athigher ones, can vary according to the discharge sce-nario. This is taken into account in a regression byincluding the central density ne(0) and the densityin the scrape-off layer averaged over 6 cm startingfrom the separatrix outwards, ne,SOL, yielding thefollowing expression:

τE ,th ∝ I1.0p P−0.56

h δ0.2ne(0)0.3n−0.17e,SOL

(RMSE: 9%)[4].

This expression reflects the good confinement forpeaked density profiles and confinement degradationwith gas puffing represented by ne,SOL. It is also inagreement with the Wth and pe,ped scalings givenbefore. The low RMSE is provided by the fact thatthis variable set takes the physics background intoaccount.

3.2. Density profile peakingwith constant gas puffing

Often in experiments the density is ramped upquickly to the anticipated value by strong gas puff-ing, and the high neutral gas flux causes high edgedensities, with their negative effect on global confine-ment. With ‘soft’ density control or low external gasflux, the line averaged density slowly increases on

2.0

0.0 0.2 0.4 0.6 0.8 1.0 1.2normalized poloidal flux radius

0

0.5

1.0

1.5

elec

tron

den

sity

(10

m

) 2.5 MW ICRH

2.5 MW NBI

2.5 MW ICRH5 MW NBI

5 MW NBI

7.5 MW NBI

20-3

ASDEX Upgrade #13660

Figure 4. Comparison of electron density profile peak-

ing in H mode at the Greenwald density with moderate

gas puffing using different heat deposition scenarios.

a timescale of a second (many energy confinementtimes) and can exceed the Greenwald density [4].After an initial decrease the confinement time itselfagain increases and HH98-P = 1 is restored. Thedensity profile shape broadens first because of theincreasing edge density, but on the longer timescalethe density profile significantly peaks, as shown inFig. 4 for 5 MW NBI. In this phase the pedestaland SOL regions, which are determined by the par-ticle source, remain unchanged, as also do the stifftemperature profiles. As explained above, the non-stiff density profiles together with stiff temperatureprofiles are the reason for the decrease of confine-ment observed at the beginning of the gas pulseand the confinement enhancement during the peak-ing phase. Finally, good confinement is lost owingto high central radiation caused by loss of sawteethleading to accumulation of heavy impurities in thecentre. Before accumulation Zeff is below 1.3 in thecore.

The operational window for the density profilepeaking seems to be limited by a correlation betweenheat and particle transport. The heat flux waschanged in the experiments by increasing the heat-ing power or its deposition profile by combining NBI(slightly hollow heating profile) and ICRF minorityheating (central deposition). This causes the regionwith density peaking to shrink with increasing heat-ing power, and the density profiles even become flatwhen central heating is increased (Fig. 4). In the lat-ter case the particle source with NBI is not changedand therefore cannot be the reason for the loss ofdensity peaking. The increase of central power depo-sition in both cases, with its increase in transport



at limited ∇T/T , is responsible for the density flat-tening. In summary, an inward particle pinch isinvolved (Ware pinch seems to be sufficient), whichis cancelled by increasing outward diffusion at highercentral heating.

These observations are supported by applyingECRH in various types of discharge which react tothis additional heating with flatter density profiles.One may speculate that ECRH increases the elec-tron heat flux and thus the ETG turbulence level.This density ‘pump-out’ was observed earlier in otherdevices with ECRH [21].

3.3. Advanced HFS pellet injection

With injection of frozen hydrogen pellets fromthe magnetic LFS, the fuelling efficiency in stronglyheated plasmas was found to be poor. In contrast,pellet injection from the HFS, first performed onASDEX Upgrade, was found to give significantlyhigher efficiency, resulting in H mode discharges wellabove the Greenwald density with high confinementproperties [22]. This is due to the fast radial driftin the toroidal field gradient of the high β plasmoidthat forms around the ablating pellet, which trans-fers the pellet particles towards the plasma centre[23]. In both injection scenarios performance is lim-ited by prompt convective particle and energy lossesresulting from pellet induced ELM bursts within10–20 ms after injection.

This problem should be overcome by applyinghigher launch velocities to deposit the pellet beyondthe ELMing region at the edge. However, stronglycurved guiding tubes with turning points of the cur-vature and the fragile nature of cryogenic deuteriumhave up to now imposed a speed limit of ≤240 m/s.After a systematic study we have designed a newHFS pellet injection system using a looping geome-try, where the pellets are ejected away from the toka-mak device by the accelerating centrifuge and thenguided back to the torus HFS by a roughly ellipticaltrack [24]. In first proof of principle experiments witha preliminary looping system, roughly 50% of the pel-lets with a velocity of 560 m/s could be injected intothe plasma with only a small mass loss [6]. At highervelocities of up to 1200 m/s, only fractions of thepellet reached the plasma. At a velocity of 560 m/s,a deeper penetration (an increase from 0.165 to0.215 m) and particle deposition towards the cen-tre were found, mitigating the fast convective lossesconnected with the ‘pellet ELMs’ (Fig. 5), but thepellet induced ELMs during the first 15 ms are also

560 m/s

240 m/s

50 ms

n e

10

m

19-3

Figure 5. Density decay after HFS pellet injection at

two velocities and ELM activity (Dα in the divertor). The

initial density decay time increases from 48 (v = 240 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 sys-tem should enable us to define the demands of pelletinjection systems for future large scale devices.

3.4. Tolerable type II ELMsclose to the Greenwald density

H modes with type I ELMs exhibit the best con-finement performance, but the concomitant largepower load during the ELM activity may not beacceptable in a reactor. Type III ELMs show almostno additional heat load on targets, but they entailconfinement degradation. Small high frequency (orgrassy) ELMs of type II were recently observed inDIII-D [25] and JT-60U [26], whereas the ‘enhancedDα’ regime was reported from Alcator C-Mod [27].These regimes were obtained for q95 > 4.5 andδ > 0.4, but only at medium densities.

On ASDEX Upgrade, type II ELMs were iden-tified in a comparable operation region close to adouble null configuration, but in contrast to the ear-lier experiments at high densities at the Greenwalddensity [4]. This new result is quite attractive fora future reactor. In addition the confinement inthis type II ELM regime is almost as good as withtype I ELMs and is significantly higher than fortype III ELMs, providing τE/τH98-P ≈ 0.95. The


O. Gruber et al.

O

I

type III ELMs4.005 4.015 4.025 4.035 time (s)

Hα outer divertor

O

I

type II ELMs Hα outer divertor

4.04 4.05 4.06 4.07 time (s)

O

I

type I ELMs

4.855 4.865 4.875 4.885

Hα outer divertor

time (s)

MW/m

01

2

3

2

0.5

0

1V

0.5

0

1V

0.5

0

1V

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 Hα traces.

thermography data of the divertor target plates showthat the peak power load is strongly reduced withtype II ELMs, reaching at most 2 MW/m2 almostwithout peaks on the outer plate, in comparison with5 MW/m2 for type I ELMs (Fig. 6). The power fluxon the inner plates is always very low under theseconditions.

Magnetic measurements indicate that type IIELMs often have a precursor at 15–30 kHz, indi-cating that they are indeed MHD events. Their sig-natures are different from those of type III ELMs,and they provide better confinement because theypreserve the high edge pressure.

4. Advanced tokamak discharges

Advanced tokamak scenarios aim at an oper-ating regime with improved confinement and sta-bility properties (high normalized beta βN =β/(Ip/aBt) > 3), thus allowing a reduced size toka-mak reactor at lower plasma current to be operatedin steady state [28–32]. ITBs, which are characterizedby large pressure gradients in the region of reduced

plasma turbulence, intrinsically drive bootstrap cur-rents, which can constitute a significant part of theplasma current. Steady state operation requires thatthe inductively driven plasma current be completelyreplaced by external non-inductively driven currents(in addition to the bootstrap current), which, how-ever, should be kept low to minimize the recircu-lating energy. On ASDEX Upgrade we have usednearly perpendicular NBI and ECCD for on-axis cur-rent drive up to now. Two advanced scenarios wereinvestigated by modifying the current density pro-file by means of early heating in the current ramp.Both involve peaked profiles of density and toroidalrotation velocity. The one gives improved core con-finement in combination with an H mode edge bar-rier and a flat central q profile, offering station-ary, inductively driven improved H mode operation[32–34]. This scenario may allow ignition even inthe ITER-FEAT device. Secondly, ITBs with mainlyreversed magnetic shear and qmin values above 1.5exhibit, at least in combination with an L mode edge,steep pressure gradients in the barrier region andtherefore bootstrap currents exceeding 50% of theplasma current [32, 34–36]. A combination of ITBand H mode edge is needed for true steady state,non-inductively driven tokamak operation combiningimproved performance and high bootstrap currentfraction fbs = Ibs/Ip ∝ βNq95

√A, at high q95 ≥ 4

and βN > 3.5.

4.1. Improved H mode

The improved H mode regime is obtained with lowpreheating in the current ramp-up and subsequentpower increase causing a transition into the H mode.A stationary state is obtained for up to 40 energyconfinement times and several internal current diffu-sion times and is limited only by the available pulselength [3, 32, 33]. Continuously reconnecting (1,1)fishbones clamp the current profile in the core andlead to stationary central zero shear with qmin ≈ 1(see Section 7.1) [33–35].

The density in this scenario was extended up toslightly below half of the Greenwald density to inte-grate power exhaust and He ash removal by appro-priate divertor conditions. With gas fuelling, thethreshold heating power for ITB formation had tobe raised by ∼50%. A density increase caused byimproved core particle confinement at more triangu-lar plasma shapes does not change the ITB onset con-ditions and allows higher βN values up to 2.6 with-out destabilizing NTMs. The confinement increasesto HH98-P = 1.7 or HL89-P = 3, this being largely



0 1 2 3 40

2

4

6

8

10

12

T at r/a = 0.4

0 0.5 1 1.5 2 2.50

2

4

6

0

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2

central density / density at r/a = 0.8

2 3 4 5 6 7 8

0.6

0.8

1.0

1.2

1.4

1.6

1.8

2.0H factor ITER98-P

3 4 5 6 7 8 9

Tbehaviour

hysteresis for gas puff ?

n [10 m ]19 -3en at r/a=0.8 [10 m ]19 -3

e

central T e

e

i

T at r/a=0.8 [keV]e

central T i

T at r/a = 0.4i

T at r/a=0.8 [keV]i

[keV]

Figure 7. Comparison of temperature and density profiles and HH-98P factor of standard

(blue symbols) and improved (red symbols) H mode (density scan with 0.3 < ne/nGW < 0.6

at Ip = 1 MA, PNBI = 5 MW, δ = 0.2).

due to the increasing pedestal pressure with higherδ [32, 34]. The more effective impurity screeningat higher edge densities reduces Zeff from 3 (at0.35nGW ) to 2 in the plasma centre. No temporalaccumulation of impurities is observed, but impuri-ties show centrally peaked densities consistent withneoclassical predictions. First measurements point tosufficient He exhaust in this scenario.

Detailed studies have shown that the confinementphysics in this regime can be largely unified with thatof the standard H mode [5]. In Fig. 7 the data of adensity scan 0.3 < ne/nGW < 0.6 with the same Ip

(1 MA), the same NBI heating power (5 MW) andlow triangularity δ = 0.2 are shown. The improvedH mode shots have in general a lower line averageddensity and higher central as well as edge ion tem-perature. In all discharges the ion temperature pro-file is stiff by virtue of the fact that core (at r = 0and r/a = 0.4) and pedestal (at r/a = 0.8) temper-atures are proportional to each other. The electrontemperature deviates from the linear dependence atlower density (with higher temperatures) where bothelectron–ion coupling and heat flux from NBI intothe electron channel are reduced. The density pro-files strongly peak at ne below 4.5 × 1019 m−3, thisbeing correlated with the flattening of the centralelectron temperature.

These data sets allow the following conclusions.First, the high ion core temperatures of the improvedH mode can be explained by T stiffness alone. Theroughly constant pressure at the pedestal top at fixedδ allows higher ion edge temperature at lower edgedensities, and owing to T stiffness the core temper-atures also increase, but the temperature gradient isstill limited by profile stiffness, as is also shown in

Fig. 8(a). The higher temperatures would not leadto an increase in the stored energy if the densityprofiles were stiff as well. The density peaking inthe improved H mode at lower line averaged den-sities, however, leads to an increase in the plasmaenergy of up to 35%. This already takes into accountthe ion dilution by increasing Zeff at lower densi-ties. The H factor compared with the H mode scal-ing ITER98(y,1), shown in Fig. 7, increases even upto 1.6, owing to the density dependence in the scal-ing law τH-98P ∝ n0.41

e . In conclusion, the improvedH mode can be defined in terms of the peaked den-sity profiles. Moreover, the confinement scaling givenin Section 3.1, τE ,th ∝ I1.0

p P−0.56h δ0.2ne(0)0.3n−0.17

e,SOL,describes the good confinement of improved H modeas well. Both Ti and Te profiles in the region 0.4 <

r/a < 0.8 can successfully be modelled with theITG/trapped electron mode (TEM) models, confirm-ing the above conclusions.

A prerequisite for the density peaking may be dueto the zero magnetic shear in the centre. Substitutionof ICRF minority heating for NBI heating leads togradual suppression of fishbone activity and the reap-pearance of sawteeth. As long as sawtooth periodsare very long compared with the energy confinementtime, the central q profile remains flat and densitypeaking together with the plasma energy are kepthigh. Increasing ICRF power finally leads to highersawtooth activity accompanied by the disappearanceof strong density peaking and a high H factor. Stabi-lization of ITG turbulence by Er×B shear flows doesnot seem to be so important, since toroidal plasmarotation, the main driver for the Er field (see alsoSection 4.2), already decreases with the addition of25% ICRH power.


O. Gruber et al.

L mode

0 1 Normalized effective radius

0

5

10

15L mode ITB

improvedH mode

standardH mode

0.5

Ion

tem

pera

ture

[keV

]

standardH modescaled

1.0 1.4 1.8 2.20

5

10

15

H modeL mode

ITB with L modeedge

Cen

tral

ion

tem

pera

ture

[keV

]

Ion temperature r = 0.6 a [keV]

(a) (b)

Figure 8. (a) Ion temperature profiles for different confinement modes. (b) Comparison of

temperature profile peaking for different confinement modes.

4.2. Ion ITBs with strong ion heating

In ASDEX Upgrade we obtain ITBs for ions(strong ion heating with NBI), for electrons (strongelectron heating with ECRF in current drive mode,see Section 5.2) and reactor relevant ion and electronITBs with Te ≈ Ti ≈ 10 keV. Most of the ion ITBsare obtained by significant heating in the currentramp-up, predominantly ion heating by NBI of up to7.5 MW. The H mode transition has to be avoided,as the edge bootstrap current, associated with theH mode, in combination with the large pressure gra-dients of the ITB results in the destabilization ofexternal kinks as soon as the edge q becomes ratio-nal (see Section 7). The q profiles range from mainlynegative shear with qmin > 1.5 to sometimes mono-tonic shape with a broad zero shear central regionwith qmin ≥ 1. Without external current drive, how-ever, the negative central shear cannot be sustainedfor more than about 1 s.

These ion ITBs exhibit strong internal pressuregradients which are characterized by an ITG lengthmuch smaller than that of the H or L mode plasmasgoverned by ITG turbulence and central ion tem-peratures in excess of 10 keV [5, 9]. This is shown inFig. 8, which compares Ti profiles and collects centraland edge Ti values of different confinement regimes.The large scatter in the ITB data suggests non-stiffbehaviour or different confinement states.

In the negative central shear regime, H mode wasuntil recently prevented by limiter contact. This sce-nario has now been extended by upper single null(SNtop) configurations with an increased H modethreshold due to unfavourable ion ∇B drift allow-ing higher power levels [9]. The Ti and q profiles(from MSE measurements in combination with theequilibrium code CLISTE) of such a discharge at 5.0and 7.5 MW of NBI are presented in Fig. 9. Withincreasing heating power the temperatures rise andan expanding radius of the ITB associated with anoutward shift of the qmin radius is obtained (HH-98P

reaches 1.2). The qmin radius is always well alignedwith the foot of the transport barrier (Figs 9 and 10).Shortly before the disruptive termination of thesehigh power SNtop discharges an H mode transitionoccurs. The often disruptive termination of thesehigh power SNtop discharges is caused by couplingof low (m, n) ideal internal modes (driven by largepressure gradients) to an external kink when q95

approaches 4. This tendency is enhanced by H modetransition, since the associated edge bootstrap cur-rent favours external kinks (as mentioned before).Since in ASDEX Upgrade the conducting walls arefar away from the plasma surface, the theoreticalstability limit was calculated without a conductingwall with the GATO code [10], yielding a criticalβN = 1.7 in agreement with the experimental limit.With strong ion ITBs the turbulence in the core is



0

5×1019

1×1020

n (m )e -3

0

100

200

300

400 v (km/s)

#10701

0.84 s

vtor

ne

0 1ρ

tor

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15

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keV

)

#10701

0.84 s

Ti

T (ECE)e

T (Th.sc.)e

0 1ρ

tor

0.5

li=0.880

2

4

6

8

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0 1ρ

tor

0.5

AUG #10701

MonteCarlo

χi,nc

χi

χe

0

5χ

[m /

s]2

0 1ρ

tor

0.5

Figure 9. Temperature, density, toroidal rotation veloc-

ity and q profiles of an ion ITB L mode discharge. Core

heat diffusivities are reduced to the ion neoclassical trans-

port level, which is calculated from Monte Carlo simula-

tions (1 MA, 2.5 T, 5 MW NBI). Standard neoclassical

and potato orbit corrected [30, 31] predictions are given

by dotted curves.

suppressed, as is confirmed by reflectometry signals.A clear reduction of the turbulence is observed whenthe position of the reflection layer falls together withthe foot of the developing transport barrier. Sup-pression of temperature gradient driven turbulenceby the sheared radial electric field Er is plausible,since the Er × B shearing rate greatly exceeds thelinear ITG growth rates γmax ,ITG obtained from theGLF23 model [37] in the centre,

ωE×B = |(RBp/Bt) ∂(Er/RBp)/∂r| > γmax ,ITG

(Bp and Bt are the poloidal and toroidal magneticfields).

Within the error bars ωE×B is close to the growthrate at the foot of the barrier (r/a = 0.6, Fig. 10)[5]. The radial electric field

Er = ∇rp/(ene)− vtBp + vpBt

(vp and vt are the poloidal and toroidal rotationvelocities), is mainly determined by the toroidal

ρtor

0.0 0.2 0.4 0.6 0.8 1.0

ρtor

0.0 0.2 0.4 0.6 0.8 1.0

0

0

5R

ates

[10

Hz]

ωExB

γmax

4

4

3

3

2

2

1

1

AUG #13149L mode ITB

0

20

5

10

15

T [k

eV]

7.5 MW(flat-top)

5 MW(ramp-up)

#13148

Ti

T (ECE)e

7.5 MW(flat-top)

5 MW(ramp-up)

q

Figure 10. Temperature and q profiles during an ion

ITB L mode discharge (1 MA, 2.5 T). ωE×B shearing

rate and linear ITG growth rate γmax are compared at

the 7.5 MW level in flat-top.

velocity, since the contributions of the poloidal rota-tion and the pressure gradient at the ITB itselfare smaller. Here, shearing rates from the measuredpoloidal rotation are in general larger than thosepredicted by neoclassical theory, but the error barson these measurements are rather large, too. There-fore neoclassical theory can be used to estimate thecontribution from poloidal rotation. We have foundthat the Shafranov shift plays an important role inreducing the linear growth rates to levels where theyare below the E ×B shearing rate [5]. The centralShafranov shift is much larger than in normal


O. Gruber et al.

discharges because both the pressure and the safetyfactor are large on-axis with negative shear.

For E × B suppressed turbulence and strongion ITBs, the ion heat transport is expected to bereduced to the neoclassical level. In fact the analysedion heat diffusivities are even below those of stan-dard neoclassical theory (Fig. 9). Standard neoclas-sical theory breaks down close to the axis because theradial size of the orbits is no longer small comparedwith the minor radius of the flux surface. These fatbanana orbits are especially large in reversed sheardischarges, and consequently the region in whichstandard theory breaks down is enlarged. As recenttheories [38, 39] describing the influence of the largeorbits on the neoclassical ion heat flux have yieldedhighly contradictory results, Monte Carlo simula-tions were performed to model the collisions, whichconsist of pitch angle scattering and a correctionfor momentum conservation [5]. Experimental andsimulated heat conductivities show reasonable agree-ment inside and at the steep gradient region aroundr/a = 0.5 (Fig. 9).

4.3. Reactor relevant combined ionand electron ITBs

In a fusion reactor, where electrons are predomi-nantly heated by fast particles, ITBs must be com-patible with Te being equal to or even exceeding Ti.Combining NBI and ECRH/ECCD during the cur-rent ramp-up, ITBs for both electrons and ions withTe ≈ Ti ≈ 10 keV have been achieved [36, 40]. WithNBI only an ion ITB with central Ti ≈ 10 keV isformed (Fig. 11). Especially when central counter-ECCD is applied in a similar discharge, an electronITB also occurs; this is manifested as a strong riseof the electron temperature in Fig. 11, with almostno change in the peaked profiles of ion temperature,density and toroidal rotation velocity. Despite thefivefold increase of the heat flux into the electronsin the plasma centre, neither the electron nor theion thermal conductivity increases [32, 36]. The ITGgrowth rates obtained from the GLF23 model [37]for the two discharges are again below the poloidalshearing rates. Here the destabilizing effect for theITG from the rising ratio of electron to ion tem-perature is largely compensated by the stabilizingShafranov shift, which is larger in the case withECRH because of the higher electron pressure in thecentre. Since the ITB is located inside ρ = 0.4 andcovers a small fraction of the plasma volume only,the maximum HL-89P is still about 1.8.

0.2 0.4 0.6 0.8 1.0 1.2Time (s)

0

5

10

15

T (

keV

) Ti

Te

q

0

2

4

6

8NBI only

#12224

qmin

q0

Ti

Te

ECCD

NBI + ctr-ECCD#12229

qmin

q0

0.2 0.4 0.6 0.8 1.0 1.2Time (s)

0.5 0.7 0.90.5 0.7 0.9

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 (q0, qmin ), central temperatures and Mirnov signals for

n = 1 modes (Ip = 1 MA, Bt = 2.45 T).

5. Fully non-inductively drivendischarge scenarios

As mentioned earlier, stationary tokamak oper-ation requires the maintenance of the plasma cur-rent via a high fraction of internal diffusion drivenbootstrap current supplemented by external non-inductively driven currents. We have started to inves-tigate current drive scenarios with on-axis currentdrive using nearly perpendicular NBI and ECCD ata level of Ip = 400 kA (q95 = 9).

5.1. High βp > 3 discharges withdominant bootstrap current driveat the Greenwald density

To achieve a high bootstrap current, the poloidalβ is maximized by operating at a low plasma cur-rent [41] of 400 kA and a high NBI power of 10 MW.Fully non-inductively driven H mode discharges athigh βp > 3 and densities at the Greenwald den-sity have been achieved. As shown in Fig. 12, thesedischarges start from an ohmic plasma and the NBIpower is ramped up in steps of 2.5 MW to 10 MW.H mode transition already occurs when the first



neutral beam source is switched on. The densityrise initially is facilitated by edge gas fuelling, whileabove 7.5 MW NBI fuelling alone is sufficient tomaintain the density at or above nGW . The currentin the ohmic transformer is kept constant after reach-ing the maximum NBI power of 10 MW. It is obvi-ous here that an ion ITB at ρ ≈ 0.25 is triggeredby the redistribution of fast particles owing to fish-bones (see Section 7.2) [9]. This is confirmed by thereduction of the core ion heat diffusivity as calcu-lated using the transport code ASTRA. Despite thefourfold increase of the heating power, χi is lowestin the high power phase except in the very centreof the plasma, where a small temperature plateau isformed. The minimum value at the position of theITB is ∼0.5 m2/s and is well below the neoclassicalvalues with and without finite orbit width corrections(see Section 4.2 [38, 39]).

The appearance of this ITB coincides with theextinction of the sawtooth activity and — as withthe improved confinement H mode (Section 4.1) —reconnecting (1,1) fishbones remain in the plasma,indicating that a stationary current profile with aq ≈ 1 surface is present. This agrees with the cur-rent or q profiles inferred from MSE measurementsby means of the CLISTE equilibrium code, show-ing a broad minimum of q ≈ 1 in the centre. Itagrees reasonably well also with current diffusionsimulations using ASTRA and assuming neoclassi-cal conductivity (Fig. 12). The large bootstrap cur-rent fraction reaches fBS = 0.56 ± 10% (error esti-mated from 10% error in Te and 1.5 < Zeff < 3) andtends to cause negative central shear in the initiallymonotonic q profile. The neutral beam current drivesthe rest of the plasma current. The plasma currentslightly decreases owing to the flux loss by the some-what decreasing vertical field (β decreases slightly)and the central fishbone activity.

5.2. ECCD discharges and electron ITB(Te ≤ 15 keV) with reversed shear

For the investigation of electron heating and cur-rent drive, ECCD has been applied in L mode plas-mas with low current and density [9]. A low cur-rent permits a large fraction of the ohmic currentto be replaced by the externally driven current,while the low density optimizes the current drive effi-ciency, which scales with Te/ne. Starting from anohmic plasma at ne ≈ 1.5 × 1019 m−3, the ECRwaves of three gyrotrons (1.2 MW) were centrallylaunched in either co- or counter-current direction,

Figure 12. Time evolution of a fully non-inductively

driven H mode discharge with integrated high perfor-

mance with respect to β, ne/nGW and H factors. The OH

transformer current is kept constant from 2.5 s onwards.

Temperature and q profiles at 2.6 s show an ion ITB.

while the remaining 0.4 MW was used for heat-ing. With the onset of ECRH, Te rises sharply inboth cases, but the temperature profiles are dis-tinctly different. While with co-ECCD the temper-ature gradient in the plasma core is nearly con-stant (Te(0) ≈ 10 keV), an electron ITB developsin the counter-ECCD discharge in the presence of areversed magnetic shear inside ρt ≈ 0.25. The elec-tron temperatures go up to 15 keV in the negativeshear region, but a non-thermal part may be includedin the ECE measurements (ions remain cold below1 keV). Again the position of the barrier foot is at theqmin radius. For electron ITBs, reversed shear seemsto be a prerequisite [36, 42], in agreement with the


O. Gruber et al.

theory of ETG driven turbulence and its possiblestabilization [20].

This difference is also reflected in the MHDbehaviour. For co-ECCD, continuous sawtooth activ-ity suggests the existence of a q = 1 surface through-out the discharge, in agreement with ASTRA simu-lations and MSE measurements. Sawteeth disappearin the case of counter-ECCD. Instead, continuouslyirregular (2,1) modes with a high growth rate causeTe collapse. These modes are located near the maxi-mum of the temperature gradient, suggesting a pres-sure driven ideal mode, which might be avoided by asmoother deposition profile and lower electron pres-sure gradients. As two q = 2 surfaces are present inthe plasma [9], the mode may be of double ideal kinkstructure.

Deposition, propagation and absorption of theelectron cyclotron were calculated with the TOR-BEAM code [43], predicting in both cases a drivencurrent of about 320 kA. This is confirmed by currentdiffusion simulations predicting for the co-ECCDcase a current drive fraction of fECCD = 0.82 anda nearly fully non-inductive drive (fBS = 0.12) [44].For the counter-ECCD case the strong MHD modesand huge current density gradients did not permit aquantitative assessment.

6. Stabilization of NTMs by ECCD

NTMs are the main β limiting MHD events inconventional H mode scenarios with positive shear inASDEX Upgrade. Their onset βN scales proportion-ally to normalized gyroradius ρ∗ without significantdependence on collisionality ν∗ over a wide range ofν∗, as already reported [3, 45]. The database support-ing this conclusion has been broadened by includ-ing NTMs occurring during pellet fuelled dischargesat substantially higher ν∗, as well as by compari-son with NTM data from the old ASDEX tokamak[11]. The data points now span more than an orderof magnitude variation in collisionality. This onsetbehaviour is consistent with the stabilizing termsof the polarization current model and, if the heatflux limit for parallel heat flow in the islands istaken into account, may also be described with theχ⊥/χ‖ model in the generalized Rutherford equationdescribing the evolution of the island width. In a den-sity range approaching the Greenwald limit, NTMscan be suppressed by increasing ν∗ [46]. This thresh-old collisionality was found to be well described againby the expression given from the polarization currentmodel.

#9323

2 2.4time (s)

b3/2

b2/1

pert

. mag

n. fi

eld

[a.u

.]

0 0.005 0.01 0.015t / τ

0

0.05

0.1

0.15

w/a

2/1

3/2

R

Figure 13. Temporal sequence of (3,2) and (2,1) NTMs

shown by the perturbed magnetic field. After the (2,1)

mode grows to a sufficient level, the (3,2) mode vanishes,

in agreement with theoretical modelling (right panel).

In ASDEX Upgrade, NTMs may occur in asequence of modes that descends from close to q = 1to a certain outermost surface determined by thelocal onset conditions. Usually, the final mode is a(3,2) or, at lower densities, a (2,1) NTM [11, 12]. Animportant observation has been that NTMs of dif-ferent helicities never coexist in a stationary state.Figure 13 shows the change from a (3,2) to a (2,1)mode, which was modelled with a non-linear cylin-drical tearing mode code [11, 12], also shown inFig. 13.

Stabilization of NTMs by using ECRH or ECCD isof great importance for next step experiments, whichalready suffer at lower βN ’s from NTMs because ofthe onset scaling given above. On ASDEX Upgrade,we inject up to 1.2 MW of ECRH (140 GHz, sec-ond harmonic X mode absorption on the HFS) intothe plasma [47, 48]. By changing the injection angle,we vary the amount of driven current, while feed-forward scans in Bt of 5–10% within 1 s, i.e. slowcompared with the island growth time, are used tofine tune the radial position of the absorption loca-tion. With about 10% of the total heating power,corresponding to 15–20 kA of helical current drivenwithin the island, a (3,2) NTM at βN = 2.5 canbe completely stabilized by co-ECCD (Fig. 14). Theexperimental results are in qualitative agreementwith the generalized Rutherford equation and inquantitative agreement with a two dimensional non-linear cylindrical tearing mode code [49], the mainstabilizing effect being the local helical current gen-erated by ECCD. There is no significant differencebetween AC and DC stabilization. This is due tothe fact that the fast electrons generated by ECCDquickly equilibrate along the field lines so that powerdeposited with finite width near the X point does notgenerate substantial current within the island. PureECRH is much less effective in stabilizing the mode.



#13631

NBI Power (MW) 510

ECRH Power (MW×10)

1.8 2.2 2.6 3 3.4 3.8 4.2 4.6time (s)

Shift of EC Resonance (cm)

48

2.4

n=2 amplitude

βN (suppressed zero)

2

Figure 14. Complete (3,2) NTM stabilization with

ECCD. Switching off ECCD in steps leads to the

reappearance of the (3,2) mode and β reduction.

In the experiments conducted so far, β does notrecover its full value at mode onset, although themode is completely stabilized. This is attributed toa confinement degradation during ECRH/ECCD inthis scenario (see also Section 3.2).

7. MHD in advanced scenarios

A wide variety of MHD instabilities have beenobserved in advanced tokamak experiments onASDEX Upgrade. Most of them limit the accessi-ble operating regime, in particular the maximumachievable normalized plasma pressure, either byterminating the improved confinement or even bycausing disruptions. Besides this unfavourable roleof MHD instabilities, they have been shown to behelpful in both triggering improved confinement andsupporting quasi-stationary discharge conditions.

7.1. Clamping of the q profileby reconnecting MHD modes

The beneficial effect of sawteeth and ELMs forachieving quasi-stationary discharge conditions inconventional H mode is well known. In the absence ofsawteeth, quasi-stationary discharge conditions areprovided by reconnecting (1,1) fishbones in improvedH mode discharges, thus limiting peaking of theimpurity density, electron temperature and also cur-rent profile [34, 35]. During the frequency whistling-down time of the fishbone, the phase of the eigen-function Te/∇Te as derived from ECE [10] does notchange, indicating an ideal mode. The fishbone oscil-lations continue afterwards at nearly constant fre-quency and show a clear island structure with aphase jump located at about ρt = 0.2 for the fishboneconsidered.

In reversed shear discharges with qmin > 1.5, fish-bones with higher mode numbers (m, n > 1) havebeen observed, in agreement with theoretical predic-tions [50]. These fishbones have a similar effect onthe q profile. Often (2,1) fishbones locally clamp thecurrent profile, thus keeping the minimum q value inthe vicinity of 2 for about 100 ms. As for the fish-bones in the improved H mode discharges with flatshear, no degradation in confinement occurs [32].

7.2. MHD instabilities as triggers of ITBs

As described in Section 4.2, the stationary largepressure gradient after the formation of an ITB canbe explained by suppression of turbulence due tosheared poloidal rotation. The dynamics of the ITBformation, however, is not yet well understood. Itseems possible that MHD phenomena influence thisprocess since the existence of low order rational sur-faces has been found to support the ITB formationin various tokamaks [51]. In ASDEX Upgrade, fish-bone activity is found to have a strong influence onthe formation of ITBs [10]. In nearly all dischargeswith a clear ITB, its onset occurs right after the firstfishbone oscillations with m > 2, n = 1, radiallylocated in the vicinity of qmin > 2. At the sametime and radial location the turbulence is suppressed,as measured by reflectometry. In discharges with-out significant MHD phenomena, however, there isusually no ITB formation. Triggering of the ITBis attributed to redistribution of resonant fast par-ticles caused by interaction between fishbones andfast particles, which results in a current across thecorresponding magnetic surfaces. Its return currentin the bulk plasma locally drives poloidal plasmarotation, equivalent to a radial electric field. Theradial current, associated with the fast particle flow,locally drives poloidal rotation, which may lead tothe suppression of gradient length limiting microin-stabilities. The calculated shearing rate ωE×B indeedexceeds the maximum linear growth rate of ITGmodes. Therefore even weaker fishbones should beable to suppress the turbulent transport, which is inagreement with the experimental observations. Oncethe ITB is formed, the stabilization has to be takenover by increasing bulk toroidal rotation and pres-sure gradient, as the fast particle induced poloidalrotation may be damped by neoclassical effects aftera few milliseconds.

Figure 15 demonstrates that fishbones indeed cansuppress turbulent transport. After a period with-out MHD activity in a discharge with an ITB at


O. Gruber et al.

0.4 0.6 0.8 1.0 1.2 1.4

1

2

1.15 1.2 1.25

120

100

60

80

1.18 1.19 1.20 1.21 1.22

f [kH

z]

t[s]

t[s]

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βN

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8

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W]

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I

P NBI

t = 1.20 st = 1.22 st = 1.25 st = 1.30 s

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4

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0 0.2 0.4 0.6 0.8 1.0

q = 5/3

q = 7/3

ρtor

T [

keV

]e

5

10

15

T [keV

]i

#13149

Figure 15. Waveforms of NBI heating power and nor-

malized plasma pressure βN together with a wavelet plot

of the Mirnov signal showing the onset of fishbone oscilla-

tions coinciding with the sudden increase in βN . Forma-

tion of a transport barrier is seen in the electron temper-

ature, which increases after the onset of n = 3 fishbones

in the vicinity of the radial location of the corresponding

rational surfaces of q = m/n (m = 5...7, n = 3).

qmin ≈ 1.5, a sudden increase in the normalized βN

value occurs right after the onset of n = 3, m = 5...7fishbones located outside the ITB at about 1.21 s.The enhancement in βN is caused by a develop-ing second transport barrier seen primarily in theelectron temperature, whose evolution is stopped bythe reduction of beam power. This transport bar-rier sits at about ρt = 0.6, which coincides with theradial location of the corresponding rational surfacesof the fishbone oscillations at ρt(q = 5/3...7/3) =0.53....0.68.

The effect of fishbones on turbulent transport iselucidated further by the effect of (5,2) fishbones onthe electron temperature in another reversed sheardischarge with an established ITB at ρt(qmin ≈ 2) ≈0.4 during the current ramp-up [10]. The fishboneactivity itself periodically modifies the temperatureprofile outside ρt(qmin) in the region close to theouter q = 5/2 surface at ρt ≈ 0.5, which is obvi-ously caused by a periodically local transport reduc-tion. Therefore, besides the transport barrier locatedat qmin that leads to a continuous T rise insideρt(q = 5/2), another one forms at the q = 5/2radius. The radial location of the latter moves out-wards together with the q = 5/2 surface as long asthe (5,2) fishbone activity is present. Obviously, thistransport barrier, located in the positive magneticshear region, cannot be sustained without the fish-bone activity.

7.3. Limitation of operational regimedue to MHD instabilities

We summarize here the MHD instabilitiesobserved with flat or reversed q profiles required forthe ITB formation in ASDEX Upgrade and the influ-ence of localized current drive and heating on MHDmodes.

Besides the fishbones (discussed in Section 7.1),continuous (2,1) double tearing mode (DTM) activ-ity often appears when the minimum q value reaches2 (eigenfunctions derived from Te/∇Te). Like thefishbones, these modes can contribute to the appear-ance of short quasi-stationary phases by clampingthe current profile, at least in the vicinity of qmin .But in contrast to the fishbone activity, these modeslead to breakdown of the ITB, as seen by decreasedtemperatures in Fig. 11(a) (NBI only) between 0.68and 0.78 s. During the DTM activity seen in theMirnov signal, qmin is clamped to 2. The end ofthe (2,1) activity coincides with a drop of qmin

below 2 owing to decoupling of the two rational



surfaces with increasing distance between the reso-nant surfaces. This was proven by stability analyseswith the CASTOR code generalized to include theeffect of differential rotation and based on equilib-rium reconstructions using the MSE diagnostic. Non-linear simulations performed in cylindrical geometryfurther showed that the coupled islands are able toflatten the current profile between the two rationalsurfaces.

Additional electron heating in the plasma centrecan lead to electron ITBs with high central electrontemperatures (Section 4.3) and significantly changesthe MHD stability of the discharges considered. Withcentral ECRF electron heating applied, either in pureheating or counter-CD mode using a movable mir-ror system, the DTM does not appear if the elec-tron heating is provided before the expected onsetof the DTM, or it disappears as soon as the ECRHis switched on (see Fig. 11(b), NBI and ctr-ECCD).A tentative explanation for this effect could be theincreased pressure gradient at the inner q = 2 sur-face, giving rise to a stabilization of the mode due tothe combined effect of bootstrap current reductionin the island and the negative magnetic shear.

The effect of co-ECCD on the MHD stability andconfinement was strongly negative, as in this case thetwo effects — the increased current density in theplasma centre due to the electron heating and thecurrent drive itself — are additive in increasing thecentral current density. Therefore co-ECCD reducesthe shear reversal significantly, which destroys theITBs, affects MHD stability and leads to strong n =1 modes [52].

In discharges with combined NBI and ECRH(Fig. 11(b)), the sawtooth-like events observed onTe after 0.8 s limit peaking of the temperature pro-file and are attributed to ideal (2,1) mode activ-ity (growth time of about 200 µs). This mode wasshown to be an infernal mode driven by the largepressure gradient in the weak magnetic shear region[52]. Coupling of this mode to an external kinkmode is responsible for the disruptive terminationat 1.1 s. Ideal modes often appear in reversed sheardischarges when a low order rational q value at theplasma edge allows the coupling of the (2,1) infer-nal mode to an external kink mode. The resultingmode mostly causes a disruption about 1 ms after itsonset owing to the global character of the eigenfunc-tion as measured by ECE, showing a large amplitudeat the plasma edge. This eigenfunction agrees verywell with those derived from the stability analysis ofm = 2, 3, ..., q95 ideal modes.

7.4. Extension of the operational regimeby non-inductive current driveand wall stabilization

Stability analysis of ASDEX Upgrade reversedshear discharges has shown that confinement degrad-ing MHD activity appears owing to the occurrence oftwo low order rational surfaces of the same helicity(DTMs) or to a large pressure gradient within a weakmagnetic shear region (infernal modes). A q profileoptimized with respect to the stability of core local-ized modes therefore needs distant double rationalsurfaces and pressure gradients in the qmin region assmall as possible. To avoid ideal kinks, high shearat the outer low order rational surfaces (q > 2) isrequired, especially in combination with large pres-sure gradients. Furthermore, in order to avoid theoccurrence of NTMs, qmin should be larger than 1.5at least.

Stationary sustainment and control of such mag-netic shear scenarios against the tendency of theinductively driven current to peak at the high-est electron temperatures require external non-inductively driven currents. Significant modificationof the plasma current profile on ASDEX Upgradehas hitherto been limited to early heating in currentramp-up (reversed shear scenarios) or to ECCD withup to 320 kA in low density discharges. The rota-tion of one of the neutral beam injectors into a moretangential direction being performed now will offerthe possibility of off-axis co-NBI current drive of upto 250 kA peaked at around ρ ≈ 0.5 with a ratherbroad profile (Fig. 16) [32]. The expected currentprofiles were calculated by also taking into account,besides the external current drive, the bootstrapcurrent determined by the pressure gradient andhence depending on the energy transport. Assum-ing either a strong ITB pressure profile or a reducedeffective transport equal to the neoclassical ion dif-fusivity inside the qmin radius results in pressureprofiles like those shown in Fig. 12 (including anH mode edge). The q profile shows a broad min-imum at qmin ≈ 1.5, avoiding close double loworder rational surfaces and having a high shear atq = 2. The plasma current is fully non-inductivelydriven (fBS = 0.65, fNBI = 0.25, fEC = 0.10), butthe high pressure gradient destabilizes core localizedideal modes similar to the reversed shear dischargesdescribed in Section 7.3 and does not allow high βN

operation, in agreement with experimental results(βN = 1.7). Reducing the pressure gradient by a fac-tor of 2 improves the MHD stability of internal ideal


O. Gruber et al.

0 1NBIj

q

j tot

ECjBSj

ρ

5

3MA/m 20

1

0

p/pmax

Figure 16. Pressure, current density composition and

q profiles expected for the more tangential NBI (5 MW,

100 keV) combined with on-axis ECCD. Current compo-

sition: fBS = 0.65, fNBI = 0.25, fEC = 0.10.

modes, and β significantly rises for the current pro-files expected with the new injector geometry (βN ≈3 for a bootstrap current fraction of 35% [10]). How-ever, in this case high non-inductive current drive isneeded.

The ultimate limit to the plasma pressure wouldbe given by the onset of n = 1, 2 and 3 externalkink modes, as li is rather small for reversed shearq profiles (GATO code). External modes, however,could be stabilized by a conducting wall, in combi-nation with a feedback system, which would haveto react on the resistive wall time. At present onASDEX Upgrade, the wall is too far away from theplasma edge to provide a significant stabilizing con-tribution. Installing additional wall structures insidethe vacuum vessel on ASDEX Upgrade is thereforeunder discussion. Since such a wall would have a 3-Dstructure allowing heating and diagnostic access, theresulting influence on MHD stability will be inves-tigated with the 3-D stability code CAS3D. As afirst approximation a continuous wall was assumedat the LFS of the torus. When a wall was placed ata distance of 1.4 times that of the plasma edge fromthe plasma centre, the external kink stability limitwas found at βN ≈ 3.5 with the weaker pressuregradient.

8. Operation with high Zwall coatings

The material for plasma facing components of afuture fusion device is still under discussion. Todaymost experiments use graphite because of its thermalproperties and the low radiation potential, but higherosion yield by physical sputtering, and especiallyco-deposition with tritium, put a question mark onthe use of graphite in a fusion reactor [53]. An alter-native may be high Z materials having a high thresh-old for physical sputtering and, particularly in thecase of tungsten, exhibiting excellent thermomechan-ical properties and low hydrogen inventories. How-ever, central concentrations have to be low because ofthe high radiation power of high Z materials. On thebasis of good experience using tungsten as a diver-tor material in ASDEX Upgrade [54], which demon-strated divertor operation with reactor compatibletungsten concentrations of less than 2 × 10−5, weadopt a step by step strategy for testing tungsten asa first wall material.

To reduce carbon sputtering (the main sourcebeing at the HFS heat shield), a first step was takenby applying siliconization instead of boronization forwall conditioning, which resulted in a typical 100 nmthick silicon coating and in reduced oxygen con-centrations as with boronization. Maximum siliconconcentrations in ASDEX Upgrade are about 0.002(∆Zeff < 0.2) [55]. Consequently the performance ofthe experiment was not influenced by silicon radia-tion. Moreover, these positive effects are much longerlasting than for boronization.

A second step was taken by tungsten coating1.2 m2 of the bottom part of the HFS heat shield(∼ 10% of the total area) with tungsten using phys-ical vapour deposition with a thickness of 0.5 µm.Experiments were restarted without any wall coat-ing to avoid covering the tungsten. Spectroscopicallymeasured central tungsten densities were alwaysbelow the detection limit of approximately 5× 10−6,which was established using laser blow-off experi-ments [2, 56]. These concentrations are below thethreshold tolerable in ITER and are more than a fac-tor of 10 below concentrations which might influencethe plasma operation of ASDEX Upgrade. Using col-lection probes, we found no tungsten in the other wallparts.

The next step will be to cover most of the HFSheat shield with tungsten (about 7 m2 or 2/3 ofthe total area) in the next experimental campaignin 2001 (plasma arcing provides a 1 µm tungsten



sheath). Only at positions which may be used as alimiter will the graphite remain.

9. Summary and outlook

In the H mode both ion and electron temper-ature profiles are self-similar, in agreement withcritical ITG transport models combined with TEMphysics. Stiff temperature profiles are seen at all den-sities and for quite different radial power deposi-tion profiles. This implies that plasma stored energyand global confinement are directly coupled withH mode pedestal pressure, and transport coefficientsare constrained by global effects and cannot be deter-mined by local conditions. Density profiles are notstiff, providing reduced confinement at high densitiesclose to the Greenwald density, but also confinementimprovement with more peaked profiles (improvedH mode). ECRH experiments strongly support thehypothesis that ETG physics, with streamers, playsa strong role in electron transport and results in stiffelectron temperatures at low densities and Te � Ti.

More triangular plasma shapes (δ of up to 0.4)showed remarkably improved H mode confinementas the pedestal pressure increases with triangular-ity owing to enhanced MHD stability. Moreover, thishigh confinement level persists up to nGW with con-finement times close to ITER H98-P scaling. H modedischarges could be sustained up to very high den-sities, reaching 10% above nGW without HFS pel-let injection. These beneficial effects allow the antic-ipated operational regime of ITER-FEAT close tonGW . At the highest triangularities (δ > 0.4) andclose to nGW a regime with small, high frequencytype II ELMs was found which exhibits full H modeconfinement and promises low power load on thedivertor in future devices. Closeness to a double nullconfiguration and q operation above q95 = 4.2 arerequired.

Favourable density peaking despite high averagedensities was obtained by controlled gas puffing inNBI heated H mode discharges on a timescale of 1 s.The needed inward pinch is quite slow and may beprovided by the Ware pinch. The operational win-dow is limited, however, at higher core heating sincethe resulting increasing core energy transport andthe particle outward transport are closely linked. Asecond promising refuelling method is provided byHFS pellet injection, where the existence of localizedhigh β plasmoids and their toroidal drift towards theplasma core were proved. First evidence was gained

that pellet launch velocities of above 500 m/s lead toparticle deposition beyond the ELMing edge regionand hence to reduced pellet particle and plasmaenergy loss rates.

In the advanced scenarios emphasis was placedon physics understanding and performance enhance-ment using extended plasma shape and heating capa-bilities (NBI, ECRF, ICRF). Stationary scenarioswith improved core confinement in combination withan H mode edge barrier and with low central shear(qmin ≈ 1) were extended towards higher plasmadensities close to half of the Greenwald density eitherby edge gas fuelling or by improved core particleconfinement with more triangular plasma shapes.Higher δ’s additionally allowed performance to beincreased up to HL-89P βN ≈ 7, this still beinglimited by NTMs. The energy confinement of theimproved H mode can be largely unified with thatof the standard H mode, as the increase in energy isdue to density peaking at T profiles that are still ITGlimited.

Ion and electron ITBs are obtained in reversedmagnetic shear discharges (qmin > 1.5) mainly withL mode edge. Suppression of fluctuations drivingtransport is observed by reflectometry, in agreementwith the paradigm of stabilization of the temperaturegradient driven turbulence by sheared E ×B veloc-ities (ωE×B > γmax ,TG). Correspondingly, core heattransport is strongly reduced, and the experimen-tal ion heat diffusivities agree well with neoclassicaltransport obtained from Monte Carlo simulations,being reduced compared with those from standardtheory. The ion ITB discharges with strong ion heat-ing using NBI were extended up to HH-98P = 1.2for 1 s (more than 10 confinement times) and Ti ≈15 keV. The addition of ECRF for central electronheating and counter-CD made it possible to achievereactor relevant simultaneous electron and ion corebarriers with Te ≈ Ti ≈ 10 keV. No reduction inperformance of the ion channel was observed.

Fully non-inductive current drive was achieved inan integrated advanced H mode scenario with NBIat high β (βp = 3.1, βN = 2.8), Greenwald den-sity and high confinement (HL-89P ≈ 1.8) but stillat low plasma currents (q95 = 9). An ion ITB isformed, contributing to the bootstrap current frac-tion of 0.56, while the neutral beam current driveaccounts for the rest. As qmin < 1.5, MHD stabilityis partly limited by NTMs.

Central ECRF in mainly current drive mode wasapplied in low current, low density discharges. Cur-rent diffusion and ECCD beam tracing calculations


O. Gruber et al.

predict for stationary co-ECCD discharges a drivencurrent of 320 kA and a nearly fully non-inductivedrive. In the case of counter-ECCD, the central mag-netic shear reverses and triggers an electron ITBwith a central Te of well above 12 keV. Here thestrong localized ECRF deposition in the plasmacentre destabilizes pressure driven ideal modes.

The MHD investigations placed further empha-sis on utilization of ECCD for control of β limitingNTMs in conventional H mode scenarios. The NTMonset βN scales proportional to ρ∗ without signifi-cant ν∗ dependence over a wide range of ρ∗. Thisbehaviour is consistent with the polarization currentmodel and may also be explained with the χ⊥/χ‖model. (3,2) NTM mode suppression was also pos-sible with unmodulated current drive in the islandO point with powers of ∼10% of the total heatingpower (fECCD < 2%). The main stabilizing effectis exerted by the local helical current generated byECCD. Pure ECRH is much less effective in stabiliz-ing the mode. Significant enhancement of β by thismethod will require feedback controlled adjustmentof the absorption location, because the efficiencycrucially depends on exact positioning.

MHD instabilities play a significant role inadvanced tokamak scenarios on ASDEX Upgrade.Besides their well known unfavourable role in lim-iting the operating regime (DTMs, pressure driveninfernal modes, external kinks and mode couplingleading to disruptions), they contribute to makingthe current density profile quasi-stationary (mainlyfishbones), and can even act as a trigger for form-ing ITBs. In particular, fishbones have proved to becapable of inducing suppression of turbulent trans-port by redistributing fast resonant particles, whichcauses a radial electric field and hence stabiliz-ing poloidal shear flows. Other MHD instabilitiesmight have a similar effect, which would explain thepreferential formation of transport barriers in theproximity of low order rational surfaces.

Extension of the operating regime calls for activecontrol of the current profile. Broad off-axis currentdrive with 250 kA will be available in 2001 withmore tangential NBI. Stable core localized modesseem possible by tailoring weakly reversed magneticshear (qmin > 1.5), allowing βN > 3, but the βN

limit will then be imposed by external kink modessince li is rather small for reversed shear q profiles.These modes require conducting structures close tothe plasma — not available at present on ASDEXUpgrade — in combination with a feedback systemto react on the resistive wall time.

Finally, first results from experiments with apartly tungsten coated HFS heat shield demon-strated the suitability of tungsten as a material notonly at the divertor but also in the main chamber of afusion device. The main advantages over graphite arereduced sputtering and low tritium retention. Spec-troscopically measured central tungsten concentra-tions always remained below the detection limit of5 × 10−6. The next step in ASDEX Upgrade willbe a largely tungsten covered HFS heat shield. Thiscomes together with a divertor adapted to optimalperformance at higher δ’s.

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(Manuscript received 2 November 2000Final manuscript accepted 2 April 2001)

E-mail address of O. Gruber: [email protected]

Subject classification: C0, Te; C0, Ti; D0, Te; D0,Ti; E0, Te; E0, Ti; F0, Te; F0, Ti; G0, Te; G0, Ti;H0, Te; H0, Ti; J0, Te; J0, Ti


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