Progress from ASDEX Upgrade experiments in preparing the physics basis of ITER operation and DEMO scenario development

First an MHD phenomenon is addressed in the plasma center, where elevated central safety factor values of q_s(0) > 1 together with sawtooth-free phases were observed, when transport modelling would predict q_s(0) < 1 and sawteeth. The experiments were carried out motivated by theoretical results from nonlinear MHD simulations with the MD3D-C code, where it was found that a dynamo driven by a (1, 1) quasi-interchange mode transports poloidal magnetic flux radially outward raising q_s(0) above one [13]. This occurs in a stationary process with continuous redistribution of magnetic flux. In the simulations, this mechanism is activated at high values of the plasma β, when the dynamo drive is sufficiently strong in order to counteract the plasma current diffusion which would lead to a centrally peaked current profile.

Recently, the β threshold for the flux pumping effect was investigated by means of on-axis electron-cyclotron current drive (ECCD) experiments (see figure 2). In a discharge with modest co-ECCD of 100 kA, the sawteeth disappeared when the β value, increased by NBI power steps, surpassed a critical value [14]. A successive increase of the driven current made the sawteeth reappear. Figure 2 also shows that the amount of tolerable co-ECCD without producing sawteeth increases with β [14]. The discharges were analyzed with a combination of the imaging motional Stark effect (IMSE) diagnostic [15] and the IDE equilibrium solver [16, 17]. When sawteeth are suppressed, the IMSE data indicate a flat central q_s profile clamped to values near unity whereas based on neoclassical current diffusion, the equilibrium solver predicts a monotonic q_s profile starting from q_s(0) < 1 [14]. The effective charge profile is flat in the plasma centre with values of Z_eff ⩽ 1.5. Therefore Z_eff can only play a minor role for the current profile.

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The observed dynamo effect can enable strong ECCD in the center of reactor plasmas, where the current drive efficiency is high, with flat central q_s profiles as required for enhanced core plasma performance.

One of the unresolved issues of core heat transport is the magnitude of the isotope effect and its cause. A novel strategy was used to disentangle the isotope effect in H-mode core and pedestal transport by matching the pedestal profiles with a slight increase of the plasma cross section's triangularity δ for the deuterium case with respect to that in the hydrogen plasmas while keeping heat and particle sources the same [18, 19]. Figure 3 shows the kinetic profiles of a pair of hydrogen and deuterium discharges which both were carried out with a high NBI heating power of about 10 MW and a particle fuelling rate of 7–10 × 10²¹ s⁻¹.

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A comparison with modelling results highlights the role of fast ion (FI) populations in reducing turbulent transport in NBI heated discharges. In agreement with results from the theory-based turbulence model TGLF, the core isotope effect was found to be small as long as the fraction of FIs remained below about 30 % of the main ion density [18]. At a higher NBI power with larger fast particle populations, a stronger decrease in ion heat diffusivity was indicated in deuterium with respect to that in hydrogen plasmas leading to substantially higher core ion temperatures, Due to the longer slowing down time of injected deuterium ions compared to that of hydrogen ions, the FI energy content in deuterium plasmas was about 50 % higher.

Nonlinear simulations of these discharges with the gyro-kinetic turbulence code GENE [20, 21] revealed the importance of turbulence stabilization by electromagnetic and FI effects. For both isotopes, the FI distribution from NBI stabilises turbulence and reduces turbulent transport. The stronger effect observed in deuterium plasmas is attributed to the higher content of non-thermal ions [18].

The energy confinement in helium plasmas with dominant electron heating is similar to that in deuterium, while a degradation with an increasing fraction of ion heating is observed. These observations can be theoretically explained by a different role of zonal flows in electron and ion dominated turbulence with different main ions [22].

Low-density L-mode plasmas with dominant electron heating were studied on AUG [23] in particular for their relevance to the first pre-fusion power operation phase of ITER. For this purpose, ECR heated hydrogen and deuterium discharges were performed and modeled with TGLF-SAT1geo, a quasi linear turbulent transport model based on TGLF that includes a turbulence saturation rule and an improvement in the description of geometrical effects [24, 25]. The model reproduced both the central electron temperature and the edge ion heat flux, which is critical for the L–H transition [26, 27], with high confidence. This is seen as a validation of TGLF-SAT1geo also in applications for the prediction of the absorption of X3-mode ECRH in ITER, which critically depends on the central electron temperature.

The same model was also used for the interpretation of cold pulse experiments using impurity laser ablation. The dynamic response of the electron temperature profile following the cooling event in the plasma edge was successfully modeled with the ASTRA transport code with the local transport model TGLF-SAT1. The reason for the—sometimes called nonlocal—fast response of the core temperature was explained by the stabilization of trapped electron modes through an induced and also correctly predicted flattening of the electron density profile [28].

The TGLF model also proved reliable in describing the role of beam ions in reducing ion heat transport in H-mode plasmas. It predicted ion temperature gradient (ITG) turbulence properties observed in experiments, where the ion stiffness was reduced simultaneously with a drop of the electron to ion temperature ratio, T_e/T_i. In addition, a potential role of electron temperature gradient (ETG) driven modes in strongly electron heated H-mode plasmas in limiting the increase of electron temperature gradient was identified by means of linear and nonlinear gyrokinetic simulations with the GKW code [29].

The core density profiles of ITER or reactor plasmas will be dominated by transport processes. AUG experiments reproducing reactor conditions of heat, particle and momentum sources and related integrated modelling using ASTRA and TGLF demonstrated the role and dominance of a collisionality dependent turbulent pinch in producing centrally peaked plasma density profiles at reactor relevant low collisionalities [30].

Fast-ion effects turn out to also be important for the description of low-Z impurity transport, which has important consequences for the operation of reactor plasmas. Especially helium transport will determine the amount of fuel dilution in the core of a burning plasma. In order to enhance the physical understanding of low-Z impurity transport, transport studies were carried out for helium and boron. In particular, a new modulation technique was developed and applied to boron transport studies to disentangle diffusive and convective transport coefficients [31, 32]. Using this technique, a database of transport coefficients covering a wide range of plasma parameters was assembled and compared to theoretical predictions.

For steep ITGs (${R}_{0}/{L}_{{T}_{\text{i}}} > 6$), which coincide with strong NBI heating, outward convection and hollow boron density profiles appear. In contrast, even low levels of electron heating increase both the diffusion and inward convection and result in peaking of the impurity density profiles. Comparisons with a combination of neoclassical and quasi-linear gyrokinetic turbulence simulations (NEO and GKW, respectively) showed good agreement for plasmas with combined NBI and ECR heating.

The hollow boron density profiles, on the other hand, are not fully reproduced by the simulations, particularly around mid-radius. This is demonstrated in the comparison of measured and modeled radial profiles of the boron density scale length in figure 4, where the inclusion of fast ions leads to a clear increase in the outward predicted transport at a normalised radius of ρ_tor = 0.4. In the modeling, it is a combination of different fast ion effects that leads to the development of hollow boron profiles [33, 35]. The fast ion distribution stabilises ITG driven turbulence and leads to an increase of neoclassical outward convection [35].

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However, at mid-radius this effect is insufficient to reach experimental levels due, at least in part, to the relatively small fast ion population at this location. Figure 4 also shows that the results of the simulations are very sensitive to the gradient scale lengths of the background profiles. Small increases in the ion temperature gradeint and rotation gradient can change the sign of the predicted convection.

A validation of the physics models of turbulence codes is also carried out directly on the microscopic level of the fluctuations [36]. For this purpose, measurements of a large set of fluctuation data is assembled on L-mode discharges at two values of the electron temperature gradient length. This implies density fluctuation wavenumber spectra [37], temperature fluctuation frequency spectra [38], the correlation lengths of both, cross phases between density and temperature fluctuations [38], turbulent flow and phase velocities including zonal flow structures. The experiments are completed and the analysis is ongoing. Comparison of these data with local and global GENE simulations will provide a comprehensive test of the physics models used in the code.

As a first example, the poloidal symmetry of the turbulent propagation velocity was tested using a Doppler reflectometer which probed a radial and poloidal region around the outer midplane. It turned out that the interpretation sensitively depends on the background density profile. From careful analysis, it was concluded, that the flows on the outboard side are poloidally symmetric, where on each flux surface fluctuations propagate with the same velocity at all wavenumbers. A comparison with spectroscopic measurements showed that this velocity is close to the E × B drift velocity [39].

Due to the absence of reliable physical models for the L-mode plasma edge, predictions for the power threshold for L–H confinement transitions, P_LH, rely on experimental scaling laws. In particular, the dependence of P_LH on the isotopic mass must be known to predict the performance of ITER in the pre-nuclear phase in hydrogen or helium and for the D–T phase. An earlier study on JET indicated a beneficial 40 % reduction of P_LH when small concentrations of helium were added to NBI heated hydrogen plasmas. Furthermore P_LH increased to the hydrogen level when even a small amount of hydrogen was admixed to deuterium plasmas [40].

Figure 5 shows comparable experiments performed on AUG. The left figure shows that the threshold power did not change and remained at the hydrogen level, when up to 20 % of helium was added to the hydrogen plasmas. The data was obtained from both ECRH and hydrogen NBI heated discharges. For the right figure, a continuous transition from a pure deuterium plasma to a pure hydrogen plasma was performed. The data from the different heating schemes consistently shows a continuous increase of P_LH with hydrogen concentration from the D level to the H level. A stronger increase starts only above a relative high hydrogen content of about 60 %. These differences from the JET results need further investigations.

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Power balance analyses show that the ion heat flux through the separatrix, which is known to be a key quantity for the L–H transition on AUG [26], is independent of the helium concentration and heating scheme [42]. In hydrogen and deuterium plasmas, the L–H transitions happen at similar values for the neoclassical E × B shearing rate [27]. Spectroscopic measurements of the edge radial electric field at the L–H transition of hydrogen and deuterium plasmas were also consistent with the observation that the neoclassical E × B shear, given by the ion pressure gradient, is the key physical parameter for turbulence suppression. The transition in hydrogen and deuterium plasmas is found to occur at the same E × B velocity [43].

On AUG, the edge density was identified to be the leading parameter for ELM mitigation with magnetic perturbation fields [45]. At low densities the smallest relative pedestal energy losses are observed when a magnetic perturbation field with a toroidal mode number of n = 2 is applied, inducing a so-called density pump-out.

In recent experiments, the physical origin of the density pump-out was investigated [46]. By toroidal rotating of the n = 2 perturbation field, a reflectometer observed toroidally asymmetric broad-band fluctuations with frequencies up to 150 kHz at the outer midplane. The fluctuation spectrum is locally quenched at a certain phase angle of the field perturbation. The fluctuations appear simultaneously with a drop in edge density associated with a density pump-out and are held responsible for it. To identify the transport losses caused by the fluctuations, their dynamics was compared to that observed with a poloidal array of Langmuir probes on the outer divertor target. It was found that the toroidal asymmetry of the turbulence at the outboard midplane maps along the magnetic field lines to the divertor. High and low fluctuation amplitudes at the plasma edge are magnetically connected to high and low particle transport measured in the divertor [46].

These observations lead to the interpretation that turbulence near the separatrix causes the density pump-out which reduces the edge pressure gradient and in turn leads to peeling-ballooning stability and ELM suppression. The formation of magnetic islands due to the perturbation field is not indicated on AUG [46].

The improved confinement mode (I-mode), which was observed on several tokamaks [47–50], has a number of attractive properties with respect to a reactor plasma. It can be achieved with pure electron heating and shows neither impurity accumulation nor large ELMs. The I-mode occurs in an unfavorable magnetic configuration, i.e. the ∇B drift points away from the active divertor, resulting in a higher L–H power threshold. On AUG, the I-mode is routinely achieved as long as the heating power remains below the H-mode threshold value [51, 52]. The operating space was also expanded to NBI heating, where stationary I-mode phases were achieved using active β feedback control.

Figure 6 depicts such a discharge, where the required β value was stepwise ramped up by feedback controlled NBI power raising to a total heating power of about 4 MW [53]. After an increase in the edge electron temperature, indicating an electron heat transport barrier, a stationary I-mode phase develops with a weakly coherent mode (WCM) in the range 50–100 kHz. In the particle transport channel, the transport barrier is not visible and the density remains constant. Intermittent transport events associated with the WCM become more dominant with increased heating power until the I-mode phase ends with a transition to H-mode when values of β_pol ≈ 0.6 and H₉₈ ≈ 0.9 were reached. In the phase before the transition, larger edge relaxation events (PREs) are observed. They are smaller than type-I ELMs but can be expected to be detrimental for a reactor divertor and therefore must to be avoided [53].

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The operational window for I-mode discharges was investigated on the basis of a database with density and electron temperature pedestal values from plasmas in different confinement regimes. The data from L-mode, I-mode and H-mode plasmas are clearly separated by electron pressure isobars. I-mode pedestal relaxation events occur only where the edge plasma pressure is close to the isobar above which the operational space of H-mode discharges begins [53]. This study highlighted the importance of the edge electron pressure for a stable operation of I-mode discharges and also the challenge of extending the regime to higher heating powers.

Furthermore, nitrogen seeding of I-mode discharges was tested as a way to combine the I-mode regime with a detached divertor [54]. In these experiments, the seeding rate was ramped to a maximum nitrogen core concentration of about 1.5 %. The β window in which the I-mode can exist becomes wider and shifts to higher values. This is not due to increased radiation, but rather due to increased edge transport as nitrogen is introduced into the plasma. While complete detachment was not obtained, the inner divertor was detached along with a reduction of the heat flux on the outer divertor at a confinement factor of H₉₈ = 0.9 [54].

A physical picture of the I-mode was derived from simulations using the gyro-fluid code GEMR [55]. Based on the experimentally justified assumption that at the separatrix the ion temperature is higher than the electron temperature and the ion temperature gradient is shallower than that of the electrons, the strong drive of the ITG turbulence is removed. Consequently, the L-mode turbulence turns out to be drift-wave like before the transition. It is shown, that the L–I transition is caused by a suppression of small scale turbulent fluctuations by finite Larmor radius effects and the large scales by phase randomization. The WCM appears as a remnant of the drift wave turbulence spectrum. The simulations reproduce the intermittent transport related to the WCM and also clarify the origin of the I-mode operation window in heating power, magnetic field strength, and collisionality. The fact that a transport barrier forms only in the electron heat channel is attributed to the dissipation of the electron temperature fluctuations by parallel heat conductivity at low collisionalities [55].

Recently, stationary ELM-free phases of H-mode discharges were observed on AUG [56]. Closer inspection revealed the similarity of key properties to those of the enhanced D_α H-mode (EDA H-mode) from Alcator C-Mod [57–59]. In contrast to the I-mode discussed above, the ion ∇B drift is in the favourable direction and the density pedestal is H-mode like. It has high energy confinement (H₉₈ = 0.9–1.3) and high density (f_GW = 0.8–0.9) without impurity accumulation. This makes the EDA H-mode a candidate for a fusion reactor regime. On AUG this ELM-free regime is accessed when the ECRH power is above the L–H threshold P_LH and it can be sustained with 50 % of the heating power coming from NBI. Key elements to reach this regime are strong gas puffing and a relatively high safety factor while plasma elongation is favourable to make it more robust [56].

Also on AUG, the EDA regime exhibits the characteristic quasi coherent mode (QCM) with frequencies decreasing from the range 40–80 kHz after the transition to 15–50 kHz during the stationary phase. The mode frequency spectrum broadens under strong gas puffing. The QCM is localized close to the separatrix or in the region of steep pressure gradient and causes transport to the divertor. The toroidal mode number is about n = 20 and the mode propagates in the electron diamagnetic direction. This leads to a pedestal pressure profile close to the ballooning but far from the peeling limit [56].

While the regime appears to be compatible with nitrogen seeding for divertor detachment, main challenges in terms of reactor relevance are the limited power window and the high safety factor under which it exists. Without impurity seeding, the EDA regime develops ELMs when the total heating power exceeds about 3 MW. Shaping, gas fuelling and higher current increase the acceptable power before ELMs occur. However, in recent experiments with argon seeding it was possible to increase the heating power up to about 8 MW [60]. In such discharges, the argon seeding rate was the feedback actuator with the power crossing the separatrix as control variable. The latter was derived from the difference of the total heating power and the real time bolometric measurements of the radiated power from inside the separatrix [61]. Since argon radiates mainly from the plasma edge region, the heat flux through the separatrix could be kept below the threshold value for the transition to an ELMy H-mode [60]. A more robust extension of the power window and the compatibility with lower safety factors remain the challenges to be overcome for this regime.

Some time ago it was observed at AUG that in strongly shaped H-mode plasmas strong gas fuelling can suppress type-I ELMs [62–64]. The large ELMs that cause high transient divertor power loads are replaced by more frequent and smaller edge relaxation events, also known as small, type-II or grassy ELMs, leading to quasi continuous power exhaust (QCE). Due to its relevance to a reactor plasma scenario, the now called QCE regime has recently been further studied and developed on AUG [65, 66], TCV [67] and EAST [68]. The small ELM regime reported from JET [69] is distinguished from the QCE regime by its low gas puff and edge density.

Figure 7 illustrates the two key ingredients for obtaining the QCE regime, which are high density (left) and the closeness of the magnetic configuration to double null (right) measured by the distance between primary and secondary separatrix at the outer midplane dR_XP. As the density rises in a discharge with a constant magnetic configuration (left figure), the type-I ELMs first become smaller and then vanish when a critical density is reached. In the right figure, type-I ELMs reappear at about 4.3 s, when during a QCE phase the elongation is reduced below a critical value of dR_XP. Both elements contribute in an additive manner to the formation of a QCE plasma edge.

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As in the other ELM-free scenarios, also in the QCE regime it is indicated that transport caused by fluctuations close to the separatrix relax locally the pressure gradient which stabilizes peeling-ballooning modes and thus the large ELMs. In addition to the small ELMs, fluctuations in a frequency range 20–50 kHz are also visible in the magnetics on AUG and JET [64, 70]. These remaining small edge fluctuations, which can also appear in-between type-I ELMs, could be attributed to high-n ballooning modes [65, 67]. These modes become more unstable due to the increasing connection length between the stable high-field and the unstable low-field side in close to double null configurations. Also a study of type-I ELMs on AUG with the non-linear MHD code JOREK [71] proposes a mechanism for the QCE regime related to resistive peeling ballooning modes which appear at high density near the separatrix and prevent pedestal build-up and type-I ELM crashes [72]. In the QCE regime, enhanced filamentary transport at the plasma edge broadens the SOL power fall-off length by up to a factor of 5 as compared to inter type-I ELM values and a density shoulder [73, 74] forms [75].

The QCE regime, with separatrix plasma parameters similar to those expected for a reactor plasma, was further developed to integrate high energy confinement and high density with a detached divertor. Furthermore, in order to demonstrate the reactor relevance of the QCE regime, a discharge was designed that did not produce a single type-I ELM [76]. To avoid large ELMs in the early phase of the discharge, it was started with a low plasma current and high safety factor q₉₅, where the ballooning modes near the separatrix are more unstable. After the transition to H-mode, the dynamic increase of the triangularity and moving towards a double-null configuration was accompanied by strong gas fuelling. After reaching a flattop, a double feed-back control was activated: the pre-programmed constant plasma β value was controlled by the NBI heating power and the divertor temperature was reduced by controlled nitrogen seeding in two steps first to 10 eV and later to 5 eV. With a combined ECR, ICR and NBI heating power of 15 MW, this type-I ELM-free demonstration discharge achieved a normalized β_N = 2.1 and a confinement factor of H₉₈ = 0.9 with a partially detached divertor. The detachment is accompanied by a modest loss of confinement [76].

This promising regime for ITER and DEMO plasmas will be further developed to achieve lower values of the edge safety factor q₉₅ and a robust detachment scenario, where the control via the XPR (see section 6), which also appears during QCE phases, can make an important contribution.

On AUG, the quiescent H-mode (QH-mode), an ELM-free operational regime previously discovered on DIII-D [77], was observed for the first time on a metal wall device [78]. The QH-regime relies on strong plasma rotation and low edge density. In a recent campaign on AUG it was obtained with 8 MW counter-NBI heating. The characteristic modes known from experiments in the carbon-wall AUG, the edge harmonic oscillation (EHO) at 5–10 kHz [79, 80] and the high frequency oscillations at 350–490 kHz [79], were also clearly observed on the tungsten-wall AUG. However, the particle transport driven by the EHO was insufficient to control the impurity content of the discharges. Consequently, the QH-mode phases remained transient and ended in a radiation collapse. Central ECRH to avoid impurity accumulation resulted in high T_e/T_i ratios, one of the fundamental differences to previously observed QH-modes on the carbon-wall AUG [79]. The stationarity of this regime in a tungsten device and the integration of divertor detachment remain open questions.

First experiments were conducted to investigate the effect of negative triangular plasma shapes on the discharge performance of AUG [81]. Shapes with upper and lower triangularity of δ_up = −0.45 and δ_lo = 0.05, respectively, with an average triangularity of δ = −0.2 were achieved on AUG. In discharges where the ECRH power was stepped up to about 5 MW at 80 % of the Greenwald density, the plasma edge remained in L-mode without ELMs. At the same time the normalized energy confinement was close to H₉₈ = 1 and also the power degradation followed that known for H-mode confinement [81].

Table 1 summarizes the main characteristics of the small or no-ELM regimes discussed in this section. Although the regimes are obtained through different approaches, transport caused by turbulence or coherent modes at the outer plasma edge was identified as a common mechanism to realize peeling-ballooning stability and type-I ELM suppression.

The goal of an integrated reactor-compatible plasma scenario at high heating power and with a detached divertor is not yet reached in all cases. In particular, it remains a challenging task to integrate high separatrix density and detached divertors into the QH-mode, negative δ plasmas, and the regime with RMP ELM suppression. On the other hand, it is difficult to extend the operation ranges of I-mode and EDA-H-mode plasmas towards high heating powers. At least for EDA H-mode plasmas, dissipating higher power by argon radiation is a promising approach. Currently, the QCE regime, which can combine high confinement with a partially detached divertor, can be operated closest to a reactor relevant integrated scenario.