OV ITER-relevant research on the COMPASS

R. PANEK1, M. HRON1, M. KOMM1, J. ADAMEK1, P. BILKOVA1, K. BOGAR1,2, O. BOGAR1, P. BOHM1, A. CASOLARI1, J. CAVALIER1, R. DEJARNAC1, M. DIMITROVA1,3, M. FARNIK1,4, O. FICKER1,4, O. GROVER1,2, P. HACEK1,2, J. HAVLICEK1, A. HAVRANEK1,4, J. HORACEK1, M. IMRISEK1,2, F. JAULMES1, M. JERAB1, K. JIRAKOVA1,4, K. KOVARIK1, J. KRBEC1,4, L. KRIPNER1,2, E. MACUSOVA1, E. MATVEEVA1, T. MARKOVIC1,2, J. MLYNAR1, D. NAYDENKOVA1, P. PAVLO1, M. PETERKA1,2, A. PODOLNIK1,2, J. SEIDL1, V. SKVARA1,4, M. SOS1,4, J. STOCKEL1, M. TOMES1,2, J. URBAN1, M. VARAVIN1, J. VARJU1, P. VONDRACEK1,2, V. WEINZETTL1, J. ZAJAC1 AND F. ZACEK1 AND THE EUROFUSION MST1 TEAM4

1 Institute of Physics of the CAS, Za Slovankou 3, 182 00 Prague 8, 2Charles Univ, Fac Math & Phys, V Holešovičkách 2, 180 00 Prague 8, Czech Republic 3Emil Djakov Institute of Electronics, Bulgarian Academy of Sciences,72, 1784 Sofia, Bulgaria 4 FNSPE, Czech Technical University in Prague, Břehová 7, Czech Republic 5See the author list of Meyer H. et al 2017 Nucl. Fusion 57 102014

Corresponding author: M. Komm, [email protected]

1. Introduction

The COMPASS tokamak (R=0.56 m, a=0.18 m, IP < 400 kA, BT < 1.6 T) [1] is one of the currently operated with ITER-like plasma cross-section. Thanks to its modest side, it is flexible in operation, capable of quickly adapting its components to studies of different topics, mostly focused on edge plasma and physics.

In the years 2017-18 the research on the COMPASS tokamak was focused on support of solution of the key challenges for the design and operation of ITER and next-step devices. This included mainly installations and upgrades of state-of-art edge plasma diagnostics, development of relevant scenarios (Type-I ELMy H-mode, detached plasmas, runaway electron generation etc.), contribution to multi-machine scalings and, finally, exploitation of unique features of COMPASS like high-field-side magnetic perturbation coils. This contribution provides an overview of such activities, summarizing the recent result.

2. Measurements of ELM properties

The COMPASS tokamak divertor has been recently equipped with an array of 155 Langmuir and Ball-pen probes, which is capable of measuring impacting heat flux with sub-microsecond temporal resolution [2], which is superior to the capabilities of the commonly used IR thermography. Thanks to the relatively large biasing voltage for the ion saturation current measurement (-270 V), this unique diagnostic allows to resolve heat flux evolution during ELMs or even during individual ELM filaments. The energy delivered by ELMs is a crucial parameter for ITER and next-step devices, as frequent ELMs can lead to overloading of the divertor plasma-facing components (PFCs) and drastic reduction of their lifetime. In order to predict the peak ELM energy density for ITER, a multimachine scaling was established and a model proposed by Eich [3].

Figure 1: ELM peak energy density measured by the divertor probes vs. model proposed by Eich. Figure reprinted from [2].

COMPASS data obtained by the divertor probes [2] (shown in Fig. 1) were able to complement the existing data obtained by IR thermography from other tokamaks and compare with model prediction [3].

The combination of ball-pen and Langmuir probes mounted on the horizontal reciprocating manipulator allows us to perform also upstream measurements of the power decay length during ELM and inter-ELM periods as seen in Fig 2. We have studied the peak heat flux caused by ELM filaments. This was facilitated by the relatively modest heat fluxes in the SOL of COMPASS, which allowed us to measure in the vicinity of the LCFS with the reciprocating probe even in H-mode plasmas.

Figure 2: Inter- and Intra-ELM power decay length measured by the reciprocating probe. Figure reprinted from [4].

It was observed, that the upstream power decay length during ELMs exhibits a significant broadening (factor of 4) compared to the inter-ELM value [4]. This observation is consistent with intra-ELM broadening of the power decay length seen with IR measurements at JET [5], which were however performed at the divertor targets.

3. Detachment studies

Detachment is the regime envisaged for ITER burning plasma scenarios. In this regime, most of the exhaust power is radiated, the power and particle flux to the target is strongly reduced and thus the localized heat flux deposition at the divertor targets is suppressed and material limits can be met.. COMPASS tokamak has short connection length and open divertor geometry, two factors which are deemed unfavorable for detachment access. However, several dedicated campaigns have focused on achievement of partial detachment in L-mode conditions. Series of experiments with nitrogen seeding in the range of 2-9x1020 molecules per second at different locations in the divertor region were performed in otherwise identical discharges (Ip=- 19 −3 210 kA, BT =-1.38 T, ne=5×10 m ). The injection of nitrogen has repeatedly caused sharp decrease of electron temperature at the outer target to values below 5 eV [6]. When the nitrogen was injected at the HFS target, it provoked so-called “cliff-edge” transition to detachment (Fig. 3A-C), while more gradual evolution was observed during injection at the LFS (Fig. D-F).

Figure 3: Temporal evolution of Te, pe and q|| in the divertor for seeding at HFS (a-c) and LFS (d-f) measured by probes in the vicinity of the outer strike point.

The comparison of upstream and divertor pressures have confirmed, that while nitrogen penetrates into the confined region and can cause cooling of the plasma inside the separatrix, the pressure drop at the outer divertor target is more substantial, confirming that significant amount of power is being dissipated in the SOL and divertor. In order to compensate for the core radiation, additional power was delivered to the plasma by means of NBI heating. This has helped to recover the upstream pressure without having a significant effect on the downstream values as demonstrated in Fig. 4.

Figure 4: Evolution of upstream (magenta) and downstream (blue) pressure in ohmic (A) and NBI-heated (B) discharge with identical amount of seeding.

The NBI-assisted seeded discharges allowed to achieve a pressure ratio pupstream/pdownstream > 10, which is considered as a proof of partially detached plasma.

4. Study of misalignment of the central solenoid Error Fields (EF) that originate from displacement or misalignment of tokamak coils during assembly or from nearby ferromagnetic structures can negatively affect the plasma discharge via destabilization of tearing modes which can consequently lead to confinement degradation or disruption. For a successful ITER operation, EF sources will have to be mitigated and above- the-threshold remaining EF corrected by Error Field Correction (EFC) coils. While the correction criteria and effects of EF that originate from Low-Field-Side (LFS) of the torus are well understood, intrinsic EF on NSTX-U that originates from High-Field-side (HFS) (presumably due to TF coil misalignments) [7] exhibits a different and less predictable behavior. The ITER organization has asked COMPASS to utilize the unique HFS-localized 3D coils of this tokamak to generate and investigate the effect of a controlled HFS EF, having in mind possible reevaluation of the constraints on installation precision of the CS and TF on ITER and possible necessity to use its top and bottom EFC coils to correct such EF.

Figure 5: Configurations of COMPASS EF and EFC coil discussed in this study. Left - HFS EF coil configuration simulating the tilt of central solenoid (see illustration). Middle - HFS EF with LFS EFC coils. Right - HFS EF coils with EFC coils on LFS, top and bottom, used in L-H transition studies. Simulations by ideal MHD code IPEC [8] that is going to be used on ITER show that LFS n=1 fields typically couple to all the dominant rational surfaces of plasma equilibrium with the same toroidal phase [9], which simplifies the identification of critical EF for mode locking as well as their correction since all the resonances can be corrected at the same time by a single row of midplane-localized EFC coils. However, the coupling of HFS EF to different rational surfaces is not toroidally aligned, especially at low beta [9], making a bifurcation in mode locking prediction criteria. Therefore, COMPASS has addressed this issue. Artificial EF that simulated the tilt of CS (shown in Fig. 5) was applied to low-beta plasmas of ITER baseline-relevant scenario (q95=3, aspect ratio=3.23), while plasma density ne was varied and other parameters kept as constant as possible. In Fig. 6 that shows the critical current in EF coils necessary to induce mode locking at different densities, a distinct parameter boundary can be seen. By its comparison to the predictions of critical EF from multi-machine empirical criterion that is based on IPEC simulations and global plasma parameters [10], the dominant role of m/n = 2/1 surface in mode locking induction is identified, as opposed to the necessity of EF having also to couple to 3/1 and 4/1 surfaces as well. Note that this bifurcation is an effect of HFS EF and that for LFS-originating EF the core and full coupling would degenerate into single a criterion.

Figure 6: ne vs. Icoil dependence to induce mode locking by HFS EF across multiple COMPASS discharges. IPEC modelling curves show critical Icoil according to the multi-machine empirical criterion if the coupling only to the 2/1 surface or all the rational surfaces is considered. Furthermore, since the data have shown that the effects by HFS EF can be predicted in the same manner as those by LFS EF, the COMPASS results were used to update the empirical multi-machine mode locking criterion, as shown in Fig. 7 [11].

Figure 7: Mode locking threshold EF scaling based on plasma parameters across multiple devices [11]. However, when HFS EF below the mode locking threshold was applied prior into L-H transiting plasma, disruptions occurred without any exception. Therefore, the LFS midplane EFC coil row (shown in Fig. 5 middle) was scanned with different Icoil magnitudes and toroidal phasing with respect the EF coil arrays in attempt to correct the EF and restore the H-mode access. Using the toroidal phasing as shown in Fig. 5 middle and currents of |Icoil-EFC| = 1.25 kA to |Icoil-EF| = 3.25 kA, the disruption rate over L-H transition due to strong HFS EF dropped down to ~54% (other toroidal phases of EFC did not show any improvement). The IPEC simulations of correcting the coupling to the 2/1 surface have identified that the optimal magnitude and toroidal phase of LFS EFC is very close to the empirically-found optimum, as it is shown in Fig. 8, further validating the reliability of ideal MHD modelling IPEC to predict behavior of plasma in HFS EF presence.

Figure 8: Coupling of HFS EF and LFS EFC to the 2/1 resonant surface in plasma according to the ideal MHD code IPEC. HFS EF are fixed in magnitude and phase, while LFS EFC is scanned in magnitude and toroidal phase. Plasma equilibrium is of low- beta, pre-L-H transition discharge [10]. Recently, by further supporting the LFS EFC by the top and bottom EFC (see Fig. 5 right), it was possible to drop the disruption rate over L-H transition all the way down to 20%. This promising result will be interpreted using the IPEC modelling in the near future. The experiment and IPEC modelling of correction of the HFS EF with LFS EFC in fully- developed H-modes are also in agreement, although the observations show that the correction criteria are less strict than those during L-H transition. IPEC simulations show shift of the optimal EFC toroidal phase by 45 degrees, but as long as the resulting coupling is within a well of certain minimal values, even wrong toroidal phasing over a rather broad (~0.5 kA) magnitude range is acceptable to mitigate the observed detrimental effects of HFS EF on plasma discharge, such as rotation braking, beta normalized drop by 10% or H-L back- transitions.

5. Runaway electron experiments Runaway electron (RE) experiments represented an important part of the COMPASS experimental programme since 2014. The experiments are motivated mainly by the urgent need to develop reliable RE beam mitigation strategy and gain better insight into the RE physics to secure safe future operation of ITER. The experiments on COMPASS are carried out in coordination with the EUROfusion MST1 devices and gained from the flexibility of the diagnostics set-up and relatively low safety constraints. The experimental programme was focused on a wide range of scenarios and phenomena, including post-disruptive RE beams [12], influence of MHD phenomena on the RE losses [13] or high Z impurity injection [14]. In low density current flattop discharges it appears that RE losses (measured using a HXR detector or a Cherenkov detector) are affected by various magnetic perturbations. First, regarding MHD instabilities, bursts of RE lossses after sawtooth crashes and periodic losses corresponding to rotation of magnetic islands in phase with maximum of current oscillation passing around low field side limiter have been observed [13]. Surprisingly, the islands not only enhance the losses, but there are discharges, where runaway losses were rather suppressed during the island existence. It seems that location of the RE population with respect to the islands plays a crucial role. Moreover, a link between the power source characteristic noise frequency and the frequency of HXR losses (1-2 kHz) in case of quiescent plasma has been confirmed with the exact mechanism yet to be uncovered. However the frequency of the HXR bursts seems also to be affected by the safety factor q at the edge. Bursts of MHD activity during the current quench of MGI disruptions also appeared to cause prompt RE losses [14]. Recently, the ramp-up MGI scenario was further investigated and it appears that higher toroidal is beneficial for the beam generation [14]. Moreover, the period just after RE beam generation is often accompanied by thin filaments observed by the fast camera that correlate with bursts in the ECE signal and fast events in other signals [15] including relativistic electron cyclotron emission [16] and detection of lost RE using the Cherenkov detector. [17]. The need of the better understanding of REs dynamics and the impact of injection of different gases or application of the resonant magnetic perturbation (RMP) on the RE beam suppression leads to development of the practically zero external loop voltage scenario via an active control of the constant current in the Central Solenoid of COMPASS [14]. The RMP experiments with n=1 and n=2 configurations using the low field side off-midplane coils were performed in the zero loop voltage scenario to study the effect on the post-disruption RE beam decay stage as well as on its pre-disruption phase. The strength of perturbation −2 depends on magnitude of current applied to RMP coils and can reach Br/BT ≈ 10 , however it was kept below the empirical threshold value of MHD mode locking. The increase of the RE current decay rate in the pre-disruption case (n=1) is proportional to RE losses demonstrated by an increase of the HXR signal (until HXR detectors reach the saturation level) and strongly depends on the magnitude and the toroidal phase angle between the bottom and upper RMP coils. According to predictive modeling obtained by the resistive MHD code MARS-F [16] the RE deconfinement is associated with the coupling between the rational magnetic equilibrium surfaces and plasma-screened RMP. These results are consistent with recent measurements [14, 20] also for the post-disruption (n=1) cases. The n=2 setting of RMP coils (2 possible orientations between the bottom and upper row) shows one more preferable configuration Fig. 9 (angle between bottom and upper row of coil is 90o) demonstrated by the RE beam suppression due to the formed MHD. This effect was less visible from the MARS-F modelling due to possible contribution of the non resonant response.

Figure 9: Spectrograms from the Mirnov coil C theta 13 localized at the HFS (upper panel) and HXR detector for the odd configuration where the angle between the top and bottom row of RMP coils is 90o. The red line - Ip, purple line - RMP, white line - HXR, cyan line - injection of impurity (Ar - 1.2 bar). The option of fine tuning of the RMP magnitude allows validation of the kinetic codes dealing with the radial transport and also analytical estimates of radiation transport as a contribution to the total losses. Lately, the studies were focused on the influence of different gases (Ar, Ne, D) on the generation and mitigation of the RE beam. In this case, a scenario with injection of rather small amount of gas into the current flattop of a low density discharge with already evolved RE population has been used. It appeared that an intensive injection of D may significantly slow down the current decay of RE beams triggered by Ar or Ne injection in the discharge phase with practically zero external loop voltage. The current decay rate scales with the amount of injected gas, however there is not as large difference between a slow piezo gas puff and the MGI. More Ne as compared to Ar is needed in order to achieve the same current decay rate as shown in Fig. 10. Also, the radiated power as calculated using AXUV tomography with limited number of views (LFS cameras affected by HXR from RE-wall) seems to be larger for Ne than for Ar in order to achieve the same current decay rate. Stability and decay of these RE beams is detailed in a separater contribution to this conference [18]. However, injection of any amount of Ar or Ne helps to prevent localised loss and damage by RE that is sometimes observed in extremely low-density discharges at COMPASS and which would be the worst case scenario in ITER.

Figure 10: The dependence of the current decay rate of the RE beam on the amount of injected gas normalised to the vessel volume for two different valves and gases. The higher amount. The flattop low density scenario has been used for this scan. Apparently more gas is more effective, however the dependence on large scale is far from linear and to some degree seems to saturate at higher gas amounts.

In present time an enhanced effort in modeling (simulations of transport of runaway electrons in the presence of MHD perturbation, kinetic Fokker-Planck modeling and coupling of kinetic and fluid codes) is ongoing. The results of the RE experiments on COMPASS should contribute to the joined European effort in the understanding of the RE physics and development of RE beam mitigation methods [13].

6. Conclusion

Despite its modest size, COMPASS tokamak is a versatile tool, which is conducting research relevant to ITER. Due to its unique diagnostics set, both inter- and intra-ELM heat fluxes were investigated and significant broadening of the power decay length was observed during ELM filaments. The peak energy density during ELMs agrees with model proposed by Eich. Recently, COMPASS tokamak has achieved the regime of partial detachment despite its short connection length and open divertor configuration, which enables further studies of this regime, which is relevant for ITER operation. Dedicated studies of the impact of misalignment of the central solenoid were performed thanks to a unique set of HFS RMP coils available at COMPASS. The influence of the misalignment on lock mode appearance and L-H transition was evaluated. An additional correction to the error field provided by the LFS error field coils proved that it is indeed possible to compensate for the central solenoid displacement and recover the original plasma performance. A complex series of experiments were performed focused on role of different gases on the decay of runaway beam current. Also, studies of the influence of RMP on the runaway beam current were performed

In all the fields above, i.e. ELM and divertor physics, error fields studies and runaway experiments on COMPASS, a significant progress under the joint EUROfusion effort has been achieved in 2017-18 and the results complemented and broadened the existing databases.

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