Fully non-inductive plasma maintenance was achieved by a microwave of 8.2 GHz and 40 kW for more than 1 h 55 min with a well-controlled plasma-facing wall (PFW) temperature of 393 K, using a hot wall in the middle-sized spherical tokamak QUEST, until the discharge was finally terminated by the uncontrollability of the density. The PFW was composed of atmospheric plasma-sprayed tungsten and stainless steel. The hot wall plays an essential role in reducing the amount of wall-stored hydrogen and facilitates hydrogen recycling. The behaviour of fuel hydrogen in the PFW was investigated by monitoring the injection and evacuation of hydrogen into and from the plasma-producing vessel. A fuel particle balance equation based on the presence of a hydrogen transport barrier between the deposited layer and the substrate was applied to the long-duration discharges. It was found that the model could readily predict the observed behaviour in which a higher wall temperature likely gives rise to faster wall saturation.
The plasma current is ramped up primarily by a 28 GHz electron cyclotron wave (ECW) in the Q-shu University experiment Steady-State Spherical Tokamak (QUEST), with multiple harmonic resonance layers from the second to the fourth stay in the plasma core. A steering antenna comprising two quasi-optical mirrors enhances the power density of ECWs. The ECW beam is injected obliquely from the low-field side where the parallel refractive index is N∥ = 0.75 at the second-harmonic resonance layer. Analysis of the resonance condition has found that energetic electrons moving forward along the magnetic field resonate more effectively than those moving backward. Such symmetry breaking is consistent with the results of the current ramp-up experiment. The peak plasma current reaches Ip>70 kA, constantly injecting a beam of radio frequency power of 100 kW. Ray-tracing by the TASK/WR code demonstrates that the power of the 28 GHz extraordinary mode is absorbed by energetic electrons via single-pass cyclotron absorption.
Initial results from the recently implemented transient coaxial helicity injection (CHI) system on QUEST are reported. QUEST uses a new CHI electrode configuration in which the CHI insulator is not part of the vacuum boundary, making this configuration easier to implement in fusion reactors. Experimental results show that transient CHI startup in this alternate electrode configuration is indeed possible. Reliable gas breakdown was achieved, and toroidal currents up to 45 kA were generated.
Coaxial Helicity Injection (CHI) has now been implemented in QUEST. The goals for the first transient CHI experiments were to establish reliable gas breakdown conditions, and to measure CHI-produced toroidal current generation. Both these objectives were successfully met. Toroidal currents up to 29 kA were measured. Interestingly, these first plasmas on QUEST also suggest the formation of small amounts of closed magnetic flux surfaces. Establishment of methods for solenoid-free plasma current start-up, with robust and stable non-inductive current drive on tokamaks increases the prospects for a compact and low aspect ratio fusion reactor, which has the advantages of lower construction cost and higher plasma beta. QUEST is a mid-size spherical tokamak (ST) device [1], in which plasmas are produced mainly by electron cyclotron heating (ECH), and with minimal use of the central solenoid. Non-inductive toroidal plasma currents up to 70 kA have been achieved on QUEST using 28 GHz microwave injection [2]. We have now implemented coaxial helicity injection (CHI) capability on QUEST. We aim to generate more than 100 kA of initial toroidal plasma current with transient CHI and ramp this seed current noninductively using the high-power 28 GHz microwave system. In this paper we report results from the first CHI experiments on QUEST.CHI [3], a useful method for non-inductive current drive, has been studied in the HIT-II and NSTX STs, in which the compatibility of CHI startup with conventional Ohmic heating and current drive has been verified [4,5]. It is projected that the capability of CHI for current startup would be significantly enhanced if ECH is used to heat the CHI plasma, and this is planned to be tested in NSTX-U for a full non-inductive current start-up and ramp-up scenario [6]. author's e-mail: kuroda@triam.kyushu-u.ac.jpQUEST is equipped with important capabilities that will extend CHI studies to new parameter regimes. The internal vessel walls on QUEST are all metallic; this should be beneficial to CHI as it would reduce the influx of low-Z impurities in to the plasma, which are the prime source of energy loss in low temperature plasmas. The CHI electrode configuration on QUEST is also simpler, and different than the ones on HIT-II and NSTX, and it may be easier to implement this configuration in a fusion reactor [7]. The primary difference is that on NSTX and HIT-II, the insulator is also the vacuum boundary, whereas on QUEST, the insulator is sandwiched between the electrode plate and the divertor plate. However, CHI start-up in this new electrode configuration needs to be demonstrated. This is an important part of the CHI program on QUEST. Figure 1 shows the side view of the QUEST internal structure. The nominal toroidal field at the machine axis (at R 0 = 0.64 m) is 0.25 T at its maximum limit. There are four poloidal field coils above and below the vessel mid-plane; these are normally operated in a series configuration, with two coils connected to a single power supply. As shown in Fig. 1, we installed an e...
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