Non-inductive Plasma Start-up Research Articles

Development of an outer-off-midplane lower hybrid wave launcher for improved core absorption in non-inductive plasma start-up on TST-2

A new outer-off-midplane lower hybrid (LH) wave launcher was developed for improved core absorption in the TST-2 spherical tokamak. In the previously developed outer-midplane and top launch scenarios, plasma current was driven only very near the plasma edge (), resulting in limited driven current density. Strong thick-target x-ray radiation has also been observed for the top launch scenario, which indicated LH wave driven fast electron losses. The outer-off-midplane launch scenario was designed to drive current at the core () at low phase velocity to suppress production of unnecessarily high energy electrons. In the non-inductive plasma start-up experiment with the new outer-off-midplane launch scenario, improved core electron heating was confirmed by the Thomson scattering diagnostic that showed the electron temperature was about twice as high as the previous two scenarios. Reduction of the LH wave driven fast electron losses was confirmed by the measured x-ray radiation intensity that was substantially lower than the top launch scenario. It was also easier to maintain the optimum level of the electron density compared to the outer-midplane launch scenario that indicated the undesired LH wave interactions with the scrape-off-layer plasma may have been reduced due to strong core absorption.

Investigation of the effectiveness of ‘multi-harmonic’ electron cyclotron current drive in the non inductive EXL-50 ST

The fully non-inductive spherical tokamak EXL-50, built and operated by the ENN private limited company, has routinely achieved high current drive efficiency of ∼ 1 A/W in only ECRH powered experiments. We have numerically investigated the effectiveness of multiple electron cyclotron resonance (ECR) harmonics in generating such a high efficiency of electron cyclotron current drive (ECCD) in non-inductive plasma start-up. The Fokker-Planck equation is numerically solved to obtain the electron distribution function, under the steady state of relativistic nonlinear Coulomb collision and quasi-linear diffusion operators, for calculating plasma current driven by the injected EC waves. Multi-pass absorption simulations, done with the CQL3D code for extra-ordinary EC waves, demonstrate over 1 A/W efficiency in current for a relatively low density (∼ 2 × 1018 m −3), and low temperature (∼ 100 eV) plasma, consistent with the experimental results observed on EXL-50. Systematic scanning of different ECR harmonics in simulation has revealed that the multi-harmonic resonance interaction in EXL-50 plays a pivotal role in generating the energetic electron tail responsible for the current.

Development of an electron cyclotron resonance heating and electron Bernstein wave current drive system on ST40

A mutli-frequency electron cyclotron resonance heating and current drive system is currently under construction at the ST40 spherical tokamak. The system employs two gyrotrons with a maximum output power of 1 MW each. Both gyrotrons can be tuned to operate at either 105 GHz or 140 GHz. The system is designed to study a non-inductive plasma start-up, current ramp-up and sustainment. Detailed modelling of the system capabilities has been conducted for various RF launch configurations in ST40.

Observation of second harmonic electron cyclotron resonance heating and current-drive transition during non-inductive plasma start-up experiment in QUEST

Noninductive plasma current start-up using 2nd harmonic electron cyclotron resonance heating (ECRH) with oblique radio frequency (RF) injection is demonstrated in a Q-shu University experiment with steady-state spherical tokamak. A strong transition was observed in the heating and plasma current ramp-up. The initial bulk electron heating regime exhibits T ebulk ∼ 140 eV and no hard x-ray (HXR) emission with a low I p of ∼15 kA; it abruptly transitions to a regime that exhibits a low T ebulk of ∼10 eV and a strong HXR emission with a high I p of ∼50 kA. This behavior is distinctly different from that observed in previous fundamental ECRH experiments. The mechanism of the heating and current drive transition are investigated considering wave power absorption and plasma power balance. The results indicate that the transition is caused by the favorable heating of tail electrons where the RF power absorption at the 2nd harmonic increases nearly linearly with T etail, while the power transfer from the tail electrons to the bulk electrons decreases with 1/T etail 0.5. This causes a rapid transition to a state with high T etail while reducing T ebulk towards colder ion temperature. The understanding of the transition mechanism helps to consider plasma current start-up using 2nd harmonic ECRH for tokamak reactors such as JT-60 SA and ITER.

Application of transient CHI plasma startup to future ST and AT devices

Employment of non-inductive plasma start-up techniques would considerably simplify the design of a spherical tokamak fusion reactor. Transient coaxial helicity injection (CHI) is a promising method, expected to scale favorably to next-step reactors. However, the implications of reactor-relevant parameters on the initial breakdown phase for CHI have not yet been considered. Here, we evaluate CHI breakdown in reactor-like configurations using an extension of the Townsend avalanche theory. We find that a CHI electrode concept in which the outer vessel wall is biased to achieve breakdown, while previously successful on NSTX and HIT-II, may exhibit a severe weakness when scaled up to a reactor. On the other hand, concepts which employ localized biasing electrodes such as those used in QUEST would avoid this issue. Assuming that breakdown can be successfully attained, we then apply scaling relationships to predict plasma parameters attainable in the transient CHI discharge. Assuming the use of 1 Wb of injector flux, we find that plasma currents of 1 MA should be achievable. Furthermore, these plasmas are expected to Ohmically self-heat with more than 1 MW of power as they decay, facilitating efficient hand-off to steady-state heating sources. These optimistic scalings are supported by Tokamak Simulation Code simulations.

Development of equilibrium fitting code using finite element method in versatile experiment spherical torus

An equilibrium fitting code using finite element method is developed for various plasma studies, especially on non-inductive start-up and ECH heating & current drive experiments in Versatile Experiment Spherical Torus. The magnetic diagnostics are applied to provide the basic equilibrium fitting constraints. As severe eddy currents are induced by not only the PF coils but also the plasma current itself, the eddy current component from the magnetic signals is dealt with the vessel wall model in the eddy current solver and carefully compensated, which is essential for accurate equilibrium fitting. The fast camera image analysis method is adopted and the limiter current monitoring system is developed for validating the results of the plasma position and shape from the code. The equilibrium fitting code is applied to the limited and diverted plasma configuration in VEST and the time evolution of the plasma position and shape and even averaged magnetohydrodynamic parameters such as the edge safety factor and the internal inductance are extracted. Additionally, the equilibrium fitting code is used to analyze the non-inductive plasma start-up experiment using the direct current helicity injection in VEST. Especially, it is the first time to apply the finite element method to direct current helicity injection experiment and ECH heated plasmas which have significant amount of plasma current flowing along the open field line.

Numerical modeling of lower hybrid current drive in fully non-inductive plasma start-up experiments on TST-2

Non-inductive plasma start-up is a critical issue for spherical tokamaks since there is not enough room to provide neutron shielding for the center solenoid. Start-up using lower hybrid (LH) waves has been studied on the TST-2 spherical tokamak. Because of the low magnetic field of a spherical tokamak, the plasma density needs to be kept at a very low value during the plasma current ramp-up so that the plasma core remains accessible to the LH waves. However, we have found that higher density was required to sustain larger plasma current. The achievable plasma current was limited by the maximum operational toroidal field of TST-2. The existence of an optimum density for LH current drive and its toroidal field dependence is explained through a numerical simulation based on a ray tracing code and a Fokker–Planck solver. In order to access higher density at the same magnetic field, a top-launch antenna was recently installed in addition to the existing outboard-launch antenna. Increase in the density limit was observed when the power was launched from the top antenna, consistently with the numerical predictions.

Fully non-inductive plasma start-up with lower-hybrid waves using the outboard-launch and top-launch antennas on the TST-2 spherical tokamak

Removal of the central solenoid is essential to realize an economical spherical tokamak fusion reactor, but non-inductive plasma start-up is a challenge. On the TST-2 spherical tokamak, non-inductive plasma start-up using lower-hybrid (LH) waves has been investigated. Using the capacitively-coupled combline (CCC) antenna installed at the outboard midplane, fully non-inductive plasma current ramp-up up to a quarter of that of the typical Ohmic discharges has been achieved. Although it was desirable to keep the density low during the plasma current ramp-up to avoid the LH density limit, it was recognized that there was a maximum current density that could be carried by a given electron density. Since the density needed to increase as the plasma current was ramped-up, the achievable plasma current was limited by the maximum operational toroidal field of TST-2. The top-launch CCC antenna was installed to access higher density with up-shift of the parallel index of refraction. Numerical analysis of LH current drive with the outboard-launch and top-launch antennas was performed and the results were qualitatively consistent with the experimental observations.

Power Balance Estimation in Long Duration Discharges on QUEST

Fully non-inductive plasma start-up was successfully achieved by using a well-controlled microwave source on the spherical tokamak, QUEST. Non-inductive plasmas were maintained for approximately 3–5 min, during which time power balance estimates could be achieved by monitoring wall and cooling-water temperatures. Approximately 70%–90% of the injected power could be accounted for by calorimetric measurements and approximately half of the injected power was found to be deposited on the vessel wall, which is slightly dependent on the magnetic configuration. The power distribution to water-cooled limiters, which are expected to be exposed to local heat loads, depends significantly on the magnetic configuration, however some of the deposited power is due to energetic electrons, which have large poloidal orbits and are likely to be deposited on the plasma facing components.

Non-inductive plasma start-up experiments on the TST-2 spherical tokamak using waves in the lower-hybrid frequency range

Non-inductive plasma current start-up and sustainment by waves in the lower-hybrid frequency range (200 MHz) have been studied on the TST-2 spherical tokamak (R0 ⩽ 0.38 m, a ⩽ 0.25 m, Bt0 ⩽ 0.3 T, Ip ⩽ 0.14 MA) using three types of antenna: the 11-element inductively-coupled combline antenna, the dielectric loaded 4-waveguide array antenna, and the 13-element capacitively-coupled combline (CCC) antenna. The maximum plasma currents of 15 kA, 10 kA and 16 kA were achieved, respectively. The highest current drive figure of merit was achieved by the CCC antenna. The efficiency of current drive should improve by reducing prompt orbit losses of high energy electrons by operating at higher plasma current (to improve orbit confinement) and higher toroidal magnetic field (to improve wave accessibility to the plasma core), while keeping the density high enough (to avoid excessive acceleration of electrons), but under the ‘density limit’.

Non-inductive plasma start-up on NSTX and projections to NSTX-U using transient CHI

Transient coaxial helicity injection (CHI) in the National Spherical Torus Experiment (NSTX) has generated toroidal current on closed-flux surfaces without the use of the central solenoid. When induction from the central solenoid was added, CHI initiated discharges in NSTX achieved 1 MA of plasma current using 65% of the solenoid flux of standard induction-only discharges. In addition, the CHI-initiated discharges have lower density and a low normalized internal plasma inductance of 0.35, as desired for achieving advanced scenarios. Transient CHI will be used for non-inductive plasma start-up on the upgrade to NSTX (NSTX-U) that is now under construction. It will have numerous improvements that significantly enhance CHI start-up capability. The TSC code is now starting to be used for full discharge simulation, which includes solenoid-less start-up with CHI and subsequent non-inductive current ramp-up using neutral beams. These results suggest that the increased injector flux capability of NSTX-U should allow CHI start-up at more than the 400 kA level and that the new tangential neutral beam system on NSTX-U should be able to ramp the current to the 1 MA levels.

Merging Experimental Study of High-Beta ST Formation for Non-Inductive Plasma Start-Up Assisted by NBI in TS-4

Spherical tokamak (ST) merging start-up for high-beta ST formation was developed in the TS-4 merging/reconnection experiment under a high external toroidal magnetic field. We optimized the formation of a stable high-beta ST (βt > 30%) by adjusting the initial parameters of two low-beta STs. We precisely analyzed the produced high-beta STs by MEUDAS using the measured and fitted profiles, and found that they exhibited low-plasma internal inductance because of a hollow current profile, a paramagnetic toroidal field, and weakly reversed shear. A pressure-driven instability analysis indicates that the merging STs become unstable when their q-values are less than 1. In the first high-power neutral beam injection experiment (#1 and #2, PNBI ∼ 0.4MW) in TS-4, the decay time and magnetic flux of the produced STs were improved, mainly because of pre-ionization and merging/reconnection effects. c

Noninductive plasma initiation and startup in the DIII-D tokamak

Noninductive plasma startup with second harmonic electron cyclotron heating has been studied in DIII-D. Plasma currents up to 33 kA have been obtained in this phase. The maximum current obtained was primarily limited by the experimental time and does not necessarily represent the highest achievable current. The dominant physical mechanism for the observed current is the Pfirsch–Schlüter current on open field lines. Closed flux surfaces have been observed, but in a rather limited region, compared with previous experiments in spherical tokamaks. This method of plasma initiation and initial current ramp might be used in future burning plasma devices in conjunction with other current drive techniques to provide a fully noninductive startup.

Implications of NSTX lithium results for magnetic fusion research

Lithium wall coating techniques have been experimentally explored on National Spherical Torus Experiment (NSTX) for the last five years. The lithium experimentation on NSTX started with a few milligrams of lithium injected into the plasma as pellets and it has evolved to a lithium evaporation system which can evaporate up to ∼100 g of lithium onto the lower divertor plates between lithium re-loadings. The unique feature of the lithium research program on NSTX is that it can investigate the effects of lithium in H-mode divertor plasmas. This lithium evaporation system thus far has produced many intriguing and potentially important results; the latest of these are summarized in a companion paper by H. Kugel. In this paper, we suggest possible implications and applications of the NSTX lithium results on the magnetic fusion research which include electron and global energy confinement improvements, MHD stability enhancement at high beta, edge localized mode (ELM) control, H-mode power threshold reduction, improvements in radio frequency heating and non-inductive plasma start-up performance, innovative divertor solutions and improved operational efficiency.

The national spherical torus experiment (NSTX) research programme and progress towards high beta, long pulse operating scenarios

A major research goal of the national spherical torus experiment is establishing long-pulse, high beta, high confinement operation and its physics basis. This research has been enabled by facility capabilities developed during 2001 and 2002, including neutral beam (up to 7 MW) and high harmonic fast wave (HHFW) heating (up to 6 MW), toroidal fields up to 6 kG, plasma currents up to 1.5 MA, flexible shape control, and wall preparation techniques. These capabilities have enabled the generation of plasmas with of up to 35%. Normalized beta values often exceed the no-wall limit, and studies suggest that passive wall mode stabilization enables this for H mode plasmas with broad pressure profiles. The viability of long, high bootstrap current fraction operations has been established for ELMing H mode plasmas with toroidal beta values in excess of 15% and sustained for several current relaxation times. Improvements in wall conditioning and fuelling are likely contributing to a reduction in H mode power thresholds. Electron thermal conduction is the dominant thermal loss channel in auxiliary heated plasmas examined thus far. HHFW effectively heats electrons, and its acceleration of fast beam ions has been observed. Evidence for HHFW current drive is obtained by comparision of the loop voltage evolution in plasmas with matched density and temperature profiles but varying phases of launched HHFW waves. Studies of emissions from electron Bernstein waves indicate a density scale length dependence of their transmission across the upper hybrid resonance near the plasma edge that is consistent with theoretical predictions. A peak heat flux to the divertor targets of 10 MW m−2 has been measured in the H mode, with large asymmetries being observed in the power deposition between the inner and outer strike points. Non-inductive plasma startup studies have focused on coaxial helicity injection. With this technique, toroidal currents up to 400 kA have been driven, and studies to assess flux closure and coupling to other current drive techniques have begun.

Exploration of spherical torus physics in the NSTX device

The National Spherical Torus Experiment (NSTX) is being built at Princeton Plasma Physics Laboratory to test the fusion physics principles for the spherical torus concept at the MA level. The NSTX nominal plasma parameters are R0 = 85 cm, a = 67 cm, R/a ⩾ 1.26, Bt = 3 kG, Ip = 1 MA, q95 = 14, elongation κ ⩽ 2.2, triangularity δ ⩽ 0.5 and a plasma pulse length of up to 5 s. The plasma heating/current drive tools are high harmonic fast wave (6 MW, 5 s), neutral beam injection (5 MW, 80 keV, 5 s) and coaxial helicity injection. Theoretical calculations predict that NSTX should provide exciting possibilities for exploring a number of important new physics regimes, including very high plasma β, naturally high plasma elongation, high bootstrap current fraction, absolute magnetic well and high pressure driven sheared flow. In addition, the NSTX programme plans to explore fully non-inductive plasma startup as well as a dispersive scrape-off layer for heat and particle flux handling.

Spherical tokamaks with outboard stellarator windings

The possibility of adding stellarator features to the extremely compact configuration of a spherical tokamak (ST) is analysed. This is accomplished by introducing outboard stellarator windings (OSWs). Two types of OSW are studied, a classical stellarator type and a torsatron type, and both are shown to be effective. Numerical calculations are presented for a simplified model of the NSTX tokamak. Among the main potential advantages of using OSWs in a ST are the possibility of noninductive plasma startup or improved inductive operation caused by the existence of closed vacuum flux surfaces and an external rotational transform