Session BO8: MF: Long Pulse Tokamaks, Spherical Tokamaks

Erosion of bulk TZM evaluated with an implanted depth marker in EAST

Erosion of molybdenum (TZM) at the mid-plane on the low and high field side in the Experimental Advanced Superconducting Tokamak (EAST) was characterized by utilizing a novel implanted depth marker. Exposure on the Material And Plasma Exposure System (MAPES) on the outboard mid-plane at low fluence (\textless 10\textasciicircum 24 m\textasciicircum -2) shows limited erosion, \textless 40 nm, near the sensitivity limit of the technique, x\textunderscore sens\textasciitilde 20 nm. Lithium (Li) and deuterium (D) content in the near surface (\textless 200 nm) was characterized by Elastic Recoil Detection (ERD) and Nuclear Reaction Analysis (NRA). Accumulation of Li from plasma exposure without prior Li wall conditioning is apparent, even with the sample surface positioned 3 cm radially outward of the limiter, exhibiting Li migration on the run-day time-scale, at low fluence. The HFS samples incorporated in modified tiles with dovetail slot design were exposed for a full run campaign, and net erosion of 0.5-1 um is observed. Erosion of bulk TZM plasma-facing components has been quantified for the first time with the novel implanted depth marker technique, which can be utilized with in situ wall monitoring diagnostics, such as Laser-Induced Breakdown Spectroscopy (LIBS) or Accelerator-based In-situ Material Surveillance (AIMS).   

\textbf{ELM Suppression by Boron Powder Injection: Analysis of Edge Turbulence}

Type I edge-localized modes (ELMs) in EAST were completely suppressed via boron powder injection into the X-point region of an upper-single null configuration over a wide range of operating conditions (2.8 \textless P$_{\mathrm{aux}}$ \textless 7.1 MW, 3.8e19 \textless n$_{\mathrm{e\thinspace }}$\textless 6e19 m$^{\mathrm{-3}}$, RF-only and RF$+$NBI heating scenarios). There appears to be a window of edge B concentration for stable long pulse operation. The injection of Boron coincides with the onset on many modes on the magnetic spectrograms showing multiple dependences with line-averaged densities. In this work, we will present analysis of mode localization using the radial density profile measured with O-mode sweeping reflectometry. Finally, initial analysis of power balance will be discussed.   

First demonstration of full suppression of type-I ELMs using n$=$4 RMP in EAST

Full suppression of Edge Localized Mode (ELM) by using ITER relevant $n=$4 resonant magnetic perturbation (RMP) is achieved for the first time in the type-I ELMy H-mode operational window with $q_{\mathrm{95}}\le $ 3.7 in the EAST tokamak with tungsten divertor. Different from previous observations using low $n$ ($n=$1 and 2) RMPs in higher $q_{\mathrm{95\thinspace }}(\ge $4) window, plasma confinement in this case does not drop during the transition from mitigation to full ELM suppression. Meanwhile, tungsten concentration is significantly reduced during ELM suppression. ELM suppression using $n=$3 RMP is also achieved in a similar window \quad with $q_{\mathrm{95}}$\textasciitilde 3.7 and $\beta _{\mathrm{N}}$\textasciitilde 2. Strong density pump-out is observed during full ELM suppression with high $n$ (3 and 4) RMPs. There are windows for both plasma density and $q_{\mathrm{95}}$ to access full ELM suppression with high $n$ RMPs. RMP spectrum also plays an important role to access ELM suppression. Footprint splitting in the heat flux on the divertor is observed for the first time in EAST. Detailed understanding of the 3D physics during ELM suppression will be presented. These results expand physics understanding and potential effectiveness of RMP for reliably controlling ELMs in future burning plasma devices such as ITER.   

Optimization of plasma current profile by lower hybrid current drive on EAST for steady-state operation with high internal inductance

A H-mode discharge with a relatively high value of the internal inductance $l_{\mathrm{i}}$ that may improve confinement and raise the stability limit to high $\beta_{\mbox{N}} $ has attracted considerable attentions in recent years[1], and it is proposed as one of promising scenarios to achieve steady-state (SS) operations with RF heating only on EAST. To assess how $l_{\mathrm{i}}$ varies with plasma parameters, predictive modeling and experiments are conducted for different operation parameters. It is shown by phase space analysis and GENRAY/CQL3D simulations that the variation of the parallel wave refractive index $N_{\mathrm{//}}$ which determines the power deposition profile of lower hybrid (LH) waves, depends mainly on the wave frequency, $q$ profile, and density profile, and so does the current profile[2]. The modeling indicates that a discharge with a lower density and plasma current is favorable to obtain a higher $l_{\mathrm{i}}$. A series of experiments are carried out in which only LH current drive and electron cyclotron heating are applied to sustain the SS H-mode discharges. The experimental results confirm that $l_{\mathrm{i}}$ decreases with the plasma density and current. The performance of plasma will be evaluated with more experiments and integrated modeling simulations over a broader parametric regime. [1] Ferron, et al., 2015 Nucl. Fusion 55, 073030. [2] Zhai, et al., 2019, Plasma Phys. Control. Fusion 61, 045002.   

The First H-mode Operation in Helium Plasma by RF-Heating on EAST

Physics understanding of helium plasmas are key issues for the first operational phase of ITER, in order to minimize activation of the in-vessel components. On EAST, a comprehensive set of H-mode experiments have been successfully performed in helium plasmas by pure RF-heating with ITER-like tungsten mono-block divertor to determine the requirement for low rotation operation and plasma-wall interaction in ITER. The H-mode threshold power is observed to be higher than the scaling law in deuterium plasma by a factor of 1.4 when density exceed n$_{\mathrm{min}}=$3.2x10$^{\mathrm{19}}$/m$^{\mathrm{3}}$ (n$_{\mathrm{min}}$: the density at the minimum power threshold) and increased rapidly towards lower density. For energy confinement and pedestal characteristics in helium plasmas, the helium concentration (He$_{\mathrm{conc}})$ is confirmed to play a crucial role. Type-I ELMy H-modes (f$_{\mathrm{ELM}}=$10-30Hz) have been achieved with high performance (H$_{\mathrm{98(y,2)\thinspace }}$\textgreater 1.0) when He$_{\mathrm{conc}}$\textless 60{\%}. Only Type-III ELMy H-mode could be realized under power injection (P$_{\mathrm{inj}})$ of 4.6MW while He$_{\mathrm{conc}}$ in the range of 60{\%}-70{\%}. It is hard to access H-mode in the condition of He$_{\mathrm{conc}}$\textgreater 70{\%} even with higher power around P$_{\mathrm{inj}}=$6.0MW. Small/no ELMs with good confinement are also found in helium H-mode plasmas. These investigations are carried out for various plasma parameters, plasma configurations and optimization of auxiliary heating methods. ELM stabilizations and suppressions are demonstrated by resonant magnetic perturbation and boron particles injection.   

Nonlinear MHD Modeling of the Effect of n$=$2 RMP on Peeling-Ballooning mode in KSTAR

To suppress edge-localized-modes (ELM) via resonant magnetic perturbation (RMP) is critical to reach and sustain high-performance steady state H-mode plasmas. Using the nonlinear 3D MHD code JOREK [1], we have successfully simulated a recent n$=$2 RMP-driven ELM-crash-suppression in KSTAR. In this study, we have found that such ELM-crash-suppression has been not only attributable to degraded pedestal but also to direct coupling between peeling-ballooning mode (PBM) and RMP-driven plasma response. Specifically, the pedestal pressure gradient was reduced, since radial transport was enhanced due to the formation of the stochastic layer and kink-peeling response (KPM) [2] driven by RMP. While the linear stability of PBM improved owing to the degraded pedestal, it was not a sole contributor to ELM-crash-suppression, in that the other nonlinear mode coupling should be simultaneously taken into account. This outcome is consistent with the previous studies [2, 3]. In addition, the locking of PBMs has been numerically simulated during the ELM suppression phase, which may support the relationship between V$_{\mathrm{ExB}}\approx $0 at the pedestal and the onset of ELM-crash-suppression. [1] G. T. A Huysmans et al., PPCF (2009) [2] F. Orain et al., Phys. Plasma (2019) [3] M. Becoulet et al., PRL (2014)   

VFT design and multi-pactor analysis for high power helicon current drive system in KSTAR

A helicon wave was confirmed to be coupled at low power and a high power coupling has been tried in KSTAR. When the RF is applied to the antenna system, the reflected power gradually increases by the timescale of sub-milliseconds. Rather slower process than usual arcing suggests that the reflection is caused by a multi-pactor discharge. In order to apply a high power helicon wave to the tokamak plasmas, it is important to mitigate or eliminate the multi-pactor discharge. A new VFT is fabricated following the design which focuses on the reduced total RF electric field and zero axial electric field on the TiN coated alumina surface. The disc type alumina window does not protrude in to the coaxial conductors so as to eliminate axial electric field. The unmatched impedance caused by the 1 cm thick alumina is compensated by the series matching element positioned in the pressurized section. For further understanding multi-pactor discharge at the helicon antenna system the multi-pactor test chamber (MTC) was developed. The high electric field is generated by the series unmaching-matching structures at the both sides of sample holder. The MTC is mainly used to evaluate muti-pactor avoiding technique and to develop conditioning processes.   

Shape and profile dependencies of RMP ELM suppression windows and their implications to 3D coils$^{\mathrm{1}}$

The operating windows for RMP ELM suppression in KSTAR have been greatly expanded in recent years across different target plasmas and 3D coil configurations. The window variations are qualitatively consistent with the multi-modal aspects of~ resonant responses, but have proven difficult to quantitatively predict when associated with modified shapes and profiles. New kinetic EFIT analyses show that a highly favorable condition features density profile broadening in addition to the higher triangularity discussed in earlier studies, which tends to increase the edge resonance relative to the core. The higher density over a wider region also implies higher tolerance to the core field penetration as seen in experiments, while its influence to the edge is relatively weaker as indicated by TM1 simulations. This improved physics understanding of RMP ELM suppression conditions has been then utilized to explore optimizations of the KSTAR 3D coils. Stellarator design tools such as FOCUS have been integrated with GPEC in the OMFIT framework to optimize 3D coil geometries and currents to efficiently access ELM suppression windows in a variety of KSTAR equilibria with differing shapes and kinetic profiles. These optimized 3D coils may be able to broaden access to ELM suppression in many KSTAR scenarios.   

Assessing the Impact of Light Impurities on Tungsten Sourcing Beyond the Divertor in WEST.

WEST is the first superconducting tokamak to have begun operations with all tungsten (W) plasma-facing components. To fully benefit from the opportunity to study W sources in this all W environment, including the relative contributions from divertor and main chamber, dedicated experimental sessions are ongoing. Initial experiments have indicated that intrinsic, light impurities (mainly oxygen and carbon) play a dominant role in the W sputtering for L-mode discharges. Nitrogen seeding experiments have also been conducted and the impact of N on W sourcing will be assessed. In the present study, the impact of light impurities on the poloidal distribution of W is also examined. Neutral W spectroscopy (W-I) measurements are taken from multiple poloidal locations to aid in characterizing the gross erosion rate around the vessel and incident particle fluxes are characterized by spectroscopy methods as well as Langmuir probes both imbedded and plunged. Far-SOL Collector Probes are used to constrain the light impurity contents in the SOL. Constrained by these measurements, Soledge2D-Eirene is used, starting with oxygen added to the background plasma, to examine the impact on W sputtering and its distribution beyond the divertor.   

Magnetic structure and activity during local helicity injection

Local helicity injection (LHI) is a non-solenoidal startup technique that uses biased plasma sources to inject DC magnetic helicity and drive current. For confident scaling to future devices, better understanding of the properties of LHI-generated plasmas and the mechanism(s) through which current is driven could prove crucial. Comparisons of Thomson scattering and radially scanning magnetic probe measurements find the magnetic boundary is shifted up to 8 cm outward relative to the kinetic edge during LHI. This suggests a two-region structure is present: an edge force-free region dominated by the current stream(s); and an inner confined region with a tokamak-like plasma. LHI-driven plasmas have greater magnetic activity than comparable Ohmic discharges, with the activity asymmetrically distributed and localized to the injector stream region. Following shutoff of the injectors, this power rapidly ($\approx $ 0.5 ms) decays and symmetrizes, and the remaining activity resembles that of an Ohmic-driven discharge. Broadband activity during LHI has power law behavior resembling that of Alfv\'{e}nic turbulence which is expected to cascade helicity to large scales and has been observed in other reconnecting systems. The role of this activity in helicity transport and current drive, along with high frequency activity $f\approx 2$ MHz ($\approx $ 2--4 $f_{ci} )$ that increases with LHI drive, is being investigated.   

\textsc{Urania:} A Spherical Tokamak for Developing Non-Solenoidal Plasma Startup Techniques

Development of a routine non-solenoidal startup technique is a critical issue facing the spherical tokamak. A major upgrade to the \textsc{Pegasus }program is underway to develop and compare leading non-solenoidal startup techniques on a dedicated solenoid-free facility---the Unified Reduced $A$ Non-Inductive Assessment (\textsc{Urania}) experiment. Facility upgrades for \textsc{Urania} include: increased $B_{T} $ to 0.6 T, extended pulse duration ($\le 100$ ms), an enhanced poloidal field set for improved shape control, and an expanded diagnostic suite, while retaining an ultra-low aspect ratio ($A\approx 1.2)$. Modelling and experiments have informed the design of a new local helicity injection (LHI) system capable of $I_{p} \le 0.3$ MA by leveraging the enhanced $B_{T} $ to increase the Taylor limit early in the discharge. The new LHI system is designed to optimize helicity input and increase Taylor limit via a non-circular current source geometry. A prototype of this new source is in fabrication for testing on \textsc{Pegasus}. A novel system capable of transient and sustained coaxial helicity injection is in design. A modest power (200--400 kW) electron Bernstein wave radiofrequency heating and current drive system will be deployed.   

Progress Toward First Plasma on MAST Upgrade

The MAST Upgrade spherical tokamak has unique capabilities to address some of the key issues facing the development of fusion energy. Its main objectives are: 1) development of novel exhaust concepts, 2) contribution to the knowledge base for ITER and 3) to explore potential routes to smaller/cheaper fusion reactors. To fulfil these aims, it is equipped with 19 new poloidal field coils and closed divertors with Super-X capability. BT has been increased by 50{\%} and the pulse length and Ip have increased to 5s and 2MA respectively. Auxiliary heating is provided by on and off axis NBI. The divertors are diagnosed with probes, bolometers, Thomson scattering, IR, visible imaging and spectroscopy. Fast ion physics studies are enhanced with a new fast ion loss detector. The construction of MAST Upgrade is complete and commissioning is well underway. Progress toward first plasma will be presented, including integrated commissioning of the tokamak subsystems and calibration of the magnetics sensors. Plans for the first experimental campaign will also be presented. Work supported by RCUK [grant number EP/I501045] and Euratom.   

Tokamak Energy and the high-field spherical tokamak route to fusion power

Tokamak Energy is a privately funded company based in the UK with a mission to deliver a faster route to fusion. Founded in 2009, Tokamak Energy is developing compact fusion power plants based on two promising technologies: Spherical Tokamaks (STs) and magnets made from High Temperature Superconductors (HTS). The inherent compactness and improved efficiency of the spherical tokamak, coupled with the favourable properties of HTS magnets, open a route to efficient power production at significantly lower net power outputs than previously considered possible. Currently, two main development streams are progressing in parallel: i) advancing high field spherical tokamak physics and engineering on ST40, the highest field ($B_T=3$T) device of its kind; and ii) HTS technology development, which has characterised tape performance and developed key technologies and is now focused on demonstrating a high field tokamak magnet system at substantial scale. In the next stage of development, these new technologies and understanding will be combined to deliver the world’s first device capable of fusion energy gain and industrial scale power production – ST-F1. ST-F1 is currently in the concept design stage and is aiming to demonstrate the viability of commercial fusion.   

Overview of the Lockheed Martin Compact Fusion Reactor (CFR) Project

The Lockheed Martin Compact Fusion Reactor (CFR) Program endeavors to quickly develop a compact fusion power plant with favorable commercial economics and military utility. The CFR uses a diamagnetic, high beta, magnetically encapsulated, linear ring cusp plasma confinement scheme. Major project activities will be reviewed, including the T4B and T5 plasma heating experiments. The goal of the experiments is to demonstrate a suitable plasma target for heating experiments, to characterize the behavior of plasma sources in the CFR configuration and to then heat the plasma with neutral beams, with the plasma transitioning into the high Beta confinement regime. The design and preliminary results of the experiments will be presented, including discussion of predicted behavior, plasma sources, heating mechanisms, diagnostics suite and relevant numerical modeling. \copyright 2019 Lockheed Martin Corporation. All Rights Reserved.