Diverted negative triangularity plasmas on DIII-D: the benefit of high confinement without the liability of an edge pedestal

Diverted discharges at negative triangularity on the DIII-D tokamak sustain normalized confinement and pressure levels typical of standard H-mode scenarios (H (98y2) similar or equal to 1, beta (N) similar or equal to 3) without developing an edge pressure pedestal, despite the auxiliary power far exceeding the L -> H power threshold expected from conventional scaling laws. The power degradation of confinement is substantially weaker than the ITER-89P scaling, resulting in a confinement factor that improves with increasing auxiliary power. The absence of the edge pedestal is beneficial in several ways, such as eliminating the need for active mitigation or suppression of edge localized modes, low impurity retention and a reconstructed scrape-off layer heat flux width at the mid-plane that exceeds the ITPA multi-machine scaling law by up to 50%. Together with technological advantages granted by placing the divertor at larger radii, plasmas at negative triangularity without an edge pedestal feature both core confinement and power handling characteristics that are potentially suitable for operation in future fusion reactors.

Advanced RF heating schemes in preparation for ICRF heating and fastion experiments in the W7-X stellarator

Jonathan Graves, Mike Machielsen, Hamish William Patten

Fast ion confinement is crucial for the demonstration of the stellarator approach towards fu- sion energy. To study confinement of fast ions with today’s stellarators advanced RF heating schemes can be used to generate fast ions. Secondly, advanced RF heating schemes are de- veloped to improve ion heating performance. Advanced ICRH schemes (e.g. the 3-ion species scheme [1]) have the advantage of improved polarization, so the wave energy is nearly com- pletely carried by the left hand polarized wave at resonance, thereby providing good power transfer to the ions. However, modelling of minority and 3-ion species heating in W7-X has shown that the fast ion tails are limited, and core heating is modest [3]. Additionally, the highly anisotropic distributions generated by the minority and 3-ion species schemes are not well con- fined. Furthermore, the high density plasma of W7-X creates difficulty heating thermal particles because of the high collisionality. An advantage of the so called RF-NBI synergetic scheme is that NBI born ions are weakly collisional at birth. Secondly, compared to other ICRH schemes the RF-NBI scheme heats more in the parallel direction, generating less trapped ions. Thirdly, a more isotropic velocity distribution is also desirable for fast ion studies since fusion born alphas are isotropic as well [2, 3]. The code SCENIC [4] is used to model ICRH scenarios in 3D. The code is comprised of a magnetic equilibrium code, a full wave code and a Fokker-Planck code. It is able to determine wave propagation and absorption in hot plasma with full FLR effects. Lastly, effects of finite orbit width effects and wave absorption on the distribution function are taken into account. This contribution will explore RF-NBI and other advanced heating schemes in W7-X relevant plasma scenarios prior to ICRF heating and fast ion experiments. [1] Ye. O. Kazakov, D. Van Eester, R. Dumont and J. Ongena, Nucl. Fusion 55, 3 (2015) [2] H. Patten et al., in preparation of Phys. Rev. Lett (2019) [3] H. Patten, "Development and optimisation of advanced auxiliary ion heating schemes for 3D fusion plasma devices", EPFL, (2018) [4] M. Jucker, "Self-consistent ICRH distribution functions and equilibria in magnetically confined plasmas", EPFL, (2010)

European Physical Society2019

Overview of ASDEX upgrade results

Timothy Goodman, Emanuele Poli, Olivier Sauter, Giovanni Tardini, Dávid Wágner, Hartmut Zohm

Recent results from the ASDEX Upgrade experimental campaigns 2001 and 2002 are presented. An improved understanding of energy and particle transport emerges in terms of a 'critical gradient' model for the temperature gradients. Coupling this to particle diffusion explains most of the observed behaviour of the density profiles, in particular, the finding that strong central heating reduces the tendency for density profile peaking. Internal transport barriers (ITBs) with electron and ion temperatures in excess of 20 keV (but not simultaneously) have been achieved. By shaping the plasma, a regime with small type 11 edge localized modes (ELMs) has been established. Here, the maximum power deposited on the target plates was greatly reduced at constant average power. Also, an increase of the ELM frequency by injection of shallow pellets was demonstrated., ELM free operation is possible in the quiescent H-mode regime previously found in DIII-D which has also been established on ASDEX Upgrade. Regarding stability, a regime with benign neoclassical tearing modes (NTMs) was found. During electron cyclotron current drive (ECCD) stabilization of NTMs, beta(N) could be increased well above the usual onset level without a reappearance of the NTM. Electron cyclotron resonance heating and ECCD have also been used to control the sawtooth repetition frequency at a moderate fraction of the total heating power. The inner wall of the ASDEX Upgrade vessel has increasingly been covered with tungsten without causing detrimental effects on the plasma performance. Regarding scenario integration, a scenario with a large fraction of noninductively driven current (greater than or equal to50%), but without ITB has been established. It combines improved confinement (tau(E)/tau(ITER98) approximate to 1.2) and stability (beta(N) less than or equal to 3.5) at high Greenwald fraction (n(e)/n(GW) approximate to 0.85) in steady state and with type 11 ELMy edge and would offer the possibility for long pulses with high fusion power at reduced current in ITER.

2003

The role of the sheath in magnetized plasma turbulence and flows

Joaquim Loizu Cisquella

Controlled nuclear fusion could provide our society with a clean, safe, and virtually inexhaustible source of electric power production. The tokamak has proven to be capable of producing large amounts of fusion reactions by confining magnetically the fusion fuel at sufficiently high density and temperature, thus in the plasma state. Because of turbulence, however, high temperature plasma reaches the outermost region of the tokamak, the Scrape-Off Layer (SOL), which features open magnetic field lines that channel particles and heat into a dedicated region of the vacuum vessel. The plasma dynamics in the SOL is crucial in determining the performance of tokamak devices, and constitutes one of the greatest uncertainties in the success of the fusion program. In the last few years, the development of numerical codes based on reduced fluid models has provided a tool to study turbulence in open field line configurations. In particular, the GBS (Global Braginskii Solver) code has been developed at CRPP and is used to perform global, three-dimensional, full-n, flux-driven simulations of plasma turbulence in open field lines. Reaching predictive capabilities is an outstanding challenge that involves a proper treatment of the plasma-wall interactions at the end of the field lines, to well describe the particle and energy losses. This involves the study of plasma sheaths, namely the layers forming at the interface between plasmas and solid surfaces, where the drift and quasineutrality approximations break down. This is an investigation of general interest, as sheaths are present in all laboratory plasmas. This thesis presents progress in the understanding of plasma sheaths and their coupling with the turbulence in the main plasma. A kinetic code is developed to study the magnetized plasma-wall transition region and derive a complete set of analytical boundary conditions that supply the sheath physics to fluid codes. These boundary conditions are implemented in the GBS code and simulations of SOL turbulence are carried out to investigate the importance of the sheath in determining the equilibrium electric fields, intrinsic toroidal rotation, and SOL width, in different limited configurations. For each study carried out in this thesis, simple analytical models are developed to interpret the simulation results and reveal the fundamental mechanisms underlying the system dynamics. The electrostatic potential appears to be determined by a combined effect of sheath physics and electron adiabaticity. Intrinsic flows are driven by the sheath, while turbulence provides the mechanism for radial momentum transport. The position of the limiter can modify the turbulence properties in the SOL, thus playing an important role in setting the SOL width.

EPFL2013