Identification of the beta limit in ASDEX Upgrade

Tokamak plasma is a subject of various resistive and ideal MHD instabilities which restrict the operation space of the device. For largest fusion outcome, it is preferable to operate the tokamak close to the stability limit with maximal possible pressure characterized by the value of normalized beta, , and thus maximal fusion power . In ASDEX Upgrade, the limit for maximal achievable is typically set by the resistive instabilities (tearing modes). If these instabilities are overcome or prevented, higher values of the normalized beta can be reached limited by the onset of the ideal kink instability. The actual limit depends on several factors, including the stabilizing influence of the conducting components facing the plasma surface. At present, the wall elements are far from the plasma and the stability boundary is expected to be close to the “no wall” limit (no stabilizing wall effect). Investigation of maximum achievable βN values for the different operation scenario in ASDEX Upgrade is presented in this work. Two indicators are used to detect the stability boundary:Increase of the resonant field amplification (RFA)Onset of the ideal kink modeRecently installed internal active coils are used to probe stability of the plasma by the RFA technique. A wide range of different MHD diagnostics are used to identify the behaviour and structure of MHD modes in different discharges with high . Experimentally obtained results are compared with the results of the numerical modelling with linear MHD codes CASTOR-FLOW and MARS-K. Such comparison allows to validate the plasma model used in the codes and therefore to make numerical projection for further experimental studies

Identification of the beta limit in ASDEX Upgrade

Tokamak plasma is a subject of various resistive and ideal MHD instabilities which restrict the operation space of the device. For largest fusion outcome, it is preferable to operate the tokamak close to the stability limit with maximal possible pressure characterized by the value of normalized beta, , and thus maximal fusion power . In ASDEX Upgrade, the limit for maximal achievable is typically set by the resistive instabilities (tearing modes). If these instabilities are overcome or prevented, higher values of the normalized beta can be reached limited by the onset of the ideal kink instability. The actual limit depends on several factors, including the stabilizing influence of the conducting components facing the plasma surface. At present, the wall elements are far from the plasma and the stability boundary is expected to be close to the “no wall” limit (no stabilizing wall effect). Investigation of maximum achievable βN values for the different operation scenario in ASDEX Upgrade is presented in this work. Two indicators are used to detect the stability boundary: Increase of the resonant field amplification (RFA) Onset of the ideal kink mode Recently installed internal active coils are used to probe stability of the plasma by the RFA technique. A wide range of different MHD diagnostics are used to identify the behaviour and structure of MHD modes in different discharges with high . Experimentally obtained results are compared with the results of the numerical modelling with linear MHD codes CASTOR-FLOW and MARS-K. Such comparison allows to validate the plasma model used in the codes and therefore to make numerical projection for further experimental studies

Overview of JET results

High density and high confinement operation in ELMy H-mode is confirmed at or above the normalized parameters foreseen for the ITER operating point (H98(y,2) 3c 1, n/nGW 3c 1, \u3b2N > 1.8 at q95 3c 3). The scaling of the ELMy H-mode with \u3b2N could be more favourable than that predicted by the IPB98(y,2) scaling. In ELMy H-mode, ion cyclotron current drive (ICCD) control of large sawteeth stabilized by fast particle has been demonstrated and the underlying neo-classical tearing modes (NTMs) and sawtooth physics is being refined. At high-density, Type I ELMy H-modes show trends that would lead to marginally acceptable ELMs on ITER. Type II ELM regime has been produced, though under very restrictive conditions. Type III ELMy operation with radiation fractions up to 95% has been demonstrated by seeding of N2 in H-modes and could extrapolate to Q = 10 ITER operation, albeit at high current (17 MA). The mitigation of Type I ELMs, nevertheless, remains a challenge. Considerable progress has been obtained in internal transport barrier (ITB) plasmas, with operation at central densities close to the Greenwald density or/and low toroidal rotation or/and high triangularity. Demonstrations of full current drive and successful simultaneous real time control of safety factor and temperature profiles have been achieved in ITB plasmas. Physics of resistive wall modes (RWMs) has been compared with theory, showing favourable scaling for ITER. High \u3b2N 3c 2.8 operation of hybrid modes (also called improved H-modes) has been obtained with dominant neutral beam heating. Hybrid modes with dominant ion cyclotron resonance heating (ICRH) have also been achieved. Trace tritium experiments yielded valuable information on particle transport in H-mode, ITB and hybrid regimes. In Type I ELMy plasmas, successful tests of the conjugate-T ICRH scheme have been achieved as well as lower hybrid coupling at ITER-relevant 10\u201311 cm distances. Reduced D and T fuel retention has been observed, which could relate to operation with vertical targets in the divertor and/or lower (ITER-like) vessel temperature. It is confirmed that erosion occurs predominantly on the main chamber surfaces, with possible benefits for T retention in ITER, although consequences for the metallic first wall lifetime need to be assessed. Disruption and ELM studies indicate that transient power deposition could be less constraining than expected for the ITER divertor, but more challenging for the metallic first wall. Alpha particle tomography and direct observation of alpha particle slowing down have been made possible by \u3b3 -spectroscopy. Measurements of Alfve \u301n cascades have been improved by a new interferometric technique. Promising tests of ITER relevant neutron counting detectors have been conducted

Real-time-capable prediction of temperature and density profiles in a tokamak using RAPTOR and a first-principle-based transport model

The RAPTOR code is a control-oriented core plasma profile simulator with various applications in control design and verification, discharge optimization and real-time plasma simulation. To date, RAPTOR was capable of simulating the evolution of poloidal flux and electron temperature using empirical transport models, and required the user to input assumptions on the other profiles and plasma parameters. We present an extension of the code to simulate the temperature evolution of both ions and electrons, as well as the particle density transport. A proof-of-principle neural-network emulation of the quasilinear gyrokinetic QuaLiKiz transport model is coupled to RAPTOR for the calculation of first-principle-based heat and particle turbulent transport. These extended capabilities are demonstrated in a simulation of a JET discharge. The multi-channel simulation requires ∼0.2 s to simulate 1 second of a JET plasma, corresponding to ∼20 energy confinement times, while predicting experimental profiles within the limits of the transport model. The transport model requires no external inputs except for the boundary condition at the top of the H-mode pedestal. This marks the first time that simultaneous, accurate predictions of Te, Tiand nehave been obtained using a first-principle-based transport code that can run in faster-than-real-time for present-day tokamaks

Comparison of runaway electron generation parameters in small, medium-sized and large tokamaks - A survey of experiments in COMPASS, TCV, ASDEX-Upgrade and JET

This paper presents a survey of the experiments on runaway electrons (RE) carried out recently in frames of EUROFusion Consortium in different tokamaks: COMPASS, ASDEX-Upgrade, TCV and JET. Massive gas injection (MGI) has been used in different scenarios for RE generation in small and medium-sized tokamaks to elaborate the most efficient and reliable ones for future RE experiments. New data on RE generated at disruptions in COMPASS and ASDEX-Upgrade was collected and added to the JET database. Different accessible parameters of disruptions, such as current quench rate, conversion rate of plasma current into runaways, etc have been analysed for each tokamak and compared to JET data. It was shown, that tokamaks with larger geometrical sizes provide the wider limits for spatial and temporal variation of plasma parameters during disruptions, thus extending the parameter space for RE generation. The second part of experiments was dedicated to study of RE generation in stationary discharges in COMPASS, TCV and JET. Injection of Ne/Ar have been used to mock-up the JET MGI runaway suppression experiments. Secondary RE avalanching was identified and quantified for the first time in the TCV tokamak in RE generating discharges after massive Ne injection. Simulations of the primary RE generation and secondary avalanching dynamics in stationary discharges has demonstrated that RE current fraction created via avalanching could achieve up to 70-75% of the total plasma current in TCV. Relaxations which are reminiscent the phenomena associated to the kinetic instability driven by RE have been detected in RE discharges in TCV. Macroscopic parameters of RE dominating discharges in TCV before and after onset of the instability fit well to the empirical instability criterion, which was established in the early tokamaks and examined by results of recent numerical simulations

Runaway electron beam control

Post-disruption runaway electron (RE) beams in tokamaks with large current can cause deep melting of the vessel and are one of the major concerns for ITER operations. Consequently, a considerable effort is provided by the scientific community in order to test RE mitigation strategies. We present an overview of the results obtained at FTU and TCV controlling the current and position of RE beams to improve safety and repeatability of mitigation studies such as massive gas (MGI) and shattered pellet injections (SPI). We show that the proposed RE beam controller (REB-C) implemented at FTU and TCV is effective and that current reduction of the beam can be performed via the central solenoid reducing the energy of REs, providing an alternative/parallel mitigation strategy to MGI/SPI. Experimental results show that, meanwhile deuterium pellets injected on a fully formed RE beam are ablated but do not improve RE energy dissipation rate, heavy metals injected by a laser blow off system on low-density flat-top discharges with a high level of RE seeding seem to induce disruptions expelling REs. Instabilities during the RE beam plateau phase have shown to enhance losses of REs, expelled from the beam core. Then, with the aim of triggering instabilities to increase RE losses, an oscillating loop voltage has been tested on RE beam plateau phase at TCV revealing, for the first time, what seems to be a full conversion from runaway to ohmic current. We finally report progresses in the design of control strategies at JET in view of the incoming SPI mitigation experiments