F. Köchl

ELM pacing and trigger investigations at JET with the new ITER Like Wall

During the installation of the new ITER-like wall (ILW) at JET, the high frequency pellet injector has been further improved. The launching system is now capable of delivering reliable fuelling size pellets from the magnetic outboard side up to 15xa0Hz repetition rate. Pacing size pellets can be produced at rates up to 50xa0Hz but pellet trains suffer some losses during the transfer to the plasma. A significant fraction of the pellet train can arrive at the plasma when launched from the outboard, while only a few pellets make it to the vessel inboard launching site. Stable and reliable edge-localized mode (ELM) control was achieved when using outboard fuelling size pellets. This tool was successfully applied for scenario development purposes in the ITER baseline H-mode scenario at 2.5xa0MA. Employed for ELM sustainment and impurity control, pellets prevented the ELM frequency from becoming so low as to cause a radiative collapse of the discharge. Despite technical limitations, injecting outboard pacing size pellets resulted in a transient enhancement of the initial ELM frequency up to a factor 4.5. This could be achieved in cases where a continuous train of sufficiently large and fast pellets were arriving in the plasma at a frequency of up to 31xa0Hz. Pacing size pellets were also used to investigate the ELM trigger threshold. Three basic parameters could be identified for outboard pellet launch. The ELM triggering probability increased with (i) the time elapsed since the previous ELM occurred, (ii) pellet mass and (iii) pellet speed. An indication for dependence of the ELM trigger threshold on the poloidal pellet launch position has been found; inboard launched pellets seem to reveal a higher trigger capability than pellets launched from the outboard. Finally, we compared the pellet penetration depth required for ELM triggering in the actual JET configuration with plasma-facing components to similar previous experiments performed with a carbon wall. This comparison indicates that pellet ELM triggering requires deeper penetration in the ILW configuration.

Tungsten impurity transport experiments in Alcator C-Mod to address high priority research and development for ITER

Experiments in Alcator C-Mod tokamak plasmas in the Enhanced D-alpha H-mode regime with ITER-like mid-radius plasma density peaking and Ion Cyclotron Resonant heating, in which tungsten is introduced by the laser blow-off technique, have demonstrated that accumulation of tungsten in the central region of the plasma does not take place in these conditions. The measurements obtained are consistent with anomalous transport dominating tungsten transport except in the central region of the plasma where tungsten transport is neoclassical, as previously observed in other devices with dominant neutral beam injection heating, such as JET and ASDEX Upgrade. In contrast to such results, however, the measured scale lengths for plasma temperature and density in the central region of these Alcator C-Mod plasmas, with density profiles relatively flat in the core region due to the lack of core fuelling, are favourable to prevent inter and intra sawtooth tungsten accumulation in this region under dominance of neoclassical...

Numerical analysis of JET discharges with the European Transport Simulator

The European Transport Simulator (ETS) (Coster et al 2010 IEEE Trans. Plasma Sci. 38 2085-92, Kalupin et al 2011 Proc. 38th EPS Conf. on Plasma Physics (Strasbourg, France, 2011) vol 35G (ECA) P. 4.111) is the new modular package for 1D discharge evolution developed within the EFDA Integrated Tokamak Modelling (ITM) Task Force. It consists of precompiled physics modules combined into a workflow through standardized input/output data structures. Ultimately, the ETS will allow for an entire discharge simulation from the start up until the current termination phase, including controllers and sub-systems. The paper presents the current status of the ETS towards this ultimate goal. It discusses the design of the workflow, the validation and verification of its components on the example of impurity solver and demonstrates a proof-of-principles coupling of a local gyrofluid model for turbulent transport to the ETS. It also presents the first results on the application of the ETS to JET tokamak discharges with the ITER like wall. It studies the correlations of the radiation from impurity to the choice of the sources and transport coefficients.

Optimizing the current ramp-up phase for the hybrid ITER scenario

The current ramp-up phase for the ITER hybrid scenario is analysed with the CRONOS integrated modelling suite. The simulations presented in this paper show that the heating systems available at ITER allow, within the operational limits, the attainment of a hybrid q profile at the end of the current ramp-up. A reference ramp-up scenario is reached by a combination of NBI, ECCD (UPL) and LHCD. A heating scheme with only NBI and ECCD can also reach the target q profile; however, LHCD can play a crucial role in reducing the flux consumption during the ramp-up phase. The optimum heating scheme depends on the chosen transport model, and on assumptions of parameters like ne peaking, edge Te,i and Zeff. The sensitivity of the current diffusion on parameters that are not easily controlled, shows that development of real-time control is important to reach the target q profile. A first step in that direction has been indicated in this paper. Minimizing resistive flux consumption and optimizing the q profile turn out to be conflicting requirements. A trade-off between these two requirements has to be made. In this paper it is shown that fast current ramp with L-mode current overshoot is at the one extreme, i.e. the optimum q profile at the cost of increased resistive flux consumption, whereas early H-mode transition is at the other extreme.

Modelling of JET hybrid scenarios with GLF23 transport model: E???B shear stabilization of anomalous transport

The E???B shear stabilization of anomalous transport in JET hybrid discharges is studied via self-consistent predictive modelling of electron and ion temperature, ion density and toroidal rotation velocity performed with the GLF23 model. The E???B shear stabilization factor (parameter ?E in the GLF23 model) is adjusted to predict accurately the four simulated quantities under different experimental conditions, and the uncertainty in ?E determined by 15% deviation between simulated and measured quantities is estimated. A correlation of ?E with toroidal rotation velocity and E???B shearing rate is found in the low density plasmas, suggesting that the turbulence quench rule may be more complicated than assumed in the GLF23 model with constant ?E. For the selected discharges the best predictive accuracy is obtained by using weak/no E???B shear stabilization (i.e. ?E???0) at low toroidal angular frequency (?? ?100?krad?s?1). Interestingly, a weak E???B shear stabilization of anomalous transport is found in the medium density strongly rotating discharge. An importance of linear ?e stabilization in this discharge is estimated and compared to the low density discharge with equally high ?e. The toroidal rotation velocity is well predicted here by assuming that the momentum diffusion coefficient is a fraction of thermal ion diffusivity. Taking into account the ?E and Prandtl number with their uncertainties determined in the modelling of JET hybrid discharges, the performance of ITER hybrid scenario with optimized heat mix (33?MW of NBI and 20?MW of ECCD) is estimated showing the importance of toroidal rotation for achieving Q?>?5.

Gyrokinetic projection of the divertor heat-flux width from present tokamaks to ITER

The XGC1 edge gyrokinetic code is used for a high fidelity prediction for the width of the heat-flux to divertor plates in attached plasma condition. The simulation results are validated against the empirical scaling

Recent results on the fuelling and control of plasmas by pellet injection, application to ITER

lambda_q propto B_P^{-gamma}

Analysis of fuelling requirements in ITER H-modes with SOLPS-EPED1 derived scalings

obtained from present tokamak devices, where

Integrated Core-SOL-Divertor Modelling for ITER Including Impurity: Effect of Tungsten on Fusion Performance in H-Mode and Hybrid Scenario

lambda_q

Understanding of the fundamental differences in JET and JT-60U AT discharges

is the divertor heat-flux width mapped to the outboard midplane and