The first rapid tokamak discharge shutdown using dispersive core payload deposition with shell pellets has been achieved in the DIII-D tokamak. Shell pellets are being investigated as a possible new ...path toward achieving tokamak disruption mitigation with both low conducted wall heat loads and slow current quench. Conventional disruption mitigation injects radiating impurities into the outer edge of the tokamak plasma, which tends to result in poor impurity assimilation and creates a strong edge cooling and outward heat flow, thus requiring undesirable high-Z impurities to achieve low conducted heat loads. The shell pellet technique aims to produce a hollow temperature profile by using a thin, low-ablation shell surrounding a dispersive payload, giving a greatly increased impurity ablation (and radiation) rate when the payload is released in the plasma core. This principle was demonstrated successfully using 3.6 mm outer diameter, 40 μm thickness diamond shells holding boron powder. The pellets caused rapid (<10 ms) discharge shutdown with low conducted divertor heat fluence (∼0.1 MJ/m^{2}). Confirmation of massive release of the boron powder payload into the plasma core was obtained spectroscopically. Some evidence for the formation of a hollow temperature profile during the shutdown was observed. These first results open a new avenue for disruption mitigation research, hopefully enabling development of highly effective methods of avoiding disruption wall damage in future reactor-scale tokamaks.
Divertor detachment offers a promising solution to the challenge of plasma-wall interactions for steady-state operation of fusion reactors. Here, we demonstrate the excellent compatibility of ...actively controlled full divertor detachment with a high-performance (β
~ 3, H
~ 1.5) core plasma, using high-β
(poloidal beta, β
> 2) scenario characterized by a sustained core internal transport barrier (ITB) and a modest edge transport barrier (ETB) in DIII-D tokamak. The high-β
high-confinement scenario facilitates divertor detachment which, in turn, promotes the development of an even stronger ITB at large radius with a weaker ETB. This self-organized synergy between ITB and ETB, leads to a net gain in energy confinement, in contrast to the net confinement loss caused by divertor detachment in standard H-modes. These results show the potential of integrating excellent core plasma performance with an efficient divertor solution, an essential step towards steady-state operation of reactor-grade plasmas.
Shattered pellet injection (SPI) is one of the prime candidates for the ITER disruption mitigation system because of its deeper penetration and larger particle flux than massive gas injection (MGI) ...(Taylor et al 1999 Phys. Plasmas 6 1872) using deuterium (Commaux et al 2010 Nucl. Fusion 50 112001, Combs et al 2010 IEEE Trans. Plasma Sci. 38 400, Baylor et al 2009 Nucl. Fusion 49 085013). The ITER disruption mitigation system will likely use mostly high Z species such as neon because of more effective thermal mitigation and pumping constraints on the maximum amount of deuterium or helium that could be injected. An upgrade of the SPI on DIII-D enables ITER relevant injection characteristics in terms of quantities and gas species. This upgraded SPI system was used on DIII-D for the first time in 2014 for a direct comparison with MGI using identical quantities of neon. This comparison enabled the measurements of density perturbations during the thermal quench (TQ) and radiated power and heat loads to the divertor. It showed that SPI using similar quantities of neon provided a faster and stronger density perturbation and neon assimilation, which resulted in a lower conducted energy to the divertor and a faster TQ onset. Radiated power data analysis shows that this was probably due to the much deeper penetration of the neon in the plasma inducing a higher core radiation than in the MGI case. This experiment shows also that the MHD activity during an SPI shutdown (especially during the TQ) is quite different compared to MGI. This favorable TQ energy dissipation was obtained while keeping the current quench (CQ) duration within acceptable limits when scaled to ITER.
A stochastic magnetic boundary, produced by an applied edge resonant magnetic perturbation, is used to suppress most large edge-localized modes (ELMs) in high confinement (H-mode) plasmas. The ...resulting H mode displays rapid, small oscillations with a bursty character modulated by a coherent 130 Hz envelope. The H mode transport barrier and core confinement are unaffected by the stochastic boundary, despite a threefold drop in the toroidal rotation. These results demonstrate that stochastic boundaries are compatible with H modes and may be attractive for ELM control in next-step fusion tokamaks.
We observe the formation of a high-pressure staircase pedestal (≈16–20 kPa) in the DIII-D tokamak when large amplitude edge localized modes are suppressed using resonant magnetic perturbations. The ...staircase pedestal is characterized by a flattening of the density and temperature profiles in midpedestal creating a two-step staircase pedestal structure correlated with the appearance of midpedestal broadband fluctuations. The pedestal oscillates between the staircase and single-step structure every 40–60 ms, correlated with oscillations in the heat and particle flux to the divertor. Gyrokinetic analysis using the cgyro code shows that when the heat and particle flux to the divertor decreases, the pedestal broadens and the E×B shear at the midpedestal decreases, triggering a transport bifurcation from the kinetic ballooning mode (KBM) to trapped electron mode (TEM) limited transport that flattens the density and temperature profiles at midpedestal and results in the formation of the staircase pedestal. As the heat flux to the divertor increases, the pedestal narrows and the E×B shear at the midpedestal increases, triggering a back transition from TEM to KBM limited transport. The pedestal pressure increases during the staircase phase, indicating that enhanced midpedestal turbulence can be beneficial for confinement.
Abstract
Type-I and type-II edge-localized-modes (ELMs) heat flux profiles measured at the DIII-D divertor feature a peak in the vicinity of the strike-point and a plateau in the scrape-off-layer ...(SOL), which extends to the first wall. The plateau is present in attached and detached divertors and it is found to originate with plasma bursts upstream in the SOL. The integrated ELM heat flux is distributed at ∼65% in the peak and ∼35% in this plateau. The parallel loss model, currently used at ITER to predict power loads to the walls, is benchmarked using these results in the primary and secondary divertors with unprecedented constraints using experimental input data for ELM size, radial velocity, energy, electron temperature and density, heat flux footprints and number of filaments. The model can reproduce the experimental near-SOL peak within ∼20%, but cannot match the SOL plateau. Employing a two-component approach for the ELM radial velocity, as guided by intermittent data, the full radial heat flux profile can be well matched. The ELM-averaged radial velocity at the separatrix, which explains profile widening, increases from ∼0.2 km s
−1
in attached to ∼0.8 km s
−1
in detached scenarios, as the ELM filaments’ path becomes electrically disconnected from the sheath at the target. The results presented here indicate filaments fragmentation as a possible mechanism for ELM transport to the far-SOL and provide evidence on the beneficial role of detachment to mitigate ELM flux in the divertor far-SOL. However, these findings imply that wall regions far from the strike points in future machines should be designed to withstand significant heat flux, even for small-ELM regimes.
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