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Register a new userL-Mode and Inter-ELM Divertor Particle and Heat Flux Width Scaling on MAST
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The distribution of particles and power to plasma-facing components is of key importance in the design of next-generation fusion devices. Power and particle decay lengths have been measured in a number of MAST L-mode and H-mode discharges in order to determine their parametric dependencies, by fitting power and particle flux profiles measured by divertor Langmuir probes, to a convolution of an exponential decay and a Gaussian function. In all discharges analysed, it is found that exponential decay lengths mapped to the midplane are mostly dependent on separatrix electron density and plasma current (or parallel connection length). The widths of the convolved Gaussian functions have been used to derive an approximate diffusion coefficient, which is found to vary from 1m2/s to 7m2/s, and is systematically lower in H-mode compared with L-mode.
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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 $lambda_q propto B_P^{-gamma}$ obtained from present tokamak devices, where $lambda_q$ is the divertor heat-flux width mapped to the outboard midplane and $gamma_q=1.19$ as defined by T. Eich et al. [Nucl. Fusion 53 (2013) 093031], and $B_P$ is the magnitude of the poloidal magnetic field at outboard midplane separatrix surface. This empirical scaling predicts $lambda_q leq 1mm$ when extrapolated to ITER, which would require operation with very high separatrix densities $(n_{sep}/n_{Greenwald} > 0.6)$ in the Q=10 scenario to achieve semi-detached plasma operation and high radiative fractions leading to acceptable divertor power fluxes. XGC1 predicts, however, that $lambda_q$ for ITER is over 5 mm, suggesting that operation in the ITER Q=10 scenario with acceptable divertor power loads could be obtained over a wider range of plasma separatrix densities and radiative fractions. The physics reason behind this difference is, according to the XGC1 results, that while the ion magnetic drift contribution to the divertor heat-flux width is wider in the present tokamaks, the turbulent electron contribution is wider in ITER. A high current C-Mod discharge is found to be in a mixed regime: While the heat-flux width by the ion neoclassical magnetic drift is still wider than the turbulent electron heat-flux width, the heat-flux magnitude is dominated by the narrower electron heat-flux.
Sustained ELM mitigation has been achieved using RMPs with a toroidal mode number of n=4 and n=6 in lower single null and with n=3 in connected double null plasmas on MAST. The ELM frequency increases by up to a factor of eight with a similar reduction in ELM energy loss. A threshold current for ELM mitigation is observed above which the ELM frequency increases approximately linearly with current in the coils. A comparison of the filament structures observed during the ELMs in the natural and mitigated stages shows that the mitigated ELMs have the characteristics of type I ELMs even though their frequency is higher, their energy loss is reduced and the pedestal pressure gradient is decreased. During the ELM mitigated stage clear lobe structures are observed in visible-light imaging of the X-point region. The size of these lobes is correlated with the increase in ELM frequency observed. The RMPs produce a clear 3D distortion to the plasma and it is likely that these distortions explain why ELMs are destabilised and hence why ELM mitigation occurs.
The application of resonant magnetic perturbations (RMPs) produces splitting of the divertor strike point due to the interaction of the RMP field and the plasma field. The application of a rotating RMP field causes the strike point splitting to rotate, distributing the particle and heat flux evenly over the divertor. The RMP coils in MAST have been used to generate a rotating perturbation with a toroidal mode number n=3. The ELM frequency is doubled with the application of the RMP rotating field, whilst maintaining the H mode. During mitigation, the ELM peak heat flux is seen to be reduced by 50% for a halving in the ELM energy and motion of the strike point, consistent with the rotation of the applied RMP field, is seen using high spatial resolution (1.5mm at the target) heat flux profiles measured using infrared (IR) thermography.
Understanding and predicting divertor heat-load width ${lambda}_q$ is a critically important problem for an easier and more robust operation of ITER with high fusion gain. Previous predictive simulation data for ${lambda}_q$ using the extreme-scale edge gyrokinetic code XGC1 in the electrostatic limit under attached divertor plasma conditions in three major US tokamaks [C.S. Chang et al., Nucl. Fusion 57, 116023 (2017)] reproduced the Eich and Goldston attached-divertor formula results [formula #14 in T. Eich et al., Nucl. Fusion 53, 093031 (2013); R.J. Goldston, Nucl. Fusion 52, 013009 (2012)], and furthermore predicted over six times wider ${lambda}_q$ than the maximal Eich and Goldston formula predictions on a full-power (Q = 10) scenario ITER plasma. After adding data from further predictive simulations on a highest current JET and highest-current Alcator C-Mod, a machine learning program is used to identify a new scaling formula for ${lambda}_q$ as a simple modification to the Eich formula #14, which reproduces the Eich scaling formula for the present tokamaks and which embraces the wide ${lambda}_q^X{GC}$ for the full-current Q = 10 ITER plasma. The new formula is then successfully tested on three more ITER plasmas: two corresponding to long burning scenarios with Q = 5 and one at low plasma current to be explored in the initial phases of ITER operation. The new physics that gives rise to the wider ${lambda}q_^{XGC} is identified to be the weakly-collisional, trapped-electron-mode turbulence across the magnetic separatrix, which is known to be an efficient transporter of the electron heat and mass. Electromagnetic turbulence and high-collisionality effects on the new formula are the next study topics for XGC1.
A potentially important feature in a divertor design for a high-power tokamak is an extended and expanded divertor leg. The upgrade to MAST will allow a wide range of such divertor leg geometries to be produced, and hence will allow the roles of greatly increased connection length and flux expansion to be experimentally tested. This will include testing the potential of the Super-X configuration [1]. The design process for the upgrade has required analysis of producing and controlling the magnetic configurations, and has included consideration of the roles that divertor closure and increasing magnetic connection length will play.
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