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25 records · Page 2

Dependence of the boundary heat flux width on core and edge profiles in Alcator C-Mod

This work presents new evidence that the heat flux width, λ q , in the Alcator C-Mod tokamak scales with the edge electron pressure, as observed in the ASDEX Upgrade (AUG) tokamak, but the scaling with volume-averaged pressure, $\bar{p}$, from the plasma stored energy, is a better predictor of λ q in Alcator C-Mod than the edge electron pressure. These previous studies, which find that λ q decreases with increasing plasma pressure, imply that a high performance core at high pressure will lead to challenging heat and particle exhaust due to very small λ q . This concern has led to our significant enlargement of the C-Mod database with the electron density, temperature, and pressure profile data from the Thomson scattering and electron cyclotron emission diagnostics. Using the C-Mod database augmented with new profile data, we find that λ q decreases with increasing edge electron pressure as ${\lambda }_{q}\propto {p}_{\mathrm{e},95}^{-0.26}$, similar to results from AUG, and showing the strength of cross-machine comparisons. We also find that ${\lambda }_{q}\propto {p}_{\mathrm{e},\mathrm{c}\mathrm{o}\mathrm{r}\mathrm{e}}^{-0.56}$, consistent with the original finding from C-Mod that the heat flux width scales as ${\bar{p}}^{-0.48}$. The scalings of λ q with separatrix pressure and gradient scale length are found to match the AUG results qualitatively. The C-Mod scalings with edge plasma quantities have more scatter than the $\bar{p}$ scaling, and, importantly, show different trends for H-modes relative to L- and I-mode. Furthermore, investigating the source of this discrepancy presents an opportunity for further study that may improve our ability to predict the heat flux width in different confinement scenarios in the pursuit of optimizing core-edge performance in future reactors.

Alcator C-Mod↗

Validation of EDGDE2D-EIRENE predicted 2D distributions of electron temperature and density against divertor Thomson scattering measurements in the low-field side divertor leg in DIII-D

Comparisons of profiles of the electron temperature (T e ), density (n e ) and pressure (p e ) measured with Divertor Thomson Scattering in DIII-D low-confinement mode discharges to predictions from the edge fluid code EDGE2D-EIRENE show that the models implemented in EDGE2D-EIRENE predict the measurements within their collective uncertainties if the T e at the separatrix (T e,sep ) is 10 eV, or higher. The simulations do not predict, however, the peaked T e and ne profiles measured adjacent to the target plate when T e,sep , is below 10 eV, i.e., for the plasma downstream from the region of ionization of deuterium atoms. Inclusion of cross-field drifts and a five-fold reduction of radial transport cannot reconcile the discrepancy between the measurements and predictions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Evaluation of SPARC divertor conditions in H-mode operation using SOLPS-ITER

The predicted divertor conditions for the SPARC tokamak are calculated using SOLPS-ITER for a range of scrape-off-layer (SOL) heat flux widths λ q , input powers, and particle fueling locations. Under H-mode scenario conditions with an upstream separatrix density of 1 x 10 20 m -3 , the most conservative range of λ q extrapolations ( 0.15 mm) results in extremely high unmitigated particle and energy fluxes to the divertor, both under full field (12.2 T) and power (P SOL = 29 MW) conditions, and 2/3 field with P SOL = 10 MW. Increasing the cross-field SOL diffusivities by 2–10× reduces the magnitude of the mitigation challenge, however strategies such as impurity seeding or strike-point-sweeping will likely still be required. A combination of steady-state and time-dependent SOLPS-ITER simulations are used to map out phase space diagrams of upstream and divertor conditions. The simulations include parallel currents but neglect cross-field drifts. At low upstream density the inner and outer divertor conditions are highly asymmetric, with a large temperature difference and significant heat fluxes driven by parallel currents. The solution has sharp bifurcations with a region of hysteresis, depending on whether the initial state is at a low or high density. This behavior is observed even when the fueling location, cross-field diffusivity, and impurity level is changed, although the density window with asymmetry is reduced with increasing diffusivity. The addition of neon impurity seeding reduces the divertor heat fluxes, but also causes a drop in the upstream electron density with fixed particle throughput. This drop can be counteracted by increased main ion throughput, however too much neon results in a back transition into the asymmetric divertor regimes suggesting a need for control of both main ion and impurity seeding levels to achieve a desired divertor state.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

A reduced-model (nSOLT) simulation of neutral recycling effects on plasma turbulence in the divertor region of MAST-U

The 2D scrape-off-layer turbulence code (nSOLT), which includes 1D kinetic neutral–plasma interactions, is applied to study effects of neutral recycling on plasma turbulence for parameters illustrative of the MAST-U divertor region. Neutral recycling is modeled by injecting a fraction of the parallel plasma flux to the divertor back into the simulation domain as a source of Franck–Condon-distributed neutrals. Stationary sources, concentrated at the magnetic separatrix, model plasma streaming into the divertor region from the upstream scrape-off-layer and sustain plasma turbulence absent neutral recycling. Starting from one such no-neutrals equilibrium, we initiate recycling in a numerical experiment designed to diagnose and identify the effects of various neutral–plasma interactions on the divertor plasma, divertor turbulence, and plasma exhaust. The onset of recycling triggers an initial burst of enhanced cross field plasma transport that is quelled by ionization cooling and charge–exchange (CX) friction, with growing neutral pressure, leading to a quiescent, turbulence-free state. Diagnosis of this transient burst reveals that (1) the sudden increase in plasma density due to ionization dominates the onset of the burst, (2) electron cooling due to ionization increases collisionality and disconnects blob filaments from the sheath, and (3) CX friction drives tripole polarization of a blob that can dominate the curvature-driven dipole polarization, leading to the stagnation of blob propagation and reduced radial turbulent transport. It is shown that CX friction is negligible compared to sheath physics in determining equilibrium mean flow shearing rates, for parameters considered herein (specifically a short connection length to the divertor target), while it can significantly reduce interchange-instability growth rates.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

High performance power handling in the absence of an H-mode edge in negative triangularity DIII-D plasmas

Abstract Experiments performed during strongly-shaped high-power diverted negative triangularity (NT) experiments in DIII-D achieved detached divertor conditions and a transient-free edge, showcasing the potential for application of NT to a core-edge integrated reactor-like scenario and providing the first characterization of the parametric dependencies for detachment onset. Detached divertor conditions will be required in future devices to mitigate divertor heat fluxes. Access to dissipative divertor conditions was investigated via an increase in upstream density. Detachment onset at the outer strike point was achieved with H-mode level confinement H 98 − y 2 ∼ 1 and reactor-relevant normalized pressures β N ∼ 2 . Confinement degradation was observed with deeper detachment, associated with the loss of an electron temperature pedestal. Differences in geometry, radial transport, impact of cross field drifts are discussed to explain differences in access to detachment in NT discharges. Higher normalized densities, with respect to equivalent discharges in positive triangularity, were necessary to achieve detachment, partially explained by the shorter parallel connection length to the targets. The effect of cross-field particle drifts ( E × B , B × ∇ B ) on access to detachment was demonstrated by the lower upstream density needed to access detachment with ion B × ∇ B drift directed outside of the active divertor (Greenwald fraction f Gw ∼ 0.9–1.0 vs f Gw ∼ 1.3). The upstream density at detachment onset was observed to increase linearly with plasma current with ion B × ∇ B drift into the divertor, consistent with the observed narrowing of the scrape-off layer heat flux width λ q . Edge fluid simulations capture separatrix densities needed to achieve detachment in NT plasma and their dependence on drift direction. The ability to reproduce detachment dynamics in NT plasma increases the confidence in future design studies for NT divertors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

UEDGE modeling of plasma detachment of CFETR with ITER ‐like divertor geometry by external impurity seeding

Abstract Efficient handling of high heat flux on the plasma‐facing components, particularly the divertor targets, poses a significant challenge for the Chinese Fusion Engineering Testing Reactor (CFETR) with fusion power of Gigawatt. This work investigates the divertor plasma detachment of CFETR with a standard ITER‐like divertor geometry by neon (Ne) or argon (Ar) impurity seeding using UEDGE code. The cross‐field drifts terms are switched off, and fluid neutral models and a “fixed‐fraction” impurity model are applied to enable efficient simulations for the study of CFETR detachment. In order to reduce the heat load on the divertor targets below the acceptable level (<10 MW/m 2 ), the impurity fraction ( f ), pumping speed ( S ), and upstream density are varied to identify the suitable operations window during Ne seeding. The effects of Ne and Ar impurities on the plasma detachment are compared. It is found that with the power across the core‐edge interface P SOL = 200 MW and separatrix density of 2.8 10 19 , Ne impurity fraction ≥1.7%, and Ar impurity fraction ≥0.24% can achieve the partial detachment. Achieving similar total radiation power (˜148 MW), the Ne fraction is 2.3% and the Ar fraction is 0.24%. Moreover, the simulation results indicate that Ar exhibits better power radiation efficiency and core compatibility compared with Ne.

Zhang, M. Z.↗

The role of B T -dependent flows on W accumulation at the edge of the confined plasma

Abstract Near-separatrix impurity accumulation between the crown and the outer midplane of tokamaks is a common feature in results from codes such as SOLPS-ITER and DIVIMP; however, experimental evidence of accumulation has only recently been obtained and is reported here. The codes find that the poloidal distribution of impurity ions in the scrape-off layer (SOL) depends primarily on toroidal field ( B T )-dependent parallel flow patterns of the background plasma and the parallel ion temperature gradient (∇ ‖ T ion ) force. Experimentally, Mach probes used in L-mode plasmas with favorable (for H-mode access) B T measure fast ( M ∼ 0.3–0.5) inner-target-directed (ITD) background plasma flows at the crown of single-null discharges. This study reports a set of DIVIMP simulations for two similar H-mode discharges from the DIII-D W metal rings campaign differing primarily in B T -direction to assess the effect that fast ITD flows have on the distribution of W ions in the SOL. It is found that for imposed ITD flows of M = 0.3, W ions that otherwise accumulate due to the ∇ ‖ T ion -force are largely flushed out. It is also found that doubling the radial diffusion coefficient from 0.3 to 0.6 m 2 s −1 prevents accumulation due to rapid cross-field transport into the far-SOL, where background plasma flows drain W ions to the divertors. Far-SOL W distributions from DIVIMP are then used to specify input to the impurity transport code 3DLIM, which is used to interpretively model collector probe (CP) deposition patterns measured in the ‘wall-SOL’. It is demonstrated that the deposition patterns are consistent with the DIVIMP predictions of near-SOL accumulation for the unfavorable- B T direction, and little/no accumulation for the favorable- B T direction. The wall-SOL CPs have thus provided the first experimental evidence, albeit indirect, of near-SOL W accumulation—finding it occurs for the unfavorable- B T direction only. For the favorable- B T direction, fast flows can largely prevent accumulation from occurring.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗