17 Search Results

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 themore » 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.« less

Abstract Extremely intense power exhaust channels are projected for tokamak-based fusion power reactors; a means to handle them remains to be demonstrated. Advanced divertor configurations have been proposed as potential solutions. Recent modelling of tightly baffled, long-legged divertor geometries for the divertor test tokamak concept, ADX, has shown that these concepts may access passively stable, fully detached regimes over a broad range of parameters. The question remains as to how such divertors may perform in a reactor setting. To explore this, numerical simulations are performed with UEDGE for the long-legged divertor geometry proposed for the ARC pilot plant conceptual design—amore » device with projected heat flux power width ( ) of 0.4 mm and power exhaust of 93 MW—first for a simplified Super-X divertor configuration (SXD) and then for the actual X-point target divertor (XPTD) being proposed. It is found that the SXD, combined with 0.5% fixed-fraction neon impurity concentration, can produce passively stable, detached divertor regimes for power exhausts in the range of 80–108 MW—fully accommodating ARC’s power exhaust. The XPTD configuration is found to reduce the strike-point temperature by a factor of  ∼10 compared to the SXD for small separations (∼1.4 ) between main and divertor X-point magnetic flux surfaces. Even greater potential reductions are identified for reducing separations to  ∼1 or less. The power handling response is found to be insensitive to the level of cross-field convective or diffusive transport assumed in the divertor leg. By raising the separatrix density by a factor of 1.5, stable fully detached divertor solutions are obtained that fully accommodate the ARC exhaust power without impurity seeding. To our knowledge, this is the first time an impurity-free divertor power handling scenario has been obtained in edge modelling for a tokamak fusion power reactor with of 0.4 mm.« less

Numerical modeling of divertor configurations with radially or vertically extended, tightly baffled, outer divertor legs has demonstrated the existence of a passively-stable fully detached divertor regime. Here in the simulations, long-legged divertors provide up to an order-of-magnitude increase in peak power handling capability compared to conventional divertors. The key physics for attaining the passively stable, fully detached regime in these simulations involves the interplay of strong convective plasma transport to the divertor leg outer sidewall, confinement of neutral gas in the divertor volume, geometric effects including a secondary X-point, and atomic radiation. New analysis shows that in this regime themore » detachment front location is set by the balance between the power entering the divertor leg and the losses to the walls of the divertor channel. Correspondingly, the maximum power that can be accommodated by the divertor, while still staying detached, increases with the poloidal length of the leg. The detached regime access window in terms of input power, density and impurity seeding concentration varies quantitatively depending on divertor geometry and modeling assumptions most specifically, cross-field transport to the side walls.« less

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Understanding the statistics of fluctuation driven flows in the boundary layer of magnetically confined plasmas is desired to accurately model the lifetime of the vacuum vessel components. Mirror Langmuir probes (MLPs) are a novel diagnostic that uniquely allow us to sample the plasma parameters on a time scale shorter than the characteristic time scale of their fluctuations. Sudden large-amplitude fluctuations in the plasma degrade the precision and accuracy of the plasma parameters reported by MLPs for cases in which the probe bias range is of insufficient amplitude. While some data samples can readily be classified as valid and invalid, wemore » find that such a classification may be ambiguous for up to 40% of data sampled for the plasma parameters and bias voltages considered in this study. In this contribution to this work, we employ an autoencoder (AE) to learn a low-dimensional representation of valid data samples. By definition, the coordinates in this space are the features that mostly characterize valid data. Ambiguous data samples are classified in this space using standard classifiers for vectorial data. In this way, we avoid defining complicated threshold rules to identify outliers, which require strong assumptions and introduce biases in the analysis. By removing the outliers that are identified in the latent low-dimensional space of the AE, we find that the average conductive and convective radial heat fluxes are between approximately 5% and 15% lower as when removing outliers identified by threshold values. For contributions to the radial heat flux due to triple correlations, the difference is up to 40%« less

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We report management of power exhaust will be a crucial task for tokamak fusion reactors. Reactor concepts are often proposed with double-null divertors, i.e. having two magnetic separatrices in an up-down symmetric configuration. This arrangement is potentially advantageous since the majority of the tokamak exhaust power tends to flow to the outer pair of divertor legs at large major radius, where the geometry is favorable for spreading the heat over a large surface area and there is more room for advanced divertor configurations. Despite the importance, there have been relatively few studies of divertor power sharing in near double nullmore » configurations and no studies at the poloidal magnetic fields and scrape-off layer power widths anticipated for a reactor. Motivated by this need we have undertaken a systematic study on Alcator C-Mod, examining the effect of magnetic flux balance on the power sharing among the four divertor legs in near double-null plasmas. Ohmic L-modes at three values of plasma current and ICRF-heated enhanced D-alpha (EDA) H-modes and I-modes at a single value of plasma current are explored, producing poloidal magnetic fields of 0.42, 0.62 and 0.85 Tesla. For Ohmic L-modes and ICRF-heated EDA H-modes, we find that the point of equal power sharing between upper and lower divertors occurs remarkably close to a balanced double null. Power sharing amongst the outer (upper versus lower) and inner (upper versus lower) pairs of divertors can be described in terms of a logistic function of magnetic flux balance, consistent with heat flux mapping along magnetic field lines to the outer midplane. Power sharing between inner and outer legs is found to follow a Gaussian-like function of magnetic flux balance with non-zero power to the inner divertors at double null. The overall behavior of H-modes operated near double null and for I-modes operating to within one heat flux e-folding of double null are found similar to Ohmic L-modes, with a significant reduction of power on the inner divertor legs. The results are encapsulated in terms of empirically-informed analytic functions of magnetic flux balance. When combined with magnetic equilibrium control system specifications, these relationships can be used to specify the power flux handling requirements for each of the four divertor target plates.« less

In order to more completely demonstrate the I-mode regime as a compelling fusion reactor operating scenario, the first dedicated attempts at I-mode radiative heat exhaust and detachment were carried out on Alcator C-Mod. Results conclusively show that within the parameter space explored, an I/L back-transition is triggered prior to meaningful reductions in parallel heat flux, q_║, target temperature, T_e,tar and target pressure, p_e,tar at the outer divertor. The exact mechanism for the I/L trigger remains uncertain, but a multi-diagnostic investigation suggests the pedestal regulation physics is impacted promptly by small amounts of N₂ seeded into the private flux region. Furthermore,more » the time delay between when N₂ contacts the plasma and the I/L transition is triggered varied from 30–120 ms, approximately 0.7-3×τ_E, and the delay varied inversely with I-mode pedestal-top pressure, p_e,95. Power and nitrogen influx scans indicate that the I/L transitions are not linked to excessive bulk-plasma impurity radiation. Additionally, it is also shown that in the subsequent L-mode following nitrogen seeding, q_║ and T_e,tar can be reduced by factors of ~10. The I/L transition and L-mode exhaust results using N₂ are compared to similar attempts using Ne where such q_║ and T_e,tar reductions in L-mode are limited to factors of 2-3. Implications for the I-mode regime are discussed, including needs for follow-up experiments on other facilities.« less

The Shoelace antenna was built to drive edge fluctuations in the Alcator C-Mod tokamak, matching the wavenumber ($$k_{\perp}\approx1.5$$ cm^-1) and frequency ($$30\lesssim f \lesssim 200$$ kHz) of the quasi-coherent mode (QCM), which is responsible for regulating transport across the plasma boundary in the steady-state, ELM-free Enhanced Dα (EDA) H-mode. Initial experiments in 2012 demonstrated that the antenna drove a resonant response in the edge plasma in steady-state EDA and transient, non-ELMy H-modes, but transport measurements were unavailable. In 2016, the Shoelace antenna was relocated to enable direct measurements of driven transport by a reciprocating Mirror Langmuir Probe, while also makingmore » available gas puff imaging and reflectometer data to provide additional radial localization of the driven fluctuation. This new data suggests a ~4 mm-wide mode layer centered on or just outside the separatrix. Fluctuations coherent with the antenna produced a radial electron flux with $${\Gamma_e}/{n_e}\sim4$$ m s^-1 in EDA H-mode, smaller than but comparable to the QCM level. But in transient ELM-free H-mode, $${\Gamma_e}/{n_e}$$ was an order of magnitude smaller, and driven fluctuations reduced by a factor of $$\gtrsim$$ 3. The driven mode is quantitatively similar to the intrinsic QCM across measured spectral quantities, except that it is more coherent and weaker. This work informs the prospect of achieving control of edge transport by direct coupling to edge modes, as well as the use of such active coupling for diagnostic purposes.« less