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Experiments indicate that the formation of density shoulders in the far scrape-off layer (SOL) is associated with an increase in cross-field transport and the propagation of coherent structures of high density known as blobs. It has been suggested that far SOL divertor collisionality (Λdiv) may be a controlling parameter in this physics, which could be challenging for heat flux mitigation schemes. At Alcator C-Mod, a new high spatial resolution divertor Langmuir probe array (rail probes) combined with a scanning Mirror Langmuir probe has been used to investigate this phenomenon in detail. Experiments were performed to mitigate divertor heat flux and thus increase Λdiv by two different methods and observe the divertor and scrape-off layer plasma response: (1) a 3-point density scan is found to increase the fluctuation intermittency at both the midplane and divertor, and produce a density shoulder at elevated Greenwald fraction; (2) a 3-point N2 seeding scan is found to increase the divertor strike point collisionality, by over two orders of magnitude yet no noticeable trend is observed in either the density profile or the fluctuation statistics. It is concluded that Λdiv near the strike point is not by itself a controlling parameter for SOL fluctuations and shoulder formation which is important for localized heat flux mitigation schemes targeting the strike point where heat flux is the greatest.

The Alcator C-Mod mirror Langmuir probe diagnostic is used to characterize plasma fluctuations in the outboard mid-plane scrape-off layer (SOL). Ohmically heated L-mode plasmas with Greenwald fractions from 0.1 to 0.45 are analyzed. In plasmas with a low line-averaged core density, the radial profiles of the average electron density and temperature feature a two-scale structure with a strong gradient in the near-SOL and a weak gradient in the far-SOL. The relative fluctuation levels of the electron density are below 25% throughout the entire SOL and the fluctuation driven heat flux is dominated by plasma density fluctuations. In the plasma with the highest line-averaged core density, the radial profiles of the electron density and temperature feature only a single weak gradient length scale across the SOL. The plasma density at the limiter radius is approximately 10 times larger than in low-density discharges. The average electron temperature at the separatrix falls to approximately 30 eV while the magnitude of the relative temperature fluctuations increases with the line-averaged core plasma density. With increasing plasma density, the fluctuation-driven heat fluxes increase by more than one order of magnitude while retaining a relative fluctuation level of approximately 3. Even though the SOL is on average cooler than in low density discharges, large electron temperature fluctuations govern the turbulence driven heat flux towards the wall while contributions from density fluctuations and triple correlations contribute approximately 35% and 15%, respectively.

A poloidal array of toroidally-extended, flush-mounted ‘rail’ Langmuir probes was recently installed on Alcator C-Mod's vertical target plate divertor. The aim was to investigate if a Langmuir probe array could be designed to survive reactor-level heat fluxes and have the ability to make measurements that could be reliably interpreted under reactor-level plasma densities, neutral densities and magnetic fields. Langmuir probes are typically built to have incident field-line angles >10° to avoid interpretation issues associated with sheath expansion. However, at the high parallel heat fluxes experienced in reactor-relevant conditions such a probe would quickly overheat and melt. To mitigate both the issues of extreme heat flux and sheath expansion, each probe was designed to be flush with the divertor surface, toroidally-extended and field-aligned, giving it a ‘rail’ geometry. The flush mounted probes have proven to be exceptionally robust surviving the 2015–2016 campaign – a first for a C-Mod probe system. Examination of the probe current-voltage (I-V) characteristics reveals that they are immune to sheath expansion at incident field angles down to ∼0.5°. Comparison of the flush probes to traditional proud probes shows that both measure the same electron pressure across the divertor plate. However, there are significant and systematic differences in the density, temperature and floating potential. This suggests that there is important physics, perhaps unique to conditions in a vertical-target plate divertor with small field-line attack angles, that affects the I-V characteristics and is not currently included in probe data analyses. Finally, the probe response is examined in the ‘death-ray’ regime, just near detachment. Previous work using proud probes has suggested that the ‘death-ray’ is an artefact of the probe bias. However, on flush mounted probes the ‘death-ray’ manifests itself under different conditions, which may provide insight into this unique phenomenon.

New experiments on Alcator C-Mod reveal that the favorable impurity screening characteristics of the high-field side (HFS) scrape-off layer (SOL), previously reported for single null geometries, is retained in double null configurations, despite the formation of an extremely thin SOL. In balanced double-null, nitrogen injected locally into the HFS SOL is better screened by a factor of 2.5 compared to the same injection into the low field side (LFS) SOL. This result is insensitive to plasma current and Greenwald fraction. Nitrogen injected into the HFS SOL is not as well screened (only a factor of 1.5 improvement over LFS) in unbalanced double-null discharges, when the primary divertor is in the direction of B×∇B. In this configuration, impurity ‘plume’ emission patterns indicate that an opposing E×B drift competes with the parallel impurity flow to the divertor. In balanced double-null plasmas, the dispersal pattern exhibits a dominant E×B motion. Unbalanced discharges with the primary divertor opposite the direction of B×∇B exhibit excellent HFS screening characteristics – a factor of 5 enhancement compared to LFS. These data support the idea that future tokamaks should locate all RF actuators and close-fitting wall structures on the HFS and employ near-double-null magnetic topologies, both to precisely control plasma conditions at the antenna/plasma interface and to maximally mitigate the impact of local impurity sources arising from plasma-material interactions. Keywords: Alcator C-Mod, Impurity screening, Double null, High field side scrape-off layer

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. 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 the 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. [ABSTRACT FROM AUTHOR]

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—a 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. [ABSTRACT FROM AUTHOR]

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, , target temperature, and target pressure, 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. The time delay between when N2 contacts the plasma and the I/L transition is triggered varied from 30–120 ms, approximately 0.7–, 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. It is also shown that in the subsequent L-mode following nitrogen seeding, and 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 and 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. [ABSTRACT FROM AUTHOR]

New data from Alcator C-Mod have extended the range of heat flux measurements and scalings to poloidal magnetic fields above ITER-level. Knowledge of how the scrape-off layer heat flux width (λq) scales with machine parameters is crucial for designing fusion reactors and developing a power exhaust solution. An international database indicated that λq scaled approximately inversely with the poloidal magnetic field (Bp) and had no other significant dependencies. However, reactor-class tokamaks are expected to have at least 50% higher Bp than the maximum of that database (0.8 T). Alcator C-Mod has been the only diverted tokamak capable of operating at reactor-level Bp, up to ~1.3 T. A major focus of the final experimental campaign on Alcator C-Mod was to characterize λq over a wide range of conditions, utilizing a unique array of heat flux sensors with unprecedented spatial resolution and heat flux dynamic range. The heat flux width scaling is found to extend up to Bp ~ 1.3 T in H-mode. Looking across confinement regimes we find the remarkable result that λq exhibits a unified dependence on volume-averaged core plasma pressure (). Within a standard deviation of about 20%, the heat flux width in any of the C-Mod plasmas studied (L-, I-, and H-mode) is proportional to the inverse square root of . It is also found that the standard prescription of representing the target plate heat flux profile as a convolution of exponential and Gaussian functions does not capture the heat flux profile measured in the private zone; a purely exponential decay fits the data better in this region to  >3 orders of magnitude in heat flux dynamic range. [ABSTRACT FROM AUTHOR]

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 null 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. [ABSTRACT FROM AUTHOR]

Halo currents generated during disruptions on Alcator C-Mod have been measured with Langmuir ‘rail’ probes. These rail probes are embedded in a lower outboard divertor module in a closely-spaced vertical (poloidal) array. The dense array provides detailed resolution of the spatial dependence (~1 cm spacing) of the halo current distribution in the plasma scrape-off region with high time resolution (400 kHz digitization rate). As the plasma limits on the outboard divertor plate, the contact point is clearly discernible in the halo current data (as an inversion of current) and moves vertically down the divertor plate on many disruptions. These data are consistent with filament reconstructions of the plasma boundary, from which the edge safety factor of the disrupting plasma can be calculated. Additionally, the halo current ‘footprint’ on the divertor plate is obtained and related to the halo flux width. The voltage driving halo current and the effective resistance of the plasma region through which the halo current flows to reach the probes are also investigated. Estimations of the sheath resistance and halo region resistivity and temperature are given. This information could prove useful for modeling halo current dynamics. [ABSTRACT FROM AUTHOR]

The Alcator C-Mod team has recently developed a feedback system to measure and control surface heat flux in real-time. The system uses real-time measurements of surface heat flux from surface thermocouples and a pulse-width modulated piezo valve to inject low-Z impurities (typically N2) into the private flux region. It has been used in C-Mod to mitigate peak surface heat fluxes  >40 MW m−2 down to  <10 MW m−2 while maintaining excellent core confinement, H98  >  1. While the system works quite well under relatively steady conditions, use of it during transients has revealed important limitations on feedback control of impurity seeding in conventional vertical target plate divertors. In some cases, the system is unable to avoid plasma reattachment to the divertor plate or the formation of a confinement-damaging x-point MARFE. This is due to the small operational window for mitigated heat flux in the parameters of incident plasma heat flux, plasma density, and impurity density as well as the relatively slow response of the impurity gas injection system compared to plasma transients. Given the severe consequences for failure of such a system to operate reliably in a reactor, there is substantial risk that the conventional vertical target plate divertor will not provide an adequately controllable system in reactor-class devices. These considerations motivate the need to develop passively stable, highly compliant divertor configurations and experimental facilities that can test such possible solutions. [ABSTRACT FROM AUTHOR]

The impurity screening response of the high-field side (HFS) scrape-off layer (SOL) to localized nitrogen injection is investigated on Alcator C-Mod for magnetic equilibria spanning lower-single-null, double-null and upper-single-null configurations under otherwise identical plasma conditions. L-mode, EDA H-mode and I-mode discharges are investigated. HFS impurity screening is found to depend on magnetic flux balance and the direction of B  ×   B relative to the most active divertor. Impurity ‘plume’ emission patterns indicate that both parallel and perpendicular (E  ×  B) flows in the SOL contribute to the ‘flushing’ of impurities towards the active divertor, thereby affecting the overall impurity screening behavior. Despite the fact that the HFS SOL is extremely narrow in near-double-null configurations, this SOL is able to screen locally injected nitrogen at least as effectively as the low-field side (LFS) SOL—up to a factor of 10 more effective, depending on specific plasma conditions and whether the magnetic geometry produces parallel flows that work with or against E  ×  B flows. For situations in which the E  ×  B drift of the impurity ions opposes parallel flow toward the primary divertor, HFS impurity screening is found to be least effective. When E  ×  B drifts assist parallel flow toward the primary divertor, HFS impurity screening is found to be very effective. These data support the idea of placing RF actuators and close-fitting wall components on the HFS of the tokamak. With this configuration, near-double-null magnetic topologies may be used for active control of plasma parameters at the antenna/plasma interface for optimal RF coupling, mitigate the generation of local impurities from plasma–material interactions and, taking advantage of favorable plasma flows and good screening properties of the HFS SOL, further minimize the impact of wall-born impurity sources on the plasma core. [ABSTRACT FROM AUTHOR]