Effect Of Divertor Geometry Research Articles

Effects of divertor geometry on detachment and core plasma performance with impurity seeding in EAST

Effects of divertor geometry on detachment and core plasma performance with impurity seeding in EAST

Improved heat and particle flux mitigation in high core confinement, baffled, alternative divertor configurations in the TCV tokamak

Nitrogen seeded detachment has been achieved in the tokamak a configuration variable (TCV) in alternative divertor configurations (ADCs), namely X-divertor and X-point target, with and without baffles in H-mode plasmas with high core confinement. Both ADCs show a remarkable reduction in the inter-ELM particle and heat fluxes to the target compared to the standard divertor configuration. 95%–98% of the inter-ELM peak heat flux to the target is mitigated as a synergetic effect of ADCs, baffling, and nitrogen seeded detachment. The effect of divertor geometry and baffles on core-divertor compatibility is investigated in detail. The power balance in these experiments is also investigated to explore the physics behind the observed reduction in heat fluxes in the ADCs.

Effects of divertor geometry on H-mode pedestal structure in attached and detached plasmas in the DIII-D tokamak

Dedicated experiments have been performed in the DIII-D tokamak to assess the influence of divertor geometry on the H-mode pedestal structure. It has been found that in both attached and detached plasmas, compared to the open divertor, the more closed divertor results in lower density at the pedestal top. Higher pedestal performance can be achieved with the closed divertor in detached plasmas, as manifested by a higher pedestal temperature at the same pedestal density, by about a factor of two, than the open divertor. In addition, approaching divertor detachment by increasing the gas-puffing rate, for different divertor geometries, the pedestal width exhibits different trends. In the attached plasma, the pedestal width agrees well with the theoretical and empirical pedestal-poloidal-beta scaling. However, during divertor detachment, in the open divertor the pedestal width is significantly reduced. In contrast, for detached plasmas with the more closed divertor, the pedestal is significantly wider, by up to 50%, compared to the theoretical scaling. Moreover, near divertor detachment, the open diverted plasma exhibits a more aligned density and temperature pedestal profile, while in the closed divertor the detachment results in a relative shift (up to 50% of the pedestal width) between the density and temperature pedestal profiles.

Investigation of conventional and Super-X divertor configurations of MAST Upgrade using scrape-off layer plasma simulation

One of the first studies of MAST Upgrade divertor configurations with scrape-off layer plasma simulation (SOLPS) 5.0 are presented. We focus on understanding the main prospects associated with the novel geometry of the Super-X divertor (SXD). This includes a discussion of the effect of magnetic flux expansion and volumetric power losses on the reduction of target power loads, the effect of divertor geometry on the divertor closure and the distribution of neutral species and radiation in the divertor, and the role of the connection length in broadening the target wetted area. A comparison in conditions typical for MAST inter-ELM H-mode plasmas confirms the improved performance of the Super-X topology resulting in significantly better divertor closure with respect to neutrals (the atomic flux from the target increased by a factor of 6, but the atomic flux from the divertor to the upper SOL reduced by a factor of 2), increased radiation volume and increased total power loss (a factor of 2) and a reduction of target power loads through both magnetic flux expansion and larger volumetric power loss in the divertor (a factor of 5–10 in attached plasmas). The reduction of the target power load by SXD further increases with collisionality (high density or detached regimes) thanks to the larger importance of volumetric power losses. It is found that a cold divertor plasma leads to stronger parallel temperature gradients in the SOL which drive more parallel heat flux, meaning that the effectiveness of perpendicular transport in spreading the power at the target can be reduced, and this needs to be taken into account in any optimization.

Feasibility of Power and Particle Handling in an ST-FNSF and the Effects of Divertor Geometry

A spherical tokamak (ST) configuration is attractive as the basis for a fusion nuclear science facility (FNSF), due to its small size and relatively low cost. However, the compactness of the ST also exacerbates the power and particle handling problems anticipated in next-step devices, since local fluxes are higher and less space is available for optimizing plasma-facing components. On the other hand, novel divertor geometries that have recently been developed such as the snowflake and super-X divertors can be especially effective at reducing heat fluxes in an ST, helping to meet the exhaust challenge. Here, we present an analysis of the power and particle handling requirements of a candidate ST-FNSF, based on 0-D exhaust projections as well as 2-D edge plasma modeling using the SOLPS code. Both conventional and novel divertor geometries are considered. These show that, for reasonable assumptions on cross-field transport, operating points can be identified that are consistent with both core plasma operation and power and particle exhaust requirements, and these operating points are more easily accessible with novel divertors.

Divertor modeling for the design of the National Centralized Tokamak with high beta steady-state plasmas

The modification of the JT-60U to a fully superconducting coil tokamak, National Centralized Tokamak (NCT) facility, has been programmed to accomplish the high beta steady-state plasma research. A 2D divertor simulation code, SOLDOR/NEUT2D, is applied to the construction of a database for optimum design of the divertor. A semi-closed divertor configuration with vertical target is adopted as the first conceptual divertor design on NCT. With an anticipated SOL power flux of 12 MW at the high beta steady-state operation, the peak heat load on the divertor target is evaluated to be ∼16 MW/m 2. Effects of divertor geometry and intense gas puffing are demonstrated with a view to reduce the heat load. For the simulation of divertor pumping, we find that the pumping efficiency increases by a factor of 2∼3 by narrowing the divertor gap from 20 to 5 cm. An attractive feature in reducing the heat load and improving the particle controllability has been obtained for a new divertor design due to a recent progress in NCT design.

球状トカマクの実用炉への展望 -トカマクの低アスペクト比化への挑戦- 5.ST で現実的な炉設計は可能か? 5.4 低アスペクト比トカマク炉のダイバータ熱流制御

Recent results of edge plasma study in spherical tokamaks (ST) are reviewed. Effect of divertor geometry on allowable steady-state power and edge localized mode (ELM) heat load to target plates were studied. Conceptual case study was made on divertor power handling in a low aspect ratio tokamak reactor. By expanding magnetic flux tube and divertor wetted area, it might be possible to handle heat load without radiative cooling in low triangularity configuration.

Chapter 8: Edge and Divertor Physics in ASDEX Upgrade

An overview of edge and divertor physics research on ASDEX Upgrade of relevance for next-step fusion devices like ITER is presented. The results described were primarily obtained in lower single-null divertor configurations with three consecutive bottom divertor designs, starting from an initial open divertor (Div I) over the closed LYRA configuration (Div II), optimized for low-triangularity single-null equilibria, to the presently operational variant Div IIb, fitting a large variety of plasma shapes. The upper, geometrically open divertor structure remained essentially unchanged. A dedicated diagnostics system in combination with advanced plasma control scenarios and extensive numerical modeling allowed for a detailed analysis of edge and divertor physics mechanisms. Main chamber edge profiles exhibit a double structure, especially pronounced in high-performance H-mode plasmas. While radial transport inside and across the separatrix is governed by critical gradients, the cold scrape-off layer wing shows rapid diffusion or even outward drift, probably related to intermittent crossfield transport. The divertor behavior has been studied for the different divertor geometries and for all operational regimes of interest. Closed divertor operation enhances divertor recycling and pumping, reduces the power load on target plates by increased upstream losses, and facilitates onset of plasma detachment. The transient power load during type I ELMs, however, remains high and problematic, while the small type III ELMs, appearing, for example, in radiative discharge scenarios, and especially the type II ELMs are nearly invisible on the target heat flux. Despite this strong effect of divertor geometry on the divertor behavior, its direct effect on core confinement remains small.

Effects of divertor geometry and chemical sputtering on impurity behaviour and plasma performance in JET

The effects of increased geometrical closure on thebehaviour of the recycling and intrinsic impurities are investigatedin JET Mark I, Mark IIA and Mark IIGB pumped divertors. Increasing the divertorclosure leads to a significant improvement in exhaust for bothdeuterium and recycling impurities. However, the impurity enrichmentin the exhaust gases remains unchanged due to a simultaneous increase indeuterium and impurity compression in the divertor. A comparison ismade for helium, neon and argon under different plasma conditions. In addition,the operation of the Mark II and Mark IIGB divertors has shown thatZeff is reduced with the improved divertor closure in the L modedischarges, although no obvious changes in the Zeff values have beenobserved in the ELMy H modes. The divertor target surface temperaturehas a strong influence on intrinsic carbon production. The carbonsource in the Mark II and Mark IIGB divertors is significantly higherthan that in the Mark I divertor, which is attributed to enhancedchemical sputtering at the increased divertor tile temperature of theMark II and Mark IIGB divertors (related to the divertor coolingsystem), as opposed to the increased closure. The consequences of thiselevated yield for plasmas under different operation conditions arediscussed, and further evidence, obtained from a specificwall/divertor temperature reduction experiment, is presented. Theeffect of the divertor screening on the chemically producedimpurities is investigated using the EDGE2D/NIMBUS/DIVIMP codes forthe different recycling regimes and comparisons are made with experimental observations from the Mark I, Mark IIA and Mark IIAP divertors taking into accountthe change in chemical sputtering yield due to the different tiletemperatures of these divertors.

The effect of divertor geometry on divertor and core plasma performance in JET

JET has completed a series of experiments in the Mk I and Mk IIA divertors on the effects of increased geometrical closure and target orientation. The potential benefits from closure were expected to be enhanced volumetric energy loss in the divertor (detachment), increased divertor neutral pressure for better pumping and He exhaust, and reduced main chamber neutral pressure for reduced sputtering. The expected effects on neutral pressures were observed. In ohmic and L-modes this led to detachment at lower upstream density and reduced density limits, in qualitative agreement with code calculations. The pumping speed was increased by about a factor of three. Z eff did not reduce, despite the reduced main chamber neutral pressure. In ELMy H-modes the effects of closure were less distinct, which may have been due in part to ELMs striking the upper surfaces of the divertor and main chamber limiting surfaces. The density limit and confinement quality were unaffected by changes in divertor geometry. Increasing triangularity increased the density limit, but also raised Z eff. Confinement was degraded by either deuterium puffing or nitrogen puffing. Detachment occurred at the inner target between ELMs, but not at the outer target until confinement was strongly degraded. Vertical target ELMy H-modes have thinner SOL's and lower midplane separatrix densities than those run on horizontal targets in Mk IIA. Given the JET observations on the lack of sensitivity of core plasma ELMy H-mode performance to divertor geometry, it appears appropriate to review the possibility of simpler, lower cost divertor options than the deep divertor design currently proposed for ITER.

Effect of divertor geometry on plasma detachment in DIII-D

UEDGE simulation of the radiative divertor plasma (RDP) (nearly closed divertor) in DIII-D indicates a 40% reduction in the core density necessary for fully detached pure deuterium plasmas at 3 MW core power. Including intrinsic carbon impurities to the UEDGE model expands the region for fully detached plasmas and further reduces the necessary core density for this state.

Effect of divertor geometry on boundary and core plasma performance in ASDEX Upgrade and JET

A key element in future toroidal magnetic fusion machines like ITER is the design of a divertor, which allows for safe particle and power exhaust in parallel with high bulk plasma performance. Correspondingly, the definition and design of an optimized divertor is a major task within the ongoing international ITER research and development effort. In order to provide a profound physics basis for such a divertor optimization, different divertor geometries are being tested on major tokamaks. This paper describes the effects of these divertor modifications on plasma performance in ASDEX Upgrade and in JET. In conclusion, increasing closure improves divertor performance without limiting the core plasma performance in ELMy H-modes.

Effect of divertor geometry on impurity exhaust

The effect of poloidal divertor geometry on the exhaust of impurities is studied experimentally. In the expanded-boundary geometry, strong accumulation of argon in the divertor region and minimal reflux into the main plasma are observed. A simple model provides an explanation for the observations and leads to the conclusion that the two key geometrical factors for effective impurity control are an expansion of the divertor channel and a narrow gap between the divertor channel and the wall.