Divertor Geometry Research Articles

Drive of parallel flows by turbulence and large-scale E × B transverse transport in divertor geometry

The poloidal asymmetries of parallel flows in edge plasmas are investigated by the 3D fluid turbulence code TOKAM3X. A diverted COMPASS-like magnetic equilibrium is used for the simulations. The measurements and simulations of parallel Mach numbers are compared, and exhibit good qualitative agreement. Small-scale turbulent transport is observed to dominate near the low field side midplane, even though it co-exists with significant large-scale cross-field fluxes. Despite the turbulent nature of the plasma in the divertor region, simulations show the low effectiveness of turbulence for the cross-field transport towards the private flux region. Nevertheless, a complex pattern of fluxes associated with the average field components are found to cross the separatrix in the divertor region. Large-scale and small-scale turbulent transport, along with the drift, drive the asymmetries in parallel flows. A semi-analytical model based on mass and parallel momentum balances allows the poloidal drift effects on the asymmetry pattern to be evaluated. As in the experiments, a reversed B T simulation provides a way of self-consistently separating the effects of turbulent transport and large-scale flows, which must be reversed for a reversed field. The large-scale contribution is found to be responsible for typically 50% of the effect on the Mach number, evaluated at the top of the machine. The presented picture shows the complex interplay between drifts and turbulence, underlining the necessity of a global approach to edge plasma modelling, including a self-consistent description of the turbulence.

Characterized the pattern of the material deposition in the HL-2A tokamak

Since the divertor geometry of a tokamak has a strong impact on the material erosion and deposition on the wall and HL-2A has a unique divertor configuration, it is necessary to investigate the material deposition pattern in HL-2A although a few results on other tokamaks have already been published.In this paper, tiles retrieved from the vessel are analyzed ex-situ by SIMS, SEM and laser-induced breakdown spectroscopy (LIBS). And deposition behind the lower divertor is in-situ measured by a quartz crystal microbalance (QMB). The deposition in HL-2A displays a complex pattern and clear localization characteristic. The thickness of the deposition layer varies in the range of 0-4μm. And in-situ diagnostic of QMB indicates that the average thickness of the deposition layer per pulse is over ten nanometers. In addition, the results imply that Si, Fe and D have different behaviors during the material deposition in HL-2A.

Possible influence of near SOL plasma on the H-mode power threshold

A strong effect of divertor configuration on the threshold power for the L-H transition (PLH) was observed in recent JET experiments in the new ITER-like Wall (ILW) [1–3]. Following a series of EDGE2D-EIRENE code simulations with Be impurity and drifts a possible mechanism for the PLH variation with the divertor geometry is proposed. Both experiment and code simulations show that in the configuration with lower neutral recycling near the outer strike point (OSP), electron temperature (Te) peaks near the OSP prior to the L-H transition, while in the configuration with higher OSP recycling Te peaks further out in the scrape-off layer (SOL) and the plasma stays in the L-mode at the same input power. Code results show large positive radial electric field (Er) in the near SOL under lower recycling conditions leading to a large E×B shear across the separatrix which may trigger earlier (at lower input power) edge turbulence suppression and lower PLH. Suppressed Te‘s at OSP in configurations with strike points on vertical targets (VT) were observed earlier and explained by a geometrical effect of neutral recycling near this particular position, whereas in configurations with strike points on horizontal targets (HT) the OSP appears to be more open for neutrals (see e.g. review paper [4]).

Effect of plasma geometry on divertor heat flux spreading: MONALISA simulations and experimental results from TCV

Safe ITER operations will rely on power spreading to keep the peak heat flux within divertor material constraints. A solid understanding and parameterization of heat flux profiles is therefore mandatory. This paper focuses on the impact of plasma geometry on the power decay length (λq) and the spreading factor (S). Numerical heat flux profiles, obtained with the simple SOL transport model MONALISA, agree with theoretical predictions for purely diffusive cylindrical plasmas: λq does not depend on the machine-specific divertor geometry but only on transport parameters and global geometry (a, R, k). A dedicated experiment on TCV was designed to further test this assumption in L-mode plasmas with similar control parameters and upstream shape but different divertor leg length (Zmag=−14, 0, 28 cm). Characterization of OSP q∥ profiles with Langmuir probes and infrared thermography enlightens unexpected behavior with the divertor leg length: λq increases, while S shows no clear trend. These findings suggest that the link between heat flux profiles, plasma geometry and transport is currently not fully understood.

Investigations on the heat flux and impurity for the HL-2M divertor

The controllability of the heat load and impurity in the divertor is very important, which could be one of the critical problems to be solved in order to ensure the success for a steady state tokamak. HL-2M has the advantage of the poloidal field (PF) coils placed inside the demountable toroidal field (TF) coils and close to the main plasma. As a result, it is possible to make highly accurate configuration control of the advanced divertor for HL-2M. The divertor target geometry of HL-2M has been designed to be compatible with different divertor configurations to study the divertor physics and support the high performance plasma operations. In this paper, the heat loads and impurities with different divertor configurations, including the standard X-point divertor, the snowflake-minus divertor and two tripod divertor configurations for HL-2M, are investigated by numerical simulations with the SOLPS5.0 code under the current design of the HL-2M divertor geometry. The plasmas with different conditions, such as the low discharge parameters with = 0.5 MA at the first stage of HL-2M and the high parameters with = 2.0 MA during the normal operations, are simulated. The heat load profiles and the impurity distributions are obtained, and the control of the peak heat load and the effect of impurity on the core plasma are discussed. The compatibility of different divertor configurations for HL-2M is also evaluated. It is seen that the excellent compatibility of different divertor configurations with the current divertor geometry has been verified. The results show that the snowflake-minus divertor and the tripod divertor with present good performance in terms of the heat load profiles and the impurity distributions under different conditions, which may not have a big effect on the core plasma. In addition, it is possible to optimize the distance between the two X-points, , to achieve a better performance in terms of the parameters of discharges.

Experimental results of H-mode power threshold with lower hybrid wave heating on the EAST tokamak

The density roll-over dependence on H-mode power threshold is observed on EAST for the first time. In campaign 2014 and 2015 shots with a toroidal field of 2.25 T have observed roll-over dependence with lower hybrid wave as the only auxiliary heating method, while shots with a toroidal field of 1.79 T and 1.9 T exhibit linear dependence consistent with scaling law. The density of minimum power for accessing H-mode on EAST has different plasma current values of 400 kA and 500 kA, and is better described in the normalized Greenwald fraction at . The absence of in 1.7 T and 1.8 T may be attributed to the positive dependence with toroidal field. Besides, correlation analysis of H-mode power threshold and divertor geometry in scanning X-point is summarized and compared. Outer leg length (distance from X-point to outer strike point) has the highest correlation coefficient with H-mode power threshold, which explains the data scattering within the same plasma parameters. A new equation of scaling law is proposed: Neutral particles are believed to be the hidden factor in different divertor geometry, and play a negative role in L–H transition via charge exchange damping.

L–H power threshold studies with tungsten/carbon divertor on the EAST tokamak

ABSTRACTThe power threshold for low (L) to high (H) confinement mode transition achieved by radio-frequency heating and molybdenum first wall with lithium coating has been experimentally investigated on the EAST tokamak for two sets of divertor geometries and materials: tungsten/carbon divertor and full carbon divertor. For both sets of divertors, the power threshold was found to decrease with gradual accumulation of the lithium wall coating, suggesting the important role played by the low Z impurities and/or the edge neutral density on the L–H power threshold. When operating in the upper single null configuration, with the ion grad-B drift direction away from the primary X-point, a lower normalized power threshold is observed in EAST with the tungsten/carbon divertor, compared to the carbon divertor after intensive lithium wall coating. A newly installed cryopump increasing the pumping efficiency also plays an important part in the observed lower threshold. In addition, the H-mode in the Quasi-Snowflake divertor configuration has been obtained on EAST, exhibiting higher L–H power threshold compared to the lower single null configuration with similar IP/BT pairs.

Continuum kinetic modeling of the tokamak plasma edge

The first 4D (axisymmetric) high-order continuum gyrokinetic transport simulations that span the magnetic separatrix of a tokamak are presented. The modeling is performed with the COGENT code, which is distinguished by fourth-order finite-volume discretization combined with mapped multiblock grid technology to handle the strong anisotropy of plasma transport and the complex X-point divertor geometry with high accuracy. The calculations take into account the effects of fully nonlinear Fokker-Plank collisions, electrostatic potential variations, and anomalous radial transport. Topics discussed include: (a) ion orbit loss and the associated toroidal rotation and (b) edge plasma relaxation in the presence of anomalous radial transport.

Impact of divertor geometry on radiative divertor performance in JET H-mode plasmas

Radiative divertor operation in JET high confinement mode plasmas with the ITER-like wall has been experimentally investigated and simulated with EDGE2D-EIRENE in horizontal and vertical low field side (LFS) divertor configurations. The simulations show that the LFS divertor heat fluxes are reduced with N2-injection in similar fashion in both configurations, qualitatively consistent with experimental observations. The simulations show no substantial difference between the two configurations in the reduction of the peak LFS heat flux as a function of divertor radiation, nitrogen concentration, or pedestal Zeff. Consistently, experiments show similar divertor radiation and nitrogen injection levels for similar LFS peak heat flux reduction in both configurations. Nevertheless, the LFS strike point is predicted to detach at 20% lower separatrix density in the vertical than in the horizontal configuration. However, since the peak LFS heat flux in partial detachment in the vertical configurations is shifted towards the far scrape-off layer (SOL), the simulations predict no benefit in the reduction of LFS peak heat flux for a given upstream density in the vertical configuration relative to a horizontal one. A factor of 2 reduction of deuterium ionization source inside the separatrix is observed in the simulations when changing to the vertical configuration. The simulations capture the experimentally observed particle and heat flux reduction at the LFS divertor plate in both configurations, when adjusting the impurity injection rate to reproduce the measured divertor radiation. However, the divertor Dα-emissions are underestimated by a factor of 2–5, indicating a short-fall in radiation by the fuel species. In the vertical configuration, detachment is experimentally measured and predicted to start next to the strike point, extending towards the far SOL with increasing degree of detachment. In contrast, in the horizontal configuration, the entire divertor particle flux profile is reduced uniformly with increasing degree of detachment.

Development of a free-boundary tokamak equilibrium solver for advanced study of tokamak equilibria

A free-boundary Tokamak Equilibrium Solver (TES), developed for advanced study of tokamak equilibra, is described with two distinctive features. One is a generalized method to resolve the intrinsic axisymmetric instability, which is encountered after all in equilibrium calculation with a free-boundary condition. The other is an extension to deal with a new divertor geometry such as snowflake or X divertors. For validations, the uniqueness of a solution is confirmed by the independence on variations of computational domain, the mathematical correctness and accuracy of equilibrium profiles are checked by a direct comparison with an analytic equilibrium known as a generalized Solovev equilibrium, and the governing force balance relation is tested by examining the intrinsic axisymmetric instabilities. As a valuable application, a snowflake equilibrium that requires a second order zero of the poloidal magnetic field is discussed in the circumstance of KSTAR coil system.

Nonlinear fluid simulation of particle and heat fluxes during burst of ELMs on DIII-D with BOUT++ code

In order to study the distribution and evolution of the transient particle and heat fluxes during edge-localized mode (ELM) bursts, a BOUT++ six-field two-fluid model based on the Braginskii equations with non-ideal physics effects is used to simulate pedestal collapse in divertor geometry. The profiles from the DIII-D H-mode discharge #144382 with fast target heat flux measurements are used as the initial conditions for the simulations. A flux-limited parallel thermal conduction is used with three values of the flux-limiting coefficient , free streaming model with , sheath-limit with , and one value in between. The studies show that a 20 times increase in leads to ∼6 times increase in the heat flux amplitude to both the inner and outer targets, and the widths of the fluxes are also expanded. The sheath-limit model of flux-limiting coefficient is found to be the most appropriate one, which shows ELM sizes close to the measurements. The evolution of the density profile during the burst of ELMs of DIII-D discharge #144382 is simulated, and the collapse in width and depth of are reproduced at different time steps. The growing process of the profiles for the heat flux at divertor targets during the burst of ELMs measured by IRTV (infrared television) is also reproduced by this model. The widths of heat fluxes towards targets are a little narrower, and the peak amplitudes are twice the measurements possibly due to the lack of a model of divertor radiation which can effectively reduce the heat fluxes. The magnetic flutter combined with parallel thermal conduction is found to be able to increase the total heat loss by around 33% since the magnetic flutter terms provide the additional conductive heat transport in the radial direction. The heat flux profile at both the inner and outer targets is obviously broadened by magnetic flutter. The lobe structures near the X-point at LFS are both broadened and elongated due to the magnetic flutter.

Study of impacts on tritium breeding ratio of a fusion DEMO reactor

The existing tritium resource available for future fusion power plant D-T operation is severely limited. It is essential to breed tritium to maintain the continuous consumption in the D-T plasma so as to sustain the required fusion power. A minimum criterion of calculated tritium breeding ratio (TBR) >1.1 is adopted to ensure tritium self-sufficiency, allowing a 10% safety margin to cover calculation uncertainties. In a DEMO reactor, there are various factors impacting the availability and effectiveness of fusion generated neutrons for production of tritium. In this paper, the study focused on the impact of first wall design including materials and first wall thickness, and blanket coverage aspects such as divertor geometry and the provision of equatorial heating and diagnostic ports.

Solid tungsten Divertor-III for ASDEX Upgrade and contributions to ITER

ASDEX Upgrade became a full tungsten experiment in 2007 by coating its graphite plasma facing components with tungsten. In 2013 a redesigned solid tungsten divertor, Div-III, was installed and came into operation in 2014. The redesign of the outer divertor geometry provided the opportunity to increase the pumping efficiency in the lower divertor by increasing the gap between divertor and vessel. In parallel, a by-pass was installed into the cryo-pump in the divertor region allowing adapting of the pumping speed to the required edge density.Safe divertor operation and heat removal becomes more and more significant for future fusion devices. This requires developing ‘tools’ for divertor heat load control and to optimize the divertor design. The new divertor manipulator, DIM-II, allows retracting a relevant part of the outer divertor into a target exchange box without venting ASDEX Upgrade. Different front-ends can be installed and exposed to the plasma. At present, front-ends for probe exposition, gas puffing, electrical probes and actively cooled prototype targets are under construction.The installation of solid tungsten, the control of the pumping speed and the flexibility for testing divertor modifications on a weekly base is a unique feature of ASDEX Upgrade and offers together with the extended set of diagnostics the possibility to investigate dedicated questions for a future divertor design.

Predictive Modeling for Performance Assessment of ITER-Like Divertor in China Fusion Engineering Testing Reactor

To facilitate the design of the China Fusion Engineering Testing Reactor (CFETR), predictive modeling for the assessment and optimization of the divertor performances is an indispensable approach. This paper presents the modeling of the edge plasma behaviors as well as the W erosion and transport properties in CFETR with ITER-like divertor by using the B2-Eirene/SOLPS 5.0 code package together with the Monte Carlo impurity transport code DIVIMP. As expected, SOLPS modeling of divertor-SOL plasmas finds that the peak heat flux onto the divertor targets greatly exceeds 10 MW/m2, an engineering limit posed to the steady-state and/or long-pulse operation of the next-step fusion devices, for a wide range of plasma conditions, and thus modeling of Ar puffing by scanning the puffing rate for radiative divertor is performed. As the increase of the Ar puffing rate, the peak target heat fluxes and plasma temperature decreases exponentially,reflecting that Ar puffing is highly effective at power exhausting. Based on the ion fluxes from SOLPS, the W erosion is calculated by taking into consideration the bombardment of both D and Ar ions, and then the W plasma concentrations are calculated based on the W erosion fluxes using DIVIMP. The calculations show that if the Ar puffing only being used to reduce the divertor heat load, the W plasma contamination in the core plasma exceeds the tolerable value (<10−5), which demonstrates that some further upgrading of the divertor geometry is still needed.

The development of safe high current operation in JET-ILW

The JET tokamak is unique amongst present fusion devices in its capability to operate at high plasma current, providing the closest plasma parameters to ITER. The physics benefits of high current operation have to be balanced against the risks to the integrity of the machine due to high force disruptions. The installation of the ITER-Like Wall (ILW) has added risks due to the thermal characteristics of the metal Plasma Facing Components. This paper describes the operational aspects of the scientific development of high current H-mode plasmas with the ILW, focusing on disruption prediction, avoidance and amelioration. The development yielded baseline H-mode plasmas up to 4MA/3.74T, comparable to the maximum current achieved in JET in Carbon-Wall (CFC) conditions with similar divertor geometry.

Engineering conceptual design of CFETR divertor

The China Fusion Engineering Test Reactor (CFETR), which is in conceptual design phase, aims at producing fusion power of 50–200MW with tritium breeding ratio of ∼1.2 and duty cycle time of 0.3–0.5. Its designed main parameters are major/minor radii of 5.7m/1.6m and plasma current of 10MA. Although the fusion power is lower than the one of ITER, the relative smaller machine dimensions and planed much higher auxiliary heating power of 100–140MW make that the power exhausting for the CFETR divertor is a very critical issue. To solve this issue, the divertor should be better designed with advanced physical operation mode, advanced configuration/geometry or high efficient cooling structure. In the paper, much effort was put on the divertor configuration and geometry. With designed magnet system, three divertor configurations can be realized, ITER-like, snowflake and super-X. However, considering structural design feasibility and remote handling compatibility, only the first two configurations were selected for the first step of engineering design. Three divertors were designed. They have different first wall geometries to accommodate with different plasma configurations, one for the ITER-like, one for the snowflake and the third one for both the configurations. All three divertors employ the same cassette body as the support and the cooling water manifold for the first wall. This feature simplifies the interface of the divertor to other components in the vacuum vessel. Besides, the cooling structure and the remote maintenance concept are also introduced in the paper.

Physics Basis for an Advanced Physics and Advanced Technology Tokamak Power Plant Configuration: ARIES-ACT1

The advanced physics and advanced technology tokamak power plant ARIES-ACT1 has a major radius of 6.25 m at an aspect ratio of 4.0, toroidal field of 6.0 T, strong shaping with elongation of 2.2, and triangularity of 0.63. The broadest pressure cases reached wall-stabilized βN ∼ 5.75, limited by n = 3 external kink mode requiring a conducting shell at b/a = 0.3, requiring plasma rotation, feedback, and/or kinetic stabilization. The medium pressure peaking case reaches βN = 5.28 with BT = 6.75, while the peaked pressure case reaches βN < 5.15. Fast particle magnetohydrodynamic stability shows that the alpha particles are unstable, but this leads to redistribution to larger minor radius rather than loss from the plasma. Edge and divertor plasma modeling shows that 75% of the power to the divertor can be radiated with an ITER-like divertor geometry, while =95% can be radiated in a stable detached mode with an orthogonal target and wide slot geometry. The bootstrap current fraction is 91% with a q95 of 4.5, requiring ∼1.1 MA of external current drive. This current is supplied with 5 MW of ion cyclotron radio frequency/fast wave and 40 MW of lower hybrid current drive. Electron cyclotron is most effective for safety factor control over ρ ∼ 0.2 to 0.6 with 20 MW. The pedestal density is ∼0.9×1020/m3, and the temperature is ∼4.4 keV. The H98 factor is 1.65, n/nGr = 1.0, and the ratio of net power to threshold power is 2.8 to 3.0 in the flattop.

Comparisons between tokamak fueling of gas puffing and supersonic molecular beam injection in 2D simulations

Plasma fueling with high efficiency and deep injection is very important to enable fusion power performance requirements. It is a powerful and efficient way to study neutral transport dynamics and find methods of improving the fueling performance by doing large scale simulations. Two basic fueling methods, gas puffing (GP) and supersonic molecular beam injection (SMBI), are simulated and compared in realistic divertor geometry of the HL-2A tokamak with a newly developed module, named trans-neut, within the framework of BOUT++ boundary plasma turbulence code [Z. H. Wang et al., Nucl. Fusion 54, 043019 (2014)]. The physical model includes plasma density, heat and momentum transport equations along with neutral density, and momentum transport equations. Transport dynamics and profile evolutions of both plasma and neutrals are simulated and compared between GP and SMBI in both poloidal and radial directions, which are quite different from one and the other. It finds that the neutrals can penetrate about four centimeters inside the last closed (magnetic) flux surface during SMBI, while they are all deposited outside of the LCF during GP. It is the radial convection and larger inflowing flux which lead to the deeper penetration depth of SMBI and higher fueling efficiency compared to GP.

Investigation of the influence of divertor recycling on global plasma confinement in JET ITER-like wall

The impact of the divertor geometry on global plasma confinement in type I ELMy H-mode has been investigated in the JET tokamak equipped with ITER-Like Wall. Discharges have been performed in which the position of the strike-points was changed while keeping the bulk plasma equilibrium essentially unchanged. Large variations of the global plasma confinement have been observed, the H98 factor changing from typically 0.7 when the outer strike-point is on the vertical or horizontal targets to 0.9 when it is located in the pump duct entrance. Profiles are mainly impacted in the pedestal but core gradient lengths, especially for the density, are also modified. Although substantial differences are observed in the divertor conditions, none seem to correlate directly with the confinement. Modelling with the EDGE2D-EIRENE and SOLEDGE2D-EIRENE transport codes exhibits differences in the energy losses due to neutrals inside the separatrix, but orders of magnitude are too low to explain simply the impact on the confinement.

Evaluation of target-plate heat flux for a possible snowflake divertor in CFETR using SOLPS

China Fusion Engineering Test Reactor (CFETR) is proposed as a good complement to ITER for demonstration of fusion energy. CFETR is based on both physics and some technologies of ITER. The main goals of CFETR are fusion power Pf=50–200MW, duty cycle time ⩾0.3–0.5, and a tritium breeding ratio ∼1.2. To explore a more effective way to manage heat exhaust in a future fusion reactor, which will have higher heating power (auxiliary heating power plus 20% of the fusion power) than ITER, a snowflake divertor (SFD) is an optional choice considered for CFETR. In this paper, the preliminary design of a SFD for CFETR is presented, including the divertor magnetic configuration and geometry. A numerical simulation is then performed to evaluate the heat flux onto the divertor targets by using SOLPS.