Single Null Divertor Configuration Research Articles

High-Z impurity neoclassical transport in the tokamak plasmas due to the single-null divertor configuration

A new mechanism for the high-Z impurity neoclassical particle transport in the tokamak plasmas due to the single-null divertor configuration is discovered for the first time. It will play an important role in the tokamak plasma pedestal region with the coexisting of the strong bulk ion radial gradients and the up/down asymmetry of the poloidal magnetic field. The outward (inward) high-Z impurity neoclassical particle transport will be driven with the B×∇B drift towards (away from) the X-point. The new finding indicates that the International Thermonuclear Experimental Reactor lower single-null divertor configuration will be beneficial for screening and flushing out the high-Z impurity in the pedestal region with the toroidal magnetic field in the normal direction ( B×∇B drift towards the X-point).

Parameter dependence of density pedestal width and correlation with pedestal turbulence in EAST type-I ELMy H-mode plasma

The density pedestal structures in a series of type-I ELMy H-mode discharges (0.6 MA/2.5 T) with upper single null (USN) divertor configuration have been studied in EAST. To obtain the pedestal width, density profiles are fitted via the modified hyperbolic tangent (mtanh) function and the two-line method. The pedestal widths from these two methods have a similar tendency. Significant impacts of pedestal poloidal beta and pedestal collisionality on the density pedestal width have been observed for the first time in EAST. The non-linear regressions suggest the density width scaling in type-I ELMy EAST plasma are wne∝c(νped⁎)0.27−0.33⋅(βp,ped)0.5−0.58. The density fluctuations in the bottom of the pedestal from the fluctuation reflectometry imply an enhanced density fluctuation (Pfluc) and an obvious spectral broadening in the high collisionality regime of νped⁎≥3. The experimental results suggest that the pedestal turbulence would play an important role in the evolution of the pedestal structure.

Analysis of edge transport in L-mode negative triangularity TCV discharges

One of the major problems for future tokamak devices are ELMs (Edge Localized Modes) as they can lead to large, uncontrolled heat fluxes at the machine targets. For this reason, different techniques and alternative magnetic configurations are under study to mitigate or avoid these phenomena. One of the most promising among these studies is the Negative Triangularity (NT) configuration, which exhibits a global confinement comparable with H-Mode operation and, staying in L-mode, could enable an easier power exhaust dissipation due to a possible bigger heat flux decay length with respect to the conventional Positive triangularity (PT) H-mode. In this work, the fluid code SOLEDGE2D-EIRENE is used to study edge transport. Studies are made on discharges performed in the TCV device (Tokamak à configuration variable), which can create a variety of different plasma geometries thanks to 16 independently powered poloidal field coils and its open vacuum vessel. In order to understand if and how power and particle exhaust in NT differs with respect to those of the PT shape, four discharges in single null magnetic divertor configuration with fixed lower triangularity (δbot=+0.5) but with different upper triangularity (from δup=-0.28 to δup=+0.45) have been modelled. All of them are ohmically heated, L-mode deuterium plasmas, and in the high recycling regime. Moreover, these discharges were previously used in [4] to measure the heat flux decay length by IRT (Infrared Thermography), allowing us to make comparisons with modelling results.

Design of a multi-configurations divertor for the DTT facility

The standard positive triangularity H-mode tokamak operation with Single Null Divertor (SND) configuration and tungsten monoblocks targets can face some difficulties in providing a solution scalable towards the realization of the fusion reactor. To provide a safer solution, the adoption of alternative divertor magnetic configurations (ADCs) has been considered to provide on the targets a stationary heat load of less than 10 MW/m2 presently considered an upper technological limit for a nuclear fusion reactor and to reduce enough plasma temperature to avoid tungsten sputtering and its possible plasma contamination. Additionally, different plasma scenarios have been considered to avoid the huge transient energy released due to the type-I ELMs of the high confinement H-mode operation. The new high field superconducting divertor tokamak test facility (DTT) [1] is presently under construction to specifically study power exhaust solutions in reactor relevant regimes. The first DTT divertor will use the ITER-like technology based on full tungsten monoblocks bonded on CuCrZr cooling tubes. To test and compare a wide set of different power exhaust solutions, the divertor is being designed to be compatible with the standard single null (SND) configuration but also with some ADCs, like the X divertor (XD) and the hybrid Super-X/long leg SN. Additionally, the l-mode negative triangularity (NT) operation is considered important to explore as a solution to avoid ELMs. Two different closed divertors with the same grazing angle (α = 2°) for the reference SND configuration have been considered: a so called “narrow” one with normalized divertor parameters similar to those presently foreseen for DEMO divertor and the “wide” one with a 50 % wider target aperture which allows testing long legs configurations as well as snowflake divertor configurations (SFD). For the “wide” divertor different dome sizes have been also considered. The comparison between different shapes and their optimization have been done with the 2D edge fluid-kinetic code SOLEDGE2D-EIRENE. The code easily manages all magnetic configurations and can resolve the heat load on all components thanks the mesh filling the whole divertor and vessel volume. In a representative sub-set of all the possible magnetic equilibrium configurations that can be realized in DTT, the edge modelling shows a better performance of the “wide” divertor compared to the “narrow” one. In fact, the “wide” divertor allows to have lower plasma temperature at the targets in pure deuterium at low PSOL power and also lower impurity concentrations is required to achieve plasma detachment by impurity seeding at the reference full power DTT scenario. The results seem to indicate that the improvement due to the longer legs possible with the “wide” divertor can overcome its lower apparent closure compared to the “narrow” divertor solution.

Predictive modeling of the deuterium gas puffing effect on impurity and heat flux compression for the HL-2M divertor by SOLPS

To study the effect of gas puffing on divertor operation regime and impurity level in the SOL region, we introduce deuterium gas puffing and pumping, based on the HL-2M lower single null (LSN) divertor configuration (Ip=2MA), using the SOLPS5.0 code. The simulation result indicates that the location of the gas puffing port impacts the divertor plasma condition, and puffing deuterium from the location near the divertor region can enhance the power dissipation and the divertor regime transformation from high recycling region to detachment. Besides, increasing deuterium gas puffing rate can strengthen the core confinement. Meanwhile, the impurity level in the SOL region is compressed rapidly, from Zeff=2.0 to 1.1, with a reasonable deuterium gas puffing rate of 2.5×1022s−1 puffed from the optimized port and a constant chemical sputtering yield Ychem of 1.0%, compared to the case with no gas puffing. The physics interpretation behind this phenomenon is also analyzed. It is demonstrated that the ionization rate distribution and the carbon ions’ poloidal velocity define the impurity flow in the SOL region. The higher ionized carbon ions (C4+ to C6+) retain in the ionization region and even flow downstream when they finish ionization before reaching the stagnation point with gas puffing. Then the impurity ions accumulate in the divertor region. The balance between the thermal and friction forces acting on the carbon ions results in the change of poloidal velocity. In general, deuterium gas puffing and pumping induces the modification of impurity flow and the decline of the Zeff. Furthermore, a chemical sputtering yield of 3.0% is also calculated to verify the efficiency of deuterium gas puffing on Zeff compression.

Effect of equilibrium radial position on horizontal instability of middle single null divertor configuration in the iron core tokamak

The unique high-field side middle single null divertor configuration in the Joint-Texas Experimental Tokamak is attractive in studying detachment, impurity control, L-H transition, and negative triangularity. Its equilibrium configuration with a horizontally elongated (or flattened) poloidal cross section causes position instability and difficulty in equilibrium control. This work disassembles the vertical magnetic field of each group of poloidal field windings and calculates the field decay index (n-index) at different radial positions. It is found that the instability growth time τg of horizontal displacement increases as the equilibrium radial position of plasma moves outward, which is beneficial to the position control. The simulation and experiments verified the effect of equilibrium position on horizontal instability. The equilibrium position slightly outward from the geometric center of the vacuum chamber effectively decreases the horizontal displacement oscillation amplitude and guarantees the controllable horizontal displacement.

Design and feasibility of breeding blanket vertical segment-based architecture

A design integration study (KDII4) was conducted in the DEMO Pre-Concept Design Phase with the primary goal to develop a pre-concept feasibility design and concept of operation for the Vertical Blanket Segment Architecture. The primary goal was to develop two workable variants for removal of large in-vessel components. This, prompted by two earlier studies, highlighting the integration challenge finding a self-consistent DEMO design point (KDII) and secondly the complexity of operation required to ensure compatibly with the proposed port-based maintenance schemes. Initially, only a single null (SN) divertor configuration was considered, but due to additional identified issues and technical challenges centred on the extraction of the BB segments, a double null (DN) alternative variant has also been investigated. This was prompted by other KDII's studies, especially KDII1 (Design, performance and feasibility of wall protection limiters during plasma transients) and KDII3 (Advanced Magnetic Configurations). Alternative forms of vertical maintenance architecture have been investigated by breaking down the study into Ports, In-vessel Components, Operations and Safety. SN and DN configurations were studied with split and full blankets numbering seven variants in total. Each variant was studied in some detail, comparatively assessing each on its merits. However, no viable solution has been found for down-selection. Nevertheless, the question of vertical maintenance is now better understood. This paper will describe the findings of KDII4 in conjunction with the Remote Maintenance (RM) technology work package, which suggests that the current ‘reference’ design envelope available for RM is too constrained.

Influence of the X-point location on edge plasma transport in the J-TEXT tokamak with a high-field-side single-null divertor

High-density experiments in the high-field-side mid-plane single-null divertor configuration have been performed for the first time on J-TEXT. The experiments show an increase in the highest central channel line-averaged density from to , while the X-point moves away from the target by increasing the divertor coil current. The corresponding Greenwald fraction rises from 0.50 to 0.79. For the impurity transport, the density normalized radiation intensity (absolute extreme ultraviolet and soft x-ray) of the central channel density decreased significantly (>50%) with an increase in the plasma density. To better understand the underlying physics mechanisms, the 3D edge Monte Carlo code coupled with EIRENE (EMC3- EIRENE) has been implemented for the first time on J-TEXT. The simulation results show good agreement with the experimental findings. As the X-point moves away from the target, the divertor power decay length drops and the scrape-off layer impurity screening effect is enhanced.

First principle-based multi-channel integrated modelling in support of the design of the Divertor Tokamak Test facility

An intensive integrated modelling work of the main scenarios of the new Divertor Tokamak Test (DTT) facility with a single null divertor configuration has been performed using first principle quasi-linear transport models, in support of the design of the device and of the definition of its scientific work programme. First results of this integrated modelling work on DTT (R 0 = 2.14 m, a = 0.65 m) are presented here along with outcome of the gyrokinetic simulations used to validate the reduced models in the DTT range of parameters. As a result of this work, the heating mix has been defined, the size of device has been increased to R 0 = 2.19 m and a = 0.70 m, the use of pellets for fuelling has been recommended and reference profiles for diagnostic design, estimates of neutron yields and fast particle losses have been made available.

The effects of E r × B t drift on LH wave coupling with divertor configuration at EAST

The coupling characteristics of the 4.6 GHz lower hybrid (LH) grill and the factors affecting LH wave coupling, mainly drift, with divertor configuration in anti-clockwise/clockwise magnetic field direction are studied. Based on the Langmuir probe measurement, the relationship between the reflection coefficient (RC) of LH power and the edge density at the grill mouth () measured by Langmuir probes is investigated. The optimum LH wave coupling with RC less than 2% is achieved when is in the range of –, which is quite close to the theoretically predicted value. In addition, the experimental results show that drift in the divertor scrape off layer plays a dominant role in affecting and LH wave coupling. Studies show that the discrepancy in and LH wave coupling in different magnetic field directions is mainly attributed to the different directions of drift, whereas the corresponding discrepancy with different plasma currents is mainly ascribed to the change in drift intensity. Therefore, for the upper single null divertor configuration, a lower/higher plasma currents is more favorable for increasing and improving LH wave coupling in anti-clockwise/clockwise . The above studies are of great significance for optimizing and improving LH wave coupling in the extension of the plasma operation regime.

Conceptual design of a collimated neutron flux monitor and spectrometer for DTT

A conceptual design and performance studies for a collimated neutron flux monitor and neutron spectrometer for the Divertor Tokamak Test (DTT) facility are presented. This study is based on the single-null divertor configuration and for “Half Power” and “Full power” scenarios with 15 MW of negative-ion NBI, 29 MW of ECH and 3 MW of ICRF heating with a maximum neutron yield of 1.5×1017 s−1. Fast ion distributions (both from auxiliary heating systems and fusion born) have been simulated in TRANSP/NUBEAM and the corresponding neutron energy spectra have been calculated using DRESS. Synthetic diagnostics have been implemented to determine the neutron fluxes and spectra at the detector location. Neutron emissivity profiles, plasma position, core ion temperature and the ratio of thermal and non-thermal D ion populations can be obtained with good accuracy and time resolution.

Simulations of divertor heat flux width using transport code with cross-field drifts under the BOUT++ framework

The fluid transport code [trans-electric field (Er) module] under the BOUT++ framework has been used to simulate divertor heat flux width and boundary Er with all drifts and the sheath potential in the scrape-off layer. The calculated steady state radial Er in the pedestal region has been compared with that of experimental measurements from the Alcator C-Mod tokamak. The magnitude and shape of Er are similar to those of the experimental data. In order to understand the relative role of cross-field drifts vs turbulent transport in setting the heat flux width, four C-Mod enhanced Dα H-mode discharges with a lower single null divertor configuration should be simulated. BOUT++ transport simulations with cross-field drifts included yield similar heat flux width λq to that of experimental measurements (within a factor of 2) from both the probe and the surface thermocouple diagnostics and show a similar trend with plasma current to that of the Eich experimental scaling. The simulations show that both drifts and turbulent transport compete to determine the heat flux width. The magnetic drifts play a dominant role in setting the divertor heat-flux width, while the E × B drift decreases the heat flux width by 10%–25%, leading to improved agreement with the experiment relative to Goldston’s model. A turbulence diffusivity scan (χ⊥) identifies two distinct regimes: a drift dominant regime when χ⊥ is small and a turbulence dominant regime when χ⊥ is large. The Goldston heuristic drift model yields a lower limit of the width λq.

Study of Temperature and Heat Flux on the EAST Divertor Target Plate in LHW+ NBI/ICRH H-Mode

The heat fluxes on the lower outer divertor plate were calculated using the DFLUX code with the surface temperature measured by the IR camera on the experimental advanced superconducting tokamak. The analyzed discharges were lower single-null divertor configuration discharges with different auxiliary heating: lower hybrid wave current drive (LHCD) combined with neutral beam injection (NBI) and LHCD combined with ion cyclotron resonant heating (ICRH). The surface temperature of the lower outer divertor target plate was about 170 °C in the LHCD + NBI H-mode, with the high temperature mostly distributed in the location near the joint of the dome and the outer target plate. The temperature increased as the NBI was turned on and fluctuated due to edge localized modes (ELMs). The heat flux reached nearly 4 MW/m2 when ELMs appeared and declined to less than 1 MW/m2 when the NBI was turned off. Temperature distribution with extreme toroidal asymmetry was observed in the selected LHCD + ICRH H-mode discharge. The surface temperature on the lower outer divertor target plate was up to 220 °C in one region with broader temperature distribution. During the LHCD + ICRH H-mode, the peak heat flux was not seen during the period of ICRH injection, though a brief spike in heat flux was observed at the beginning of the ICRH injection. The full-width at half-maximum of the heat flux profile as a function of divertor target plate location of the LHCD + ICRH H-mode is three times that of the LHCD + NBI H-mode.

Simulation Study of Large Power Handling in the Divertor for CFETR Phase II

The Chinese Fusion Engineering Testing Reactor (CFETR) is a next fusion device proposed in China. The ability to exhaust the plasma power loss is a critical issue for the successful production of a fusion power reactor. For Phase II of CFETR with fusion power ~1 GW, the extremely large divertor heat load would be a severe challenge to achieve steady-state operation. In this paper, we have simulated the divertor performance by using SOLPS5.0 (B2.5-EIRENE) code package for a lower single null divertor configuration based on a core self-consistent steady-state scenario with Ar seeding. The core self-consistent steady-state scenario is calculated by a 1.5-D modeling workflow utilizing the One Modeling Framework for Integrated Tasks (OMFIT) framework. The modeling shows that Ar puffing from the lower divertor region together with D2 puffing from upstream is highly effective in mitigation of the divertor peak heat flux and temperature for the whole outer target and near strike point on the inner divertor target, however, electron and ion temperature ( $T_{e}$ and $T_{i})$ at far scrape-off layer on inner divertor is still higher than the erosion mitigation. In addition, comparison simulation of different divertor geometries has been performed in this paper. Further work on self-consistent core-edge coupling to estimate the effect of impurity seeding on core performance as well as the optimization of divertor geometry and operations will be studied.

Modeling of combined effects of divertor closure and advanced magnetic configuration on detachment in DIII-D by SOLPS

One of the major challenges facing the design and operation of next-step high-power steady-state fusion devices is to develop a divertor solution for handling power exhaust, while ensuring acceptable divertor target plate erosion, which necessitates access to divertor detachment at relative low main plasma densities compatible with current drive and high plasma confinement. Detailed modeling with SOLPS is carried out to examine the effect of divertor closure on detachment with the normal single null divertor (SD) configuration, as well as one of the advanced divertor configurations, such as x-divertor (XD) respectively. The SOLPS modeling for a high confinement plasma in DIII-D finds that increasing divertor closure with SD reduces the upstream separatrix density at the onset of detachment from to . Moreover, coupling the divertor closure with XD further promotes the onset of divertor detachment at a still lower upstream separatrix density, down to the value of , thus, showing that divertor closure and advanced magnetic configuration can work synergistically to facilitate divertor detachment.

Design of snowflake-diverted equilibria of CFETR

The Chinese Fusion Engineering Test Reactor (CFETR) represents the next generation of full superconducting fusion reactors in China. Recently, CFETR was redesigned with a larger size and will be operated in two phases. To reduce the heat flux on the target plate, a snowflake (SF) divertor configuration is proposed. In this paper we show that by adding two dedicated poloidal field (PF) coils, the SF configuration can be achieved in both phases. The equilibria were calculated by TEQ code for a range of self-inductances li3. The coil currents were calculated at some fiducial points in the flattop phase. The results indicate that the PF coil system has the ability to maintain a long flattop phase in 7.5 and 10 MA inductive scenarios for the single null divertor (SND) and SF divertor configurations. The properties of the SF configuration were also analyzed. The connection length and flux expansion of the SF divertor were both increased significantly over the SND.

Heat Flux on EAST Divertor Plate in H-mode with LHCD/LHCD+NBI**Supported by the National Natural Science Foundation of China under Grant Nos 11505290, 51576208 and 11575239, and the National Magnetic Confinement Fusion Science Program of China under Grant Nos 2013GB113004 and 2015GB102004.

Based on the surface temperature measured by the infrared camera on the experimental advanced superconducting tokamak (EAST), the heat fluxes on the lower outer divertor target plate during H-mode with the lower-hybrid wave current drive (LHCD) only and with the LHCD combined with the neutral beam injection (NBI) are calculated by the DFLUX code and compared. The analyzed discharges are lower single null divertor configuration discharges. In the case with the LHCD only (, ), ELM-free appears after L-H transition with the peak heat flux on the lower outer target plate less than 1 MW/m2. However, there is no ELM-free appearing after the L-H transition in the case with the LHCD+NBI (, ). The results show that the peak heat fluxes on the lower outer target plate in the LHCD+NBI H-mode cases are larger than those in the LHCD H-mode under the similar auxiliary heating power. This is because the heat flux profiles of the lower outer target plate as a function of plate location in ELMing with the LHCD+NBI are narrower than those with the LHCD only. The results are consistent with the results in terms of the scrape-off layer width observed in the EAST.

Preliminary study of divertor particle exhaust in the EAST superconducting tokamak

The particle exhaust of the upper tungsten and lower carbon divertors in EAST has been preliminarily studied during the 2016 experimental campaign. The density decay time during terminating gas puffing has been employed as a key parameter to evaluate the divertor particle exhaust performance. Comparative plasma discharges have been carried out on the particle exhaust performance between two toroidal field directions in the upper single null and lower single null divertor configurations. This work has enhanced the understanding of the effects of the in–out asymmetry and divertor geometry on the efficiency of the divertor particle exhaust. In addition, the sensitivity of the particle exhaust capability on different strike point locations has been analyzed. The experimental results are expected to provide important information on the future upgrade of EAST bottom divertor and facilitate the realization of longer pulse operation.

SOL effects on the pedestal structure in DIII-D dischargesaaThis manuscript has been authored by UT-Battelle, LLC under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form

Analysis with the SOLPS code suite (Schneider R. et al 1992 J. Nucl. Mater. 196 810; Wiesen S. et al 2015 J. Nucl. Mater. 463 480–4; Bonnin X. et al 2016 Plasma Fusion Res. 11 1403102) explains the differences in pedestal structure associated with different ion drift directions in DIII-D. Core transport models predict that fusion power scales roughly as the square of the pressure at the top of the pedestal, so understanding the effects that determine pedestal structure in steady-state operational scenarios is important to help project scenarios developed in DIII-D to ITER and other devices. Both experiments and modeling indicate that scrape off layer (SOL) conditions are important in optimizing the pedestal structure for high-beta steady-state scenarios. The SOLPS code is used to provide interpretive analysis of the pedestal, and SOL conditions are used to examine the nature of flows and fueling on the pedestal structure; including the effect of drifts in the fluid model. This analysis shows that flows driven by the ion drift outwards when the drift flows toward the x-point, in a single-null divertor configuration (favorable direction for reduced H-mode power threshold), and they drift inwards when the drift flow is away from the x-point (unfavorable direction). It is hypothesized that these flows decrease the density gradient of the pedestal in the favorable direction, thereby stabilizing the kinetic ballooning mode (KBM) and increasing the pedestal width. Comparisons of the pedestal structures in similarly shaped DIII-D steady-state plasmas confirm this change, showing increased density pedestal width, lower peak density and lower separatrix density with the favorable drift direction. The pedestal temperature is higher at a lower density, resulting in an increased pedestal pressure, which indicates that the increased particle flux does not significantly degrade energy confinement. Modeling cases with constant drift direction, but with a change between the more open lower divertor and more closed upper divertor, show that there is increased fueling inside a pedestal with a more open geometry. The pedestal fueling rate for both attached and detached cases is always lower when there is a more closed divertor geometry than in cases when there is a more open divertor geometry.

Scaling of energy confinement time in the Globus-M spherical tokamak

The paper is devoted to an energy confinement study at the Globus-M spherical tokamak (ST). Experiments were performed in single null divertor configuration with elongation as high as 1.8–1.9 for variable plasma current and fixed toroidal magnetic field. The confinement time (τE) dependence on density for ohmic-heated (OH) deuterium plasma is presented. It was found that τE rises linearly with plasma current in H-mode with pure ohmic heating. Pronounced electron and ion heating was achieved in discharges with neutral beam injection at a moderate density level. The dependence of τE on absorbed power was weak.