Double Null Configuration Research Articles

Investigation of the effects of impurity seeding under different magnetic configurations in L-mode plasma in EAST tokamak

Experiments with a small amount of Ar puffed from the lower outer target into L-mode plasma in double null (DN), lower single null (LSN), and upper single null (USN) configurations studied the effects of magnetic configurations and the seeded location relative to the X-point on the performance of Ar seeding. With Ar seeded, the peak heat flux and the integrated power load onto the target decreased most significantly in the LSN configuration, although the increase of the line integrated radiation in the main plasma was the lowest in this case and the highest in the USN configuration. The radiation fraction in the divertor rose from about 10% to 23% of the heating power in LSN, and from around 15% to 22% in DN, while it almost did not change in the USN configuration with Ar seeding. The line emission of Ar XVI increased following the puff and then started to degrade near the end of LSN and USN phases. No such degradation was seen in the DN phase, meaning that some Ar dwelt in the core till the LSN phase began, which explained the peaking of radiation in the core prior to the seeding in the LSN phase. Dedicate Modeling indicates that the LSN configuration has the highest divertor radiation fraction. The highest radiation from outside the divertor is in the USN configuration. With increasing Ar seeding, the LSN configuration sees an increase of the radiation inside the divertor while in the USN configuration, it is the radiation from outside the divertor that increases most strongly. In the DN configuration, the radiation from both inside and outside the divertor increases with seeding, with the former more strongly.

Conceptual design of the COMPASS upgrade tokamak

The paper describes the conceptual design of the COMPASS Upgrade (COMPASS-U) tokamak which represents a compact, medium-size, high-magnetic-field and high-density device with a flexible set of poloidal field coils for generation of single, double null and snowflake configurations. In addition, COMPASS-U will be equipped with a closed divertor, which will be operated at high plasma and neutral density and with high optical opacity similar to future reactors. COMPASS-U is capable of addressing some of the key challenges in the field of the plasma exhaust physics, reactor-relevant edge plasma physics, advanced confinement regimes, advanced magnetic configurations and materials for next-step devices. The conceptual design of the individual COMPASS-U systems is presented with emphasis on magnets, support structure and power supply system enhancement.

Commissioning and experimental validation of SST-1 plasma facing components

Plasma facing components of SST-1 are designed to withstand an input heat load of 1.0 MW/m2. They protect vacuum vessel, auxiliary heating source i.e. RF antennas, NBI and other in-vessel diagnostic from the plasma particles and high radiative heat loads. PFC’s are positioned symmetric to mid-plane to accommodate with circular, single and double null configuration. Graphite is used as plasma facing material, back made of copper alloy and SS cooling/baking tubes are brazed on copper alloy back plates for efficient heat removal of incident heat flux. Benchmarking of PFC assembly was first carried out in prototype vacuum vessel of SST-1 to develop understanding and methodology of co-ordinate measurements. Based on such hands-on-experience, the final assembly of PFC’s in vacuum vessel of SST-1 was carried out. Initially, PFC’s are to be baked at 250 °C for wall conditioning followed with cooling for heat removal of incident heat flux during long pulse plasma operation. For this purpose, the supply and return headers are designed and installed inside the vacuum vessel in such a way that it will cater water as well as hot nitrogen gas depending up on the cycle. This paper will discuss the successful installation of PFC’s and its plasma operation respecting all design criteria.

High Field Side Lower Hybrid Current Drive Simulations for Off- axis Current Drive in DIII-D

Efficient off-axis current drive scalable to reactors is a key enabling technology for developing economical, steady state tokamak. Previous studies have focussed on high field side (HFS) launch of lower hybrid current drive (LHCD) in double null configurations in reactor grade plasmas and found improved wave penetration and high current drive efficiency with driven current profile peaked near a normalized radius, ρ, of 0.6-0.8, consistent with advanced tokamak scenarios. Further, HFS launch potentially mitigates plasma material interaction and coupling issues. For this work, we sought credible HFS LHCD scenario for DIII-D advanced tokamak discharges through utilizing advanced ray tracing and Fokker Planck simulation tools (GENRAY+CQL3D) constrained by experimental considerations. For a model and existing discharge, HFS LHCD scenarios with excellent wave penetration and current drive were identified. The LHCD is peaked off axis, ρ∼0.6-0.8, with FWHM Δρ=0.2 and driven current up to 0.37 MA/MW coupled. For HFS near mid plane launch, wave penetration is excellent and have access to single pass absorption scenarios for variety of plasmas for n|| =2.6-3.4. These DIII-D discharge simulations indicate that HFS LHCD has potential to demonstrate efficient off axis current drive and current profile control in DIII-D existing and model discharge.

Achieving temporary divertor plasma detachment with MARFE events by pellet injection in the EAST superconducting tokamak

A new pellet injection system has been equipped on the experimental advanced superconducting tokamak (EAST) in the 2012 campaign, with a pellet size of ϕ 2 mm × 2 mm, a frequency of 1 Hz–10 Hz and velocity of 150 m s−1–300 m s−1. The deuterium pellet is well-known for plasma fuelling as well as for triggering the edge localized mode (ELM). In the 2012 campaign, pellet injection experiments were successfully carried out on EAST. Temporary plasma detachment achieved by deuterium pellets has been observed in a double null (DN) divertor configuration, with multi-pellet injections at a repetition frequency of 2 Hz. The partial detachment of the outer divertors and complete detachment of the inner divertors was achieved after 35 ms of each pellet injection, which have a duration of 30–60 ms with the maximum degree of detachment (DOD) reaching 3.5 and 37, respectively. Meanwhile, the multifaceted asymmetric radiation from the edge (MARFE) phenomena was also observed at the high field side (HFS) near both the lower and upper X-points with radiation loss suddenly increased to about 15%–70%, which may be the main cause of divertor plasma detachment. The temporary detachment induced by pellet injection may act as a new way to study divertor detachment behaviors.

High-field side scrape-off layer investigation: Plasma profiles and impurity screening behavior in near-double-null configurations

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.

A Preliminary Development of the K-DEMO Divertor Concept

A preconceptual design study for the Korean fusion demonstration tokamak reactor (K-DEMO) has been started in 2012. The results of the preliminary concept study on the K-DEMO divertor are summarized in this paper. An up–down symmetric double-null configuration is selected as a primary choice for study with high elongation of 2 and the triangularity of 0.625. For the vertical maintenance of the K-DEMO in-vessel components, upper and lower divertors are subdivided into 32 toroidal modules, respectively. With maintaining its functionality, the poloidal coverage of divertor is restricted to maximize the breeding area of blanket. The plasma radiation from the core and the edge plasma and the fusion neutron wall loading onto the first walls of the in-vessel components were calculated as heat sources to the structures of the in-vessel components. The divertor targets are tilted at $\sim 10^{\circ }$ against separatrix field line to handle the conceived engineering limit of steady 10 MW/ $\text{m}^{2}$ of peak heat flux with $\sim 50$ % of core plasma radiation and $\sim 85$ % of divertor plasma radiation. Divertor adopts tungsten monoblock targets, reduced activation steel-based structures carrying the pressurized water coolant. Also, the preliminary electromagnetic force extrapolated from the ITER disruption events was considered as a mechanical load to the structures in addition to the coolant pressure. The developed concept on the target and structure is supported by preliminary thermohydraulic and structural analyses using ANSYS, which show that the temperature and the stresses are within their allowables.

A Scoping Study for High-Field-Side Launch of Lower Hybrid Waves on ADX MIT

Launching lower hybrid (LH) waves from the high field side (HFS) of a tokamak offers significant advantages over low-field-side (LFS) launch with respect to both wave physics and plasma material interactions (PM!s). The higher magnetic field opens the window between wave accessibility and the condition for strong electron Landau damping, allowing LH waves from the HFS to penetrate into the core of burning plasma, while waves launched from the LFS are restricted to the periphery of the plasma. The lower parallel refractive index (n||) of the waves launched from the HFS yields a higher current drive efficiency as well. The absence of turbulent heat and particle fluxes on the HFS, particularly in double null configuration, makes it the ideal location to minimize PM! damage to the antenna structure. The quiescent scrape off layer (SOL) also eliminates the need to couple LH waves across a long distance to the separatrix, as the antenna can be located close to plasma without risking damage to the structure. The Advanced Divertor eXperiment (ADX) will include an LH launcher located on the HFS. The LH system for ADX will make use of the existing infrastructure from Alcator C-Mod, including sixteen 250-kW klystrons at 4.6 GHz (total source power of 4 MW), high-voltage power supply, and controls. The ADX vacuum vessel design includes dedicated space for waveguide runs, pressure windows, and vacuum feedthroughs for accessing the HFS wall. Compact antenna designs based on proven technologies (e.g., multijunction and four-way splitter antennas) fit within the available space on the HFS of the ADX. Wave coupling simulations of these launchers with HFS SOL density profiles showing good coupling can be obtained by adjusting the distance between the separatrix and the HFS wall. Guard limiters on each side of the LH antenna protect the structure during ramp-up, ramp-down, and off-normal events.

Advanced divertor concept design and analysis for HL-2M

HL-2M is a tokamak device that is under construction and will be put into operation in the near future. Based on the magnetic coil design of HL-2M, standard divertor, snowflake divertor and tripod divertor configurations have been designed. The potential properties of snowflake divertor configurations have been analyzed, such as the low poloidal field (Bp) area around the X-point, the connection length, target plate and magnetic field shear. The linear peeling-ballooning (P-B) mode is studied by BOUT++ code for snowflake divertor configurations. According to the divertor configuration properties of HL-2M, asymmetric target plates have been concept designed to be compatible with the intended single null (SN) divertor configurations as well as double null (DN) divertor configurations. The SOLPS5.0 code is used to predict the details of the divertor plasma under the conditions of the divertor configurations noted above without impurities. This result shows that the peak heat load on an outer target plate of the advanced divertor is about 40% of that of the standard divertor. But more power will be transported to the inner target plate of advanced divertor, and this will cause a higher peak heat load on the inner target plate. The advanced divertor will also have to work under low plasma recycling conditions with high particle temperature and low density in an open divertor target geometry. When the SN configuration changes to a DN tripod divertor configuration, most of the power exhaust is handled by the outer divertor target plates and the peak heat load on these is about 4.1MW/m2 (with a power exhaust of 20MW). This range of optimized divertor configurations and target geometry will enable the study of advanced divertor physics and high performance plasmas in HL-2M tokamak.

Study on the L–H transition power threshold with RF heating and lithium-wall coating on EAST

The power threshold for low (L) to high (H) confinement mode transition achieved by radio-frequency (RF) heating and lithium-wall coating is investigated experimentally on EAST for two sets of walls: an all carbon wall (C) and molybdenum chamber and a carbon divertor (Mo/C). For both sets of walls, a minimum power threshold Pthr of ~0.6 MW was found when the EAST operates in a double null (DN) divertor configuration with intensive lithium-wall coating. When operating in upper single null (USN) or lower single null (LSN), the power threshold depends on the ion ∇B drift direction. The low density dependence of the L–H power threshold, namely an increase below a minimum density, was identified in the Mo/C wall for the first time. For the C wall only the single-step L–H transition with limited injection power is observed whereas also the so-called dithering L–H transition is observed in the Mo/C wall. The dithering behaves distinctively in a USN, DN and LSN configuration, suggesting the divertor pumping capability is an important ingredient in this transition since the internal cryopump is located underneath the lower divertor. Depending on the chosen divertor configuration, the power across the separatrix Ploss increases with neutral density near the lower X-point in EAST with the Mo/C wall, consistent with previous results in the C wall (Xu et al 2011 Nucl. Fusion 51 072001). These findings suggest that the edge neutral density, the ion ∇B drift as well as the divertor pumping capability play important roles in the L–H power threshold and transition behaviour.

Vacuum modeling of three-dimensional magnetic field topology under resonant magnetic perturbations on EAST

A numerical model using field line tracing for modeling of three-dimensional magnetic field topology under resonant magnetic perturbations (RMPs) on experimental advanced superconducting tokamak (EAST) is presented. The topological structure is calculated in the vacuum paradigm. The modeling result predicts that the possible strike point splitting on a plasma-facing component and the lobes-like structure on the boundary are observable in the diagnostics at different locations. It is shown that the magnetic perturbations with a resonant dominant spectrum can induce a large footprint splitting effect as well as a wide stochastic layer. This is useful for observations using diagnostics with limited spatial resolution. The impact of RMP fields on marginally disconnected double null configurations is investigated. To avoid the transient heat load on the upper divertor or plasma-facing components near the upper x-point, it is necessary to keep the distance between two separatrices of a near double null configuration larger than a threshold value that depends on the RMP strength and the equilibrium properties. A preliminary RMP experiment on EAST shows that there is a good agreement between the splitting width predicted by the code and that of the particle flux measured by divertor probes. An enhancement of particle flux on the upper divertor during the RMP phase is observed in the lower single null discharge.

Numerical investigation of disruption characteristics for the snowflake divertor configuration in HL-2M

Cold and hot vertical displacement events (VDEs) are frequently related to the disruption of vertically-elongated tokamaks. The weak poloidal magnetic field around the null-points of a snowflake divertor configuration may influence the vertical displacement process. In this paper, the major disruption with a cold VDE and the vertical disruption in the HL-2M tokamak are investigated by the DINA code. In order to better illustrate the effect from the weak poloidal field, a double-null snowflake configuration is compared with the standard divertor (SD) configuration under the same plasma parameters. Computational results show that the weak poloidal magnetic field can be partly beneficial for mitigating the vertical instability of the plasma under small perturbations. For major disruption, the peak poloidal halo current fraction is almost the same between the snowflake and the SD configurations. However, this fraction becomes much larger for the snowflake in the event of a hot VDE. Furthermore, during the disruption for a snowflake configuration, the distribution of electromagnetic force on a vacuum vessel gets more non-uniform during the current quench.

Effects of discharge operation regimes and magnetic field geometry on the in-out divertor asymmetry in EAST

This paper aims to investigate the reason why the divertor in-out asymmetry was not obvious as experimentally observed in EAST only considering the classical drifts from previous simulations (Guo et al., J. Nucl. Mater. 438 (2013) 280; Du et al., J. Nucl. Mater. 463 (2015) 485). With consideration of the classical drifts, a series of different typical discharge scenarios in EAST in different magnetic field geometries were simulated by using the SOLPS5.2 code package. The simulated results reveal that the classical drifts make a major contribution to the in-out divertor asymmetry in the high recycling regime (HRR) and partial detachment (one divertor target begins to detach, while the other divertor remains attached) regime. In comparison, in low recycling regime the classical drifts play a much smaller role in the contributions to the in-out divertor asymmetry, which can explain reasonably the reason for it in Guo et al. (J. Nucl. Mater. 438 (2013) 280). In addition, the magnetic field geometry also has a great impact on the classical drifts inducing the asymmetry; it is found that for lower single-null, upper single-null and connected double-null topologies, in HRR the classical drifts play an dominant role in the contribution to the in-out divertor asymmetry, while for a disconnected double null magnetic field configuration, they play a minor role, which is the reason why the in-out asymmetry was unobvious by considering the drifts in Du et al. (J. Nucl. Mater. 463 (2015) 485).

Observation of Central Toroidal Rotation Induced by ICRF on EAST**supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB112004 and 2015GB103002), National Natural Science Foundation of China (Nos. 11175208, 11305212, 11375235, 11405212 and 11261140328), the Innovative Program of Development Foundation of Hefei Center for Physical Science and Technology (2014FXCX003) and Brain Korea 21 Program for Leading Universities

Core plasma rotation of both L-mode and H-mode discharges with ion cyclotron range of frequency (ICRF) minority heating (MH) scheme was measured with a tangential X-ray imaging crystal spectrometer on EAST (Experimental Advanced Superconducting Tokamak). Co-current central impurity toroidal rotation change was observed in ICRF-heated L- and H-mode plasmas. Rotation increment as high as 30 km/s was generated at ∼1.7 MW ICRF power. Scaling results showed similar trend as the Rice scaling but with significant scattering, especially in L-mode plasmas. We varied the plasma current, toroidal field and magnetic configuration individually to study their effect on L-mode plasma rotation, while keeping the other major plasma parameters and heating unchanged during the scanning. It was found that larger plasma current could induce plasma rotation more efficiently. A scan of the toroidal magnetic field indicated that the largest rotation was obtained for on-axis ICRF heating. A comparison between lower-single-null (LSN) and double-null (DN) configurations showed that LSN discharges rendered a larger rotation change for the same power input and plasma parameters.

Numerical modeling of the transition from low to high confinement in magnetically confined plasma

The transition dynamics from low (L) to high (H) mode confinement in magnetically confined plasmas is investigated using a four-field drift fluid model—HESEL (Hot Edge-Sol-Electrostatic). The model includes profile evolution and is solved in a 2D domain at the out-board mid-plane of a tokamak including both open and closed field lines. The results reveal different types of L–H-like transitions in response to ramping up the input power by increasing the ion temperature in the edge region. For a fast rising input power we obtain an abrupt transition, and for a slow rising power we obtain a L–I–H transition with an intermediate I-phase displaying limit-cycle oscillations (LCO). The model recovers the power threshold for the L–H transition, the scaling of the threshold with the density and with the loss-rate in the SOL, indicating a decrease in power threshold when switching from single to double null configuration. The results hold promises for developing full predictive modeling of the L–H transition, which is an essential step in understanding and optimizing fusion devices.

Simulation of transition dynamics to high confinement in fusion plasmas

The transition dynamics from the low (L) to the high (H) confinement mode in magnetically confined plasmas is investigated using a first-principles four-field fluid model. Numerical results are in agreement with measurements from the Experimental Advanced Superconducting Tokamak – EAST. Particularly, the slow transition with an intermediate dithering phase is well reproduced at proper parameters. The model recovers the power threshold for the L–H transition as well as the decrease in power threshold switching from single to double null configuration observed experimentally. The results are highly relevant for developing predictive models of the transition, essential for understanding and optimizing future fusion power reactors.

The magnet system of the Tokamak T-15 upgrade

Presently, the Tokamak T-15 is being upgraded. The magnet system of the Tokamak T-15 upgrade will obtain and confine the hot plasma in the divertor configuration. Plasma parameters are a major radius of 1.48m, a minor radius of 0.67m, an elongation of 1.7–1.9 and a triangularity of 0.3–0.4. The magnet system includes the toroidal winding and the poloidal magnet system. The poloidal magnet system generates the divertor with single null and double null magnetic configurations. The power supply system provides the necessary current scenarios in the windings of the magnet system. All elements of the magnet system will be manufactured by the end of 2015. The Tokamak T-15 upgrade should begin operations in 2016.

Study of Heat Load on EAST Divertor Plates by an Infrared Camera

The infrared diagnosis system is an efficient tool to provide the immediate temperature distribution on the divertor plate, which can be converted to heat flux by solving the heat conduction equation with the finite difference method. This paper concentrates on the heat flux on the divertor with different discharge conditions on EAST. Experimental results show that the heat flux on the divertor plate has a strong in–out asymmetry under different configurations. The in–out asymmetry in double null configuration is stronger than that in upper single null configuration. Meanwhile, Argon gas puffing has been introduced in EAST to control the divertor heat flux, which shows that argon injection from either the upper outer divertor or the upper dome is very effective at reducing the divertor heat flux. Whether the argon gas is introduced from the upper outer divertor or the upper dome, heat load on the inner divertor plate tends to be more easily reduced than that on the outer divertor plate, the reason for which is discussed in this paper.

The comparison of heat flux pattern on lower divertor in KSTAR

The heat flux in KSTAR is estimated for various discharge conditions by using thermocouple arrays. The heat flux on the divertor is higher than that on inboard limiter or passive stabilizer by a factor of 2. Although the plasma configuration in KSTAR has been set to a double-null configuration, the heat flux on lower divertor is higher than that on upper divertor by 3–8 times, indicating a lower-single-null-like configuration. It is observed that the operation of the in-vessel cryo-pump (IVCP) changes the heat flux pattern significantly: When the IVCP was not operated, the heat fluxes on inboard divertor (ID), central divertor (CD) and outboard divertor (OD) were similar, but when the IVCP was operated, the heat fluxes on ID and CD were increased slightly and that on OD was decreased by 2–3 times. The heat flux on divertor was decreased from 35 to 26kW/m2 with the use of the resonant magnetic perturbation (RMP), especially that on CD was decreased by 2–4 times, while that on OD is increased by 2–3 times than without RMP. For the longest H-mode pulse of 22s shot, the heat flux on lower OD was 73kW/m2, which is the maximum heat flux among the shots obtained in 2013 campaign.

Effects of drifts and ballooning instability on the divertor in–out asymmetry in EAST tokamak

The divertor in–out asymmetry (DIOA) of the disconnected double null (DDN) configuration with the active X-point at the bottom in EAST has been investigated using the SOLPS code package to explore its forming mechanism. Effects of classical drifts and ballooning-like transport (BLT) instability are evaluated individually and compared to the DIOA during H-mode. The results show that the BLT plays a much larger role in forming the DIOA. It is further found that the combined effects of the classical drifts and BLT can account for the formation of the DIOA observed experimentally in EAST. In addition, the results also imply that reversing the toroidal magnetic field may be used to mitigate the EAST DIOA under certain circumstances.