Divertor Operation Research Articles

Effect of a baffle divertor structure on neutral hydrogen and helium transport in the Large Helical Device

Control of the peripheral plasma density is one of urgent tasks for improving the plasma performance in the Large Helical Device. Test modules with a baffle divertor structure have been installed in the inboard side of the torus for controlling the neutral density in the plasma periphery. Neutral hydrogen and helium transport in an original open divertor and the baffle divertor configurations have been monitored with filtered CCD cameras to observe the intensity profile of visible emission by neutral hydrogen (Hα) and helium (HeI). A detailed analysis using a three-dimensional neutral particle transport simulation code reasonably explains the dependence of the observed intensity profiles on the divertor and the magnetic configurations. It also predicts that the baffle divertor significantly reduces the emission of neutral hydrogen in the plasma periphery in the inboard side, and the compression of neutral helium density by the baffle divertor is comparable to that of hydrogen.

Divertor heat and particle control experiments on the large helical device

In the Large Helical Device (LHD), studies of divertor heat and particle control have been conducted with a helical divertor magnetic structure which exists naturally in the heliotron-type magnetic configuration to explore operation scenarios in a heliotron-type fusion reactor. Reduction of the divertor heat load has been one of the focused studies in the recent LHD experiment. Impurity seeding and modification of the edge magnetic structure with a resonant magnetic perturbation field have been investigated to enhance radiation power. To control the fueled particles and impurities, the helical diveror will be closed with a baffle structure and in-vessel cryo-sorption pump in 2012. For the examination of the baffle structure, it was partially installed in LHD in 2010. Comparison of the neutral particle pressure in the divertor with the baffle structure and with the existing open divertor shows that the baffle structure efficiently compresses fueled particles in the divertor.

Advanced divertor configurations with large flux expansion

Experimental studies of the novel snowflake divertor concept (D. Ryutov, Phys. Plasmas 14 (2007) 064502) performed in the NSTX and TCV tokamaks are reviewed in this paper. The snowflake divertor enables power sharing between divertor strike points, as well as the divertor plasma-wetted area, effective connection length and divertor volumetric power loss to increase beyond those in the standard divertor, potentially reducing heat flux and plasma temperature at the target. It also enables higher magnetic shear inside the separatrix, potentially affecting pedestal MHD stability. Experimental results from NSTX and TCV confirm the predicted properties of the snowflake divertor. In the NSTX, a large spherical tokamak with a compact divertor and lithium-coated graphite plasma-facing components (PFCs), the snowflake divertor operation led to reduced core and pedestal impurity concentration, as well as re-appearance of Type I ELMs that were suppressed in standard divertor H-mode discharges. In the divertor, an otherwise inaccessible partial detachment of the outer strike point with an up to 50% increase in divertor radiation and a peak divertor heat flux reduction from 3–7MW/m2 to 0.5–1MW/m2 was achieved. Impulsive heat fluxes due to Type-I ELMs were significantly dissipated in the high magnetic flux expansion region. In the TCV, a medium-size tokamak with graphite PFCs, several advantageous snowflake divertor features (cf. the standard divertor) have been demonstrated: an unchanged L–H power threshold, enhanced stability of the peeling–ballooning modes in the pedestal region (and generally an extended second stability region), as well as an H-mode pedestal regime with reduced (×2–3) Type I ELM frequency and slightly increased (20–30%) normalized ELM energy, resulting in a favorable average energy loss comparison to the standard divertor. In the divertor, ELM power partitioning between snowflake divertor strike points was demonstrated. The NSTX and TCV experiments are providing support for the snowflake divertor as a viable solution for the outstanding tokamak plasma–material interface issues.

Scaling radiative divertor solutions to high power in DIII-D

Detached radiative divertor plasmas are examined in DIII-D as a function of power. The 2D profile of plasma electron density and temperature is reconstructed from divertor Thomson data while energy transport is deduced from power balance measurements. The combination of plasma and energy diagnostics reveals that parallel energy transport transitions to convective flow at a constant 5–10 eV as a function of power. Simple considerations would imply divertor density proportional to power density for Mach 1 flow. However, the divertor, upstream separatrix and core plasma densities increase significantly less than linear with power density. During detached divertor operation H-mode core confinement is maintained though the edge pedestal can experience moderate degradation.

The effect of the divertor operation regimes on the plasma parallel flow in the edge of a tokamak

Based on the variations of the static pressure along the magnetic field line in different divertor operation regimes, the effects of the divertor operation regimes on the plasma parallel flow at the edge of a tokamak are investigated using a one-dimensional fluid model. In low recycling regime, the variation of the static pressure along the field line is obvious from the scrape-off layer (SOL) region near the X-point, and the variation tendency is the same as that of the density. The Mach number of the plasma parallel flow increases along the magnetic field line and the variation is from gentle to sharp. In high recycling regime, the static pressure does not change much except in the near divertor plate region, as a result, the Mach number of the plasma parallel flow varies gently in the SOL region and the most of the divertor region, and it increases rapidly in the near divertor plate region. The variation of the static pressure in weak divertor detachment regime is similar to that in high recycling regime, but the static pressure shows decrease tendency in the SOL region near the X-point, consequently, the Mach number of the plasma parallel flow at X-point is larger than that in high recycling regime. In strong divertor detachment regime, static pressure decreases obviously in the SOL region and away from the divertor plate region, where the static pressure decreases rapidly, and a high Mach plasma parallel flow is observed. Static pressure decreasing while dynamic pressure increasing to keep the total pressure conservation is shown to be a possible cause of the high Mach parallel flow.

Poloidal distribution of recycling sources and core plasma fueling in DIII-D, ASDEX-Upgrade and JET L-mode plasmas

Deuterium fueling profiles across the separatrix have been calculated with the edge fluid codes UEDGE, SOLPS and EDGE2D/EIRENE for lower single null, ohmic and low-confinement plasmas in DIII-D, ASDEX Upgrade and JET. The fueling profiles generally peak near the divertor x-point, and broader profiles are predicted for the open divertor geometry and horizontal targets in DIII-D than for the more closed geometries and vertical targets in AUG and JET. Significant fueling from the low-field side midplane may also occur when assuming strong radial ion transport in the far scrape-off layer. The dependence of the fueling profiles on upstream density is investigated for all three devices, and between the different codes for a single device. The validity of the predictions is assessed for the DIII-D configuration by comparing the measured ion current to the main chamber walls at the low-field side and divertor targets, and deuterium emission profiles across the divertor legs, and the high-field and low-field side midplane regions to those calculated by UEDGE and SOLPS.

Recent progress in RF heating and long-pulse experiments on EAST

New capabilities including radio-frequency (RF) powers and diagnostics have been developed since the last IAEA meeting. Divertor performance has been systematically assessed for both single null and double null configurations with normal and reversed BT directions to address asymmetries correlated with classic drift and the issues of heat load on divertor plates. 1 MW lower hybrid wave power injection has produced fully non-inductive diverted plasma discharges. Such discharges have been readily extended to well over 1 min duration. With the assistance of lower hybrid current drive (LHCD), EAST can be operated at 1 MA with both RTEFIT/Iso-flux or pre-programmed control algorithms. A later operation mode could produce an improved confinement plasma by increasing the ohmic power through regulating the loop voltage. Both direct electron and ion heating by ion cyclotron resonant frequency power have been observed for the first time on EAST. Intrinsic toroidal rotation and momentum study in ohmic and LHCD plasmas reveals the importance of the balance between neoclassical viscosity and neutral friction at the edge and finds clear linkage between edge and core toroidal rotation. The expected capabilities on EAST in the next two years will allow access to high-performance regimes, such as H-modes, and enable specific physics studies.

Concept design of the cassette toroidal mover

A full scale physical development and test facility, Divertor Test Platform 2 (DTP2), has been established in Finland for the purpose of demonstrating and developing the remote handling (RH) equipment designs for ITER using prototypes and virtual models. The major objective of the DTP2 environment is to verify and develop ITER divertor RH devices and operations. In practice this means various test trials and measurements of performance characteristics. This paper describes the design process of the Cassette Toroidal Mover (CTM). The main purpose of this design task was the development of the CTM concept. The goal of the design process was to achieve compatibility between CTM and the latest ITER divertor design. The design process was based on using a variety of tools, i.e. Catia V5, Delmia, Ansys, Mathcad and project management tools. Applicable European Standards were applied to the concept design. CTM is the cassette transporter, which carries divertor cassettes on the toroidal rails inside the ITER Vacuum Vessel (VV) during the divertor maintenance. The operation environment differs from a common industrial environment. Radiation level is 100 Gy/h. The temperature during RH operations can be 50 °C. Clearances are less than 20 mm and the loads carried weigh 9000 kg. These conditions require special solutions during the product development process. The design process consisted of defining and developing of the CTM operational sequence. This sequence includes the procedure of how the CTM – with it is onboard manipulator – prepares for and handles the divertor cassettes during RH operations. RH operations are essential part when defining CTM functions. High reliability is required in order to carry out RH tasks successfully. The recoverability of CTM is also an important design criteria. This paper describes the design process and the structure of the CTM concept.

Predictive modelling for EAST divertor operation

The predictive modelling study of the divertor operation in EAST tokamak [B. Wan et al., Nucl. Fusion 49, 104011 (2009)] with double null (DN) configuration is carried out by using the two-dimensional edge plasma code B2.5-SOLPS5.0 [D. P. Coster, X. Bonnin et al., J. Nucl. Mater. 337–339, 366 (2005)]. The modelling study includes the particle and power balance in the scrape-off-layer (SOL), the operation parameters of plasma density, temperature and plasma heat fluxes at the separatrix, the target plates and the wall, and the effect of the gas puffing, drifts, and vertical target plate on the divertor operation. The fluid model for the edge plasma is applied using the real magnetohydrodynamic (MHD) equilibrium from the MHD equilibrium code EFIT [L. L. Lao et al., Nucl. Fusion 25, 1611 (1985)] and the real divertor geometry in the device. Before EAST tokamak starts its experimental programme of divertor operation, the modelling plays an important role in the design of its experimental programme and the optimization of the divertor operation parameters. Based on the modelling results, EAST divertor can operate over a large wide of plasma parameters with different regimes. For a heating power of 8 MW and an edge density at core-SOL interface Nedge = 0.8 × 10191/m3 and Nedge = 1.3 × 10191/m3, the EAST divertor begins access to the high recycling operation regime at the outer and inner target plates, respectively, where the plasma temperature and the heat fluxes at the target plates decrease. The gas puffing can increase the plasma density at the separatrix and trigger the transition from the high recycling operation into detachment at the target plates. When E × B and B × ▿B drifts are taken into account, the asymmetry of plasma parameters and heat fluxes between up-down SOLs can be found. The vertical target plate in EAST divertor can reduce the peak values of heat fluxes at the target plate and enables detachment at lower plasma density. The divertor with the vertical target plate can be considered as a possible option for EAST divertor upgrade in the future.

Development of KSTAR in-vessel components and heating systems

In-vessel components of the Korea Superconducting Tokamak Advanced Research (KSTAR) were developed for 2010 campaign to provide a crucial circumstance for achieving the strongly shaped and diverted plasma. Moreover, the in-vessel components such as limiter, divertor, passive stabilizer, in-vessel control coil (IVCC) system demonstrated good performances satisfying the original design concepts. In addition to the plasma facing components and the IVCC, in-vessel cryo-pump (IVCP) system was also installed to leverage divertor operation. Besides the in-vessel components, there have been substantial progresses in development of the heating and current drive system. The KSTAR heating and current drive system includes all kinds of the major heating systems such as neutral beam injection (NBI), ion cyclotron range of frequency (ICRF), electron cyclotron resonance heating and current drive (ECH and ECCD), lower hybrid current drive (LHCD) systems. As an initial stage for full equipment of the heating systems to total power of 26 MW, several key systems such as 1st NBI (called NBI-1), ICRF, and ECH-assisted startup system successfully demonstrated their excellent feasibilities in the design and performances for dedication to the 2010 campaign.

Tritium control techniques in ITER by ammonia injection

A technique based on ammonia injection in the afterglow region of the divertor plasma is proposed for the inhibition of tritium/carbon co-deposits in remote areas of ITER when operated with carbon-based divertor targets. Supporting experiments in the PILOT-PSI divertor simulator plasma device and in a RF methane plasma with afterglow ammonia injection are described. The potential of ammonia glow discharges for the removal of carbon layer in easy to reach areas and in gaps is also addressed, thus opening the possibility of an integrated tritium control scenario with water-free chemical products in carbon-based divertor operation during the active phase of ITER.

First study of EAST divertor by impurities puffing

A series of experiments has recently been carried out in EAST under different plasma conditions to investigate the basic divertor performance, divertor and SOL screening efficiency and radiative divertor effect. Detached divertor plasmas have been achieved by density ramp-up. It is found that CIII emission from the low-field side (LFS) exhibits a strong dependence on poloidal locations and plasma operation regimes from methane (CH4) puffing experiments. In addition, the radiative divertor experiments by injection of mixed Ar (5.7% Ar in D2) into the outer divertor chamber reduce the peak heat flux by 50% at the outer target plate, which also reduce the divertor plasma temperature. The in–out heat flux distribution asymmetry is improved.

Recent progress on divertor operations in EAST

The first systematic assessment of divertor performance was made in the EAST superconducting tokamak for both single null (SN) and double-null (DN) divertor configurations under Ohmic and L-mode plasma conditions. Particle and heat fluxes are higher at outboard divertor targets than those at inboard divertor targets for both SN and DN. However, the DN divertor configuration exhibits a much stronger in-out asymmetry favoring outer divertors. DN operation also leads to an up-down asymmetry with higher particle fluxes to the divertors with their ∇B drift toward the X-point. Reversing toroidal field direction shows a strong influence on the divertor asymmetries. The plasma exhibits a large amplitude broadband turbulence in the outer divertors for both SN and DN, which is significantly reduced in the inner divertors, especially for DN. This coincides with a broad divertor SOL profile on the outboard side. Initial comparisons are also made with SOLPS-B2/EIRENE modeling.

The relation between upstream density and temperature widths in the scrape-off layer and the power width in an attached divertor

The target power width is one of the most critical practical quantities in the development of magnetic fusion energy. It is essential to know how to scale this quantity to future devices. At present the controlling physics is not adequately understood, making reliable prediction difficult. It seems likely that two important processes effecting are (a) cross-field transport, e.g. D⊥, and (b) volumetric power loss processes in the edge plasma, with the latter tending to occur mainly in the divertor for attached divertor conditions. It is hypothesized that a simple relation exists between the ‘upstream’ radial profiles of ne and Te in the main scrape-off layer, , , and the parallel power flux density at the divertor entrance, . Such a simple relation is found here in 2D SOLPS edge code simulations of attached divertor conditions, which contain a wide range of more or less complex edge physics effects. It is found that , as can be expected on the basis of flux-limited parallel heat conduction, rather than Spitzer–Harm conduction for which is expected. For the relatively open divertor configuration considered, and for attached divertor conditions, it is found that the flux-limited relationship also holds for the SOLPS power flux density deposited on the target , even including the radiation load; this despite the fact that up to half the power into the SOL is dissipated radiatively. Comparing with experimentally measured target power widths for H-mode discharges, better agreement is found assuming flux limited rather than Spitzer–Harm transport although definitive conclusions will require analysis of specific discharges in specific tokamaks. This study is a necessary preliminary work to an equivalent treatment of the case where volumetric losses in the divertor are stronger, including the detached, strongly radiating divertor case with momentum loss.

Design and installation of the closed helical divertor in LHD

Neutral particle behavior in the Large Helical Device heliotron has been investigated to conduct the effective particle control using the intrinsic helical divertor. It was revealed that the torus in–out asymmetry observed in the neutral pressure distribution depended on the divertor particle flux distribution, and thus, on the operational magnetic configuration. It was also revealed that the neutral pressure in the vacuum vessel in LHD was below 0.1 Pa. Degradation of the plasma confinement with increasing of the neutral pressure was observed, and that suggested the effective particle control is necessary for the sustaining of long discharge with high performance plasma and the further improvement of the confinement. The modification of the open helical divertor to the closed one was investigated for the particle control using helical divertor by using EIRENE code. Results of the calculation showed that proper rearrangement of divertor plates and additional components, such as dome structure make the neutral particles to be compressed well in the divertor region, and effective divertor pumping to be possible. Based on the simulation and experimental results, design of the closed helical divertor was completed and it will be partially installed in the Large Helical Device before the experimental campaign in 2010.

He-cooled divertor for DEMO: Status of development and HHF tests

A modular He-cooled divertor concept for DEMO has been pursued at Karlsruhe Institute of Technology (KIT) in cooperation with Efremov with the goal of reaching a DEMO-relevant heat flux of 10 MW/m 2 and 1000 load cycles. The reference design is based on the use of small cooling finger modules made of tungsten and tungsten alloy and on impingement cooling with helium. For the purpose of experimental verification of the design, a combined helium loop and electron beam facility was built at Efremov. Technological studies on fabrication of tungsten divertor mock-ups were carried out. Results of the first HHF test series 2006 confirmed already the divertor performance of 10 MW/m 2. After some improvements of mock-ups the results of the following second and third test series in 2007–2008 showed noticeable enhancement in performance and resistance against thermal cyclic loadings. The fourth test series was continued in early 2010 after an upgrade of the EB facility. Six surviving mock-ups from the last test series were further tested. They survived differently between 180 and 1100 cycles in total under the maximum heat load of at least 10 MW/m 2 before failure. Thus the first breakthrough is reached.

Fuel Retention in LHD

A global particle balance study has been investigated in the Large Helical Device (LHD) in which the first wall and the divertor tiles are made of stainless steel (SUS-316L) and carbon, respectively. The carbon area is less than 10% of the stainless steel area. The analyzed discharges have been conducted under an intrinsic helical divertor (HD) or a local island divertor (LID). The HD is an open divertor at this stage, and the LID is a closed divertor equipped with a baffle structure and a pump system. In the HD configuration, fuel retention up to 75% of injected hydrogen was observed, and the retained hydrogen affected the plasma density control. On the other hand, almost all fueled hydrogen was evacuated by the pumps in the LID configuration. After each experimental campaign, detailed analyses of the in-vessel material probes (SUS-316L stainless steel) and a divertor tile exposed to various plasma discharges during each experimental campaign were conducted. The areal density of the retained hydrogen both in the material probes and the divertor tile was in the range 1021 to 1022 H/m2, and it corresponded to the averaged areal density that was observed after an experimental day with high-density discharges.

Effects of Net Currents on the Magnetic Configuration of W7‐X

AbstractConfigurations deviating from the optimum high‐mirror configuration of W7‐X may show significant bootstrap currents being detrimental for proper island divertor operation. Two basic scenarios to reconcile this are investigated with respect to equilibrium, stability and boundary structures. The first scenario with freely evolving bootstrap current realizes proper divertor operation by adjusting the vacuum magnetic field. As the bootstrap current evolves on the L/R time scale (40s) this needs long discharges. The second scenario compensates the net current by proper ECCD keeping the boundary structures basically constant sacrificing the low‐shear t ‐profile (mismatching ECCD and bootstrap current). Both scenarios seem to be viable experimental candidates. The present study needs further elaboration for more consistency between transport and equilibrium calculations (© 2010 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

Multi‐Fluid Modeling of Low‐Recycling Divertor Regimes

AbstractThe low‐recycling regimes of divertor operation in a single‐null NSTX magnetic configuration are studied using computer simulations with the edge plasma transport code UEDGE. The edge plasma transport properties pertinent to the low‐recycling regimes are demonstrated. These include the flux‐limited character of the parallel heat transport and the high plasma temperatures with the flattened profiles in the scrape‐off‐layer. It is shown that to maintain the balance of particle fluxes at the core interface the deuterium gas puffing rate should increase as the divertor recycling coefficient decreases. The radial profiles of the heat load to the outer divertor plate, the upstream radial plasma profiles, and the effects of the cross‐field plasma transport in the low‐recycling regimes are discussed. It is also shown that recycling of lithium impurities evaporating from the divertor plate at high surface temperatures can reverse the low‐recycling divertor operational regime to the high‐recycling one and may cause thermal instability of the divertor plate (© 2010 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

A Self-Consistent Analysis of Impurity Radiation by SOL-PSI Model

A model of scrape-off layer (SOL)-plasma surface interaction (PSI) coupling is presented for divertor operation. This model can treat the interrelation between impurity production and impurity radiation self-consistently. The model is based on a ‘two-point’ model and a 0-D steady state impurity particle balance model. The fraction of power radiated in SOL is calculated as a function of the line average density. Compared to the former simple coupling model, this model takes into account the wall-produced impurities that are sputtered by charge-exchange (CX) neutrals. A simple retention factor is also added in this model. A comparison with the experiments is made. The simulation results show the same trend upon the plasma density as shown in experiments. Reasonable qualitative agreement is reached between the results by using the model and those from experiments through the adjustment of the fitted factor.