Outer Divertor Target Research Articles

Particle exhaust and recycling control by active divertor pumping in EAST

The EAST tokamak was built to carry out long pulse plasma discharges up to 1000s. To facilitate this, an active divertor pumping (ADP) system was installed under the lower outer (LO) divertor target plate (DTP) for plasma density control and impurity exhaust. The ADP system was used in the first EAST divertor physics campaign, and achieved very promising results on neutral gas particle exhaust and recycling control. The system can improve divertor screening for neutrals and reduce the ratio H/(H+D), and appears to be effective at controlling the main chamber hydrogenic recycling. The influence of ADP on lower inner (LI) and LO DTP is observed. Furthermore, we attempt to find an optimized strike point to mitigate DTP heat load during ADP operation. This article reports on these interesting results, in particular, for the lower single null (LSN) plasma configuration.

Type-I ELM power deposition profile width and temporal shape in JET

A new infra red camera (IR) for high resolution infra red studies for the outer divertor target plate in JET has been installed. Shot integrated energy balance between tile embedded thermocouples and IR based estimation of deposited energy on the outer tile gives fair agreement in the range of 80–120%. The assumptions of the temporal evolution of type-I ELMs power load as made for ITER define a lower, conservative boundary within the observed variation of the data. The broadening of the ELM induced power profiles are, in contrast to earlier results based on a lower resolution IR system at JET, found to be in the range of 1.4–4.3 when compared to the inter-ELM wetted area.

Effect of neutrals on the power decay length at the divertor target

EDGE2D modelling of JET divertor plasmas with wide variation of main plasma parameters was aimed at identification of operational domains of the applicability of the 2/7×λTe scaling for predicting the power deposition profile width at the outer divertor target plate in high recycling conditions. Contrary to expectations, under no conditions, including lowest density plasmas just entering the conduction regime, was this scaling capable of correctly predicting the width and peak value of target power deposition profiles. Poorer ion parallel heat conduction, compared to the electron channel, in combination with reduced energy convection and volumetric power losses in divertor, of which charge exchange was found to play the dominant role, resulted in large deviations from the predictions of the 2/7×λTe scaling.

ERO modelling of local deposition of injected 13C tracer at the outer divertor of JET

The 2004 tracer experiment of JET with the injection of 13CH4 into H-mode plasma at the outer divertor has been modelled with the Monte Carlo impurity transport code ERO. EDGE2D solutions for inter-ELM and ELM-peak phases were used as plasma backgrounds. Local two-dimensional (2D) deposition patterns at the vertical outer divertor target plate were obtained for comparison with post-mortem surface analyses. ERO also provides emission profiles for comparison with radially resolved spectroscopic measurements. Modelling indicates that enhanced re-erosion of deposited carbon layers is essential in explaining the amount of local deposition. Assuming negligible effective sticking of hydrocarbons, the measured local deposition of 20–34% is reproduced if re-erosion of deposits is enhanced by a factor of 2.5–7 compared to graphite erosion. If deposits are treated like the substrate, the modelled deposition is 55%. Deposition measurements at the shadowed area around injectors can be well explained by assuming negligible re-erosion but similar sticking behaviour there as on plasma-wetted surfaces.

Conceptual Design of a New Outer Divertor for C-Mod

The next generation outer divertor target proposed for C-Mod is intended to operate with edge physics behavior that is ’Demo-like’, i.e., it will be capable of operating at a bulk tile and structure temperature of 600C. The proposed design exposes a vertical cylinder covered with tungsten lamellae tiles to the divertor heat flux. Heat load variation along the height of the cylindrical target has been specified and is being considered in the tile design. The design must allow for differential radial thermal expansion of the cylindrical structure. It is intended to be toroidally continuous with a high tolerance on axisymmetry to improve alignment with the plasma and limit interactions of disruption induced currents with the toroidal field. Inductively driven axisymmetric disruption currents are calculated using electromagnetic transient simulations previously employed for RF antennas and the cryopump. Disruption-induced halo currents are expected to flow though the structure, which have proved troublesome for the old outer divertor structure. The new toroidally continuous structure will be intrinsically strong with respect to axisymmetric mechanical loads, although the support hardware will also need to be robust to resist movement during non-axisymmetric halo loads. Halo current specifications for the outer divertor have been developed, and halo current paths that minimize loading are “forced” with appropriate use of insulation and grounding straps. Radiative energy transfer to other components in the vessel makes sustained operation of the outer divertor at elevated temperatures difficult.

Simulation study for divertor design to handle huge exhaust power in the SlimCS DEMO reactor

By the SOLDOR/NEUT2D simulation for divertor design study on a compact DEMO reactor, SlimCS, we estimated the prospect of handling the huge exhaust power in the divertor. Assuming exhaust power of 500 MW and ion outflux of 0.5 × 1023 s−1 into the scrape-off-layer, the peak heat load is estimated to be 70 MW m−2 on the outer target on the initial divertor design (vertical target) with the introduction of moderate gas puff flux and Ar fraction. This value significantly exceeds the allowable level of 10 MW m−2 which is an initial design target. By installing the ‘V-shaped corner’ in the bottom of the outer divertor target, and using strong gas puffing or Ar impurity injection, the detached condition with high particle recycling and radiation loss conditions is formed, and the peak heat load is successfully reduced below 10 MW m−2. It can also be demonstrated properly for the dependence of the exhaust power on the divertor heat load. Peak heat load is reduced exponentially with a decrease in the exhaust power and reaches 7 MW m−2 at Qtotal = 300 MW for moderate gas puff flux and Ar fraction.

ELM induced divertor heat loads on TCV

Results are presented for heat loads at the TCV outer divertor target during ELMing H-mode using a fast IR camera. Benefitting from a recent surface cleaning of the entire first wall graphite armour, a comparison of the transient thermal response of freshly cleaned and untreated tile surfaces (coated with thick co-deposited layers) has been performed. The latter routinely exhibit temperature transients exceeding those of the clean ones by a factor ∼3, even if co-deposition throughout the first days of operation following the cleaning process leads to the steady regrowth of thin layers. Filaments are occasionally observed during the ELM heat flux rise phase, showing a spatial structure consistent with energy release at discrete toroidal locations in the outer midplane vicinity and with individual filaments carrying ∼1% of the total ELM energy. The temporal waveform of the ELM heat load is found to be in good agreement with the collisionless free streaming particle model.

A conceptual model of the magnetic topology and nonlinear dynamics of ELMs

A conceptual model is introduced describing the 3D magnetic topology and nonlinear evolution of Type-I edge localized modes (ELMs), which immediately follow the initial linear peeling-ballooning growth phase. The model requires feedback amplification of stable and unstable invariant manifolds that increases the helical perturbation. The amplification process is caused by a rapid growth of field-aligned helical thermoelectric currents that flow through relatively short pedestal plasma flux tubes connecting the inner and outer divertor target plates. It is shown that the model qualitatively agrees with the formation of global field-aligned emission filaments and with fast IR heat flux splitting patterns of the inner and outer strike points observed experimentally. In addition, the model predicts an increase in the size of the ELMs as the pedestal collisionality drops. Experimental data and modeling results are presented supporting the basic conceptual elements of the model.

‘Mixed-material evolution analysis of the ITER divertor’

Evolution of the ITER outer divertor target surface is analyzed for the case of wall-sputtered beryllium transported to an initially tungsten divertor. Coupled codes for the convective edge plasma, impurity transport, and mixed-surface sputtering, give the wall-source beryllium flux to the divertor, the sputtering erosion/redeposition response, and the resulting beryllium surface content/growth. The analysis shows zero net Be growth over most (∼80%) of the divertor, due to high re-sputtering and reflection—with an equilibrium Be/W surface quickly formed (∼10 s)—but with a region near the strike point having substantial (∼1 nm/s) growth.

Design study of JT-60SA divertor for high heat and particle controllability

The modification of JT-60 to a fully superconducting coil tokamak, JT-60SA (JT-60 Super Advanced) device, has been programmed to contribute and supplement ITER toward to DEMO. Lower divertor design with the ITER-like lower single null divertor configuration is studied to obtain high heat and particle controllability using SOLDOR/NEUT2D code. With anticipated total power flux into SOL of 37 MW (90% of input power), the peak heat load on outer divertor target can be reduced to 5.8 MW/m 2 at the detached condition by gas puffing in the vertical divertor target with the “V-shaped corner”. It is ∼2/5 of the allowable level of 15 MW/m 2. On the other hand, particle controllability such as control of detached to attached condition by divertor pumping is improved by increase the strike point distance from 20 to 120 mm with above divertor geometry, suggesting that recover from severe detachment at the small distance case can be achieving by elevation of the strike point locations. Optimization of upper divertor design is in progress for high β steady-state operation using upper single null divertor configuration.

Sheath heat transmission factors on TCV

Despite its importance in determining the calculated surface power flux predicted, for example, by scrape-off layer simulation codes, the sheath heat transmission factor, γ remains a quantity whose theoretically predicted value has rarely been tested experimentally. This contribution attempts such a test by comparing direct, infra-red measurements of the outer divertor target power fluxes in single null lower discharges on the TCV tokamak with tile embedded Langmuir probe data. A selection of L-mode plasmas have been used, including varying outer target flux expansion and a pair of matched D–He pulses. Particular care is taken to identify all possible sources of error and estimate their effect on the derived values of γ. Within the admittedly rather large uncertainty, L-mode values of γ are found to be consistent with simple sheath theory.

Compatibility of the radiating divertor with high performance plasmas in DIII-D

A radiating divertor approach was successfully applied to high performance ‘hybrid’ plasmas [M.R. Wade et al., in: Proceedings of the 20th IAEA Fusion Energy Conference, Vilamoura, Portugal, 2004]. Our techniques included: (1) injecting argon near the outer divertor target, (2) enhancing the plasma flow into the inner and outer divertors by a combination of particle pumping and deuterium gas puffing upstream of the divertor targets, and (3) isolating the inner divertor from the outer by a structure in the private flux region. Good hybrid conditions were maintained, as the peak heat flux at the outer divertor target was reduced by a factor of 2.5; the peak heat flux at the inner target decreased by 20%. This difference was caused by a higher concentration of argon at the outer target than at the inner target. Argon accumulation in the main plasma was modest (nAr/ne⩽0.004 on axis), although the argon profile was more peaked than the electron profile.

Deuterium depth profiling in JT-60U tiles using the D(3He, p)4He resonant nuclear reaction

Deuterium depth profiles in graphite tiles of JT-60U have been determined by means of the D(3He,p)4He nuclear reaction. Tiles with net carbon deposition showed high deuterium retention. A D/C ratio of 0.07 was found in the outer dome wing tile. In the inner divertor tile covered with a thick deposition layer, the depth profile shows a broad peak in a deep region. This indicates that deuterium trapped in the shallow region can be replaced by hydrogen with hydrogen discharges due to isotope exchange, while deuterium trapped in the deeper region can not. Deuterium in the redeposited layers can be more easily removed than that in the substrate. The deuterium concentration in the outer divertor target tile was very low (D/C<0.003). The highest D/C ratio, 0.09, was found in the upper region of the first wall.

Migration of artificially introduced micron-size carbon dust in the DIII-D divertor

Migration of pre-characterized carbon dust in a tokamak environment was studied by introduction of ∼30mg of micron-size (∼6μm median diameter) dust in the lower divertor of DIII-D using a DiMES sample holder. Direct exposure of the dust to high particle and heat fluxes in ELMing H-mode lower single-null discharges resulted in part of the dust being injected into the plasma. Following a brief exposure (∼0.1s) at the outer strike point, 1.5–2% of the total dust carbon content (2 –3×1019carbon atoms, equivalent to a few million dust particles) penetrated the core plasma, raising the core carbon density by a factor of 2–3. Individual dust particles were observed moving at velocities of 10–100m/s, predominantly in the toroidal direction for deuteron flow to the outer divertor target, consistent with the ion drag force. The observed behavior of the dust is in qualitative agreement with modeling by the 3D DustT code.

Deuterium depth profiling in JT-60U W-shaped divertor tiles by nuclear reaction analysis

Deuterium concentrations and depth profiles in plasma-facing graphite tiles used in the divertor of JAERI Tokamak-60 Upgrade (JT-60U) were investigated by nuclear reaction analysis (NRA). The highest deuterium concentration of D/ 12C of 0.053 was found in the outer dome wing tile, where the deuterium accumulated probably through the deuterium–carbon co-deposition. In the outer and inner divertor target tiles, the D/ 12C data were lower than 0.006. Additionally, the maximum (H + D)/ 12C in the dome top tile was estimated to be 0.023 from the results of NRA and secondary ion mass spectroscopy (SIMS). Orbit following Monte-Carlo (OFMC) simulation showed energetic deuterons caused by neutral beam injections (NBI) were implanted into the dome region with high heat flux. Furthermore, the surface temperature and conditions such as deposition and erosion significantly influenced the accumulation process of deuterium. The deuterium depth profile, scanning electron microscope (SEM) observation and OFMC simulation indicated the deuterium was considered to accumulate through three processes: the deuterium–carbon co-deposition, the implantation of energetic deuterons and the deuterium diffusion into the bulk.

Divertor Experiments with MBI and Strong Gas Puffing on HL-2A

In the HL-2A 2004 experiment campaign, pulsed molecular beam injection (MBI) and strong hydrogen gas puffing under the divertor configuration were used for gas fueling. The experimental results show that the MBI of hydrogen can reduce the heat flux to the divertor target plate. The electron temperature measured by the Langmuir probe array decreases significantly during the injection of the molecular beam whereas the electron density increases. This indicates that the plasma pressure near the target plates tends to be constant at a new equilibrium level. In the divertor plasmas with strong hydrogen gas puffing a high plasma density up to 4.4 × 1019 m−3 was achieved. In addition, a phenomenon similar to the partially detached divertor regime was observed, which is being studied in open divertor tokamaks such as DIII-D to reduce the peak heat flux on the target plates near the separatrix. After a strong gas puffing the electron temperature measured on the outer divertor target plate near the separatrix decreases till below 5 eV or even lower, but that of the farther outer divertor target plate does not change obviously; and the CIII and the Hα emissions at the plasma edge decrease as expected, but the Hα emission near the X-point increases. These results reflects some interesting characteristics, which needs to be studied by further modeling and experiments.

Variation of particle exhaust with changes in divertor magnetic balance

Recent experiments on DIII-D point to the importance of two factors in determining how effectively the deuterium particle inventory in a tokamak plasma can be controlled through pumping at the divertor target(s): (1) the divertor magnetic balance, i.e. the degree to which the divertor topology is single-null or double-null (DN) and (2) the direction of the of B × ∇B ion drift with respect to the X-point(s). Changes in divertor magnetic balance near the DN shape have a much stronger effect on the particle exhaust rate at the inner divertor target(s) than on the particle exhaust rate at the outer divertor target(s). The particle exhaust rate for the DN shape is strongest at the outer strike point opposite the B × ∇B ion particle drift direction. Our data suggests that the presence of B × ∇B and E × B ion particle drifts in the scrape-off layer and divertor(s) play an important role in the particle exhaust rates of DN and near-DN plasmas. Particle exhaust rates are shown to depend strongly on the edge (pedestal) density. These results have implications for particle control in ITER and other future tokamaks.

Outer divertor target deposited layers during reversed magnetic field operation in JET

Divertor surface temperatures are significantly affected by the presence of deposited surface layers. This phenomenon can be used to monitor deposited layer evolution on a shot-by-shot basis. It was found that during an experimental campaign where the B×∇B direction was reversed that the outer target, normally an erosion zone, became a deposition zone.

Variation in particle pumping due to changes in topology in high performance DIII-D plasmas

Recent experiments on DIII-D point to the importance of two factors in determining the rate at which deuterium particles can be pumped at the divertor target(s): (1) the divertor magnetic balance, i.e., the degree to which the divertor topology is single-null (SN) or double-null (DN), and (2) the direction of the of B×∇B ion drift with respect to the X-point(s). Changes in divertor magnetic balance near the DN shape have a much stronger effect on the pumping rate at the inner divertor target(s) than on the pumping rate at the outer divertor target(s). The behavior in the particle pumping observed at the inner and outer divertor target(s) in the DN and near-DN shapes suggests a redistribution of particles that would be expected in the presence of B×∇B and E×B ion particle drifts in the scrapeoff layer (SOL) and divertor(s).

Analyses of Hydrogen Isotope Distributions in the Outer Target Tile used in the W-shaped Divertor of JT-60U

Hydrogen isotope distributions of the outer divertor target CFC tile, in the W-shaped divertor of JT-60U where erosion dominated, were analyzed by secondary ion mass spectroscopy (SIMS) to study the correlation between the concentrations of hydrogen isotopes and erosion/deposition rate. The chemical states of the CFC tile surfaces were analyzed by x-ray photoelectron spectroscopy (XPS). The existence of re-deposition layers was observed by scanning electron microscope (SEM). The erosion/deposition depth was also evaluated by a surface profilometer, namely dial gauge. The hydrogen and deuterium retention profiles were compared with the tritium profile obtained by the tritium imaging plate technique (TIPT).It was found that almost all the deuterium once retained in the near surface region of the erosion dominated area was replaced by hydrogen during the final HH discharge and/or exposure to the atmosphere. The hydrogen retention profile was controlled by surface temperature and plasma flux, and the total hydrogen retention was much less than that in the re-deposition layers observed at the plasma shadowed area in this particular tile.