Measurements and modeling of type-I and type-II ELMs heat flux to the DIII-D divertor

Type-I and type-II edge-localized-modes (ELMs) heat flux profiles measured at the DIII-D divertor feature a peak in the vicinity of the strike-point and a plateau in the scrape-off-layer (SOL), which extends to the first wall. The plateau is present in attached and detached divertors and it is found to originate with plasma bursts upstream in the SOL. The integrated ELM heat flux is distributed at ∼65% in the peak and ∼35% in this plateau. The parallel loss model, currently used at ITER to predict power loads to the walls, is benchmarked using these results in the primary and secondary divertors with unprecedented constraints using experimental input data for ELM size, radial velocity, energy, electron temperature and density, heat flux footprints and number of filaments. The model can reproduce the experimental near-SOL peak within ∼20%, but cannot match the SOL plateau. Employing a two-component approach for the ELM radial velocity, as guided by intermittent data, the full radial heat flux profile can be well matched. The ELM-averaged radial velocity at the separatrix, which explains profile widening, increases from ∼0.2 km s−1 in attached to ∼0.8 km s−1 in detached scenarios, as the ELM filaments’ path becomes electrically disconnected from the sheath at the target. The results presented here indicate filaments fragmentation as a possible mechanism for ELM transport to the far-SOL and provide evidence on the beneficial role of detachment to mitigate ELM flux in the divertor far-SOL. However, these findings imply that wall regions far from the strike points in future machines should be designed to withstand significant heat flux, even for small-ELM regimes

Poloidal inhomogeneity of the particle fluctuation induced fluxes near of the LCFS at lower hybrid heating and improved confinement transition at the FT-2 tokamak

This paper deals with the new spectral and microturbulence experimental data and their analysis, which show, that the radial electric field Er generated at the LH heating (LHH) in the FT-2 is high enough to form the transport barriers. The ETB is formed when LHH is switched off. The radial fluctuation-induced EB drift flux densities near LCFS in SOL are measured at two different poloidal angles. For this purpose two Langmuir probes located at low and high field sides of the torus are used. Registration of the poloidal and radial components of the electric field and density fluctuations at the same time during one discharge permits to measure the poloidal asymmetry of the transport reduction mechanism of the radial and poloidal particle fluxes in the SOL. The absolute E(~) fluctuation levels show dependence on the sign of Er shear. The modification of the microscale turbulence by the poloidal Er x B rotation shear EB at the L - H transition near LCFS is also studied by X-mode fluctuation Reflectometry. The new data were obtained by spatial spectroscopic technique.Comment: 12th International Congress on Plasma Physics, 25-29 October 2004, Nice (France

Physics conclusions in support of ITER W divertor monoblock shaping

The key remaining physics design issue for the ITER tungsten (W) divertor is the question of monoblock (MB) front surface shaping in the high heat flux target areas of the actively cooled targets. Engineering tolerance specifications impose a challenging maximum radial step between toroidally adjacent MBs of 0.3 mm. Assuming optical projection of the parallel heat loads, magnetic shadowing of these edges is required if quasi-steady state melting is to be avoided under certain conditions during burning plasma operation and transiently during edge localized mode (ELM) or disruption induced power loading. An experiment on JET in 2013 designed to investigate the consequences of transient W edge melting on ITER, found significant deficits in the edge power loads expected on the basis of simple geometric arguments, throwing doubt on the understanding of edge loading at glancing field line angles. As a result, a coordinated multi-experiment and simulation effort was initiated via the International Tokamak Physics Activity (ITPA) and through ITER contracts, aimed at improving the physics basis supporting a MB shaping decision from the point of view both of edge power loading and melt dynamics. This paper reports on the outcome of this activity, concluding first that the geometrical approximation for leading edge power loading on radially misaligned poloidal leading edges is indeed valid. On this basis, the behaviour of shaped and unshaped monoblock surfaces under stationary and transient loads, with and without melting, is compared in order to examine the consequences of melting, or power overload in context of the benefit, or not, of shaping. The paper concludes that MB top surface shaping is recommended to shadow poloidal gap edges in the high heat flux areas of the ITER divertor targets

Scrape Off Layer (SOL) transport and filamentary characteristics in high density tokamak regimes

A detailed cross-device investigation on the role of filamentary dynamics in high density regimes has been performed within the EUROfusion framework comparing ASDEX Upgrade (AUG) and TCV tokamaks. Both devices have run density ramp experiments at different levels of plasma current, keeping toroidal field or q95 constant in order to disentangle the role of parallel connection length and the current. During the scan at constant toroidal field, in both devices SOL profiles tend to develop a clear Scrape Off Layer (SOL) density shoulder at lower edge density whenever current is reduced. The different current behavior is substantially reconciled in terms of edge density normalized to Greenwald fraction. During the scan at constant q95 AUG exhibits a similar behaviour whereas in TCV no signature of upstream profile modification has been observed at lower level of currents. The latter behaviour has been ascribed to the lack of target density roll-over. The relation between upstream density profile modification and detachment condition has been investigated. For both devices the relation between blob-size and SOL density e-folding length is found independent of the plasma current, with a clear increase of blob-size with edge density normalized to Greenwald fraction observed. ASDEX Upgrade has also explored the filamentary behaviour in H-Mode. The experiments on AUG focused on the role of neutrals, performing discharges with and without the cryogenic pumps, highlighting how large neutral pressure not only in the divertor but at the midplane is needed in order to develop a H-Mode SOL profile shoulder in AUG

Modification of SOL profiles and fluctuations with line-average density and divertor flux expansion in TCV

A set of Ohmic density ramp experiments addressing the role of parallel connection length in modifying scrape off layer (SOL) properties has been performed on the TCV tokamak. The parallel connection length has been modified by varying the poloidal flux expansion f x . It will be shown that this modification does not influence neither the detachment density threshold, nor the development of a flat SOL density profile which instead depends strongly on the increase of the core line average density. The modification of the SOL upstream profile, with the appearance of what is generally called a density shoulder , has been related to the properties of filamentary blobs. Blob size increases with density, without any dependence on the parallel connection length both in the near and far SOL. The increase of the density decay length, corresponding to a profile flattening, has been related to the variation of the divertor normalized collisionality ##IMG## [http://ej.iop.org/images/0029-5515/57/11/116014/nfaa7db3ieqn001.gif] {$\Lambda_{\rm div}$} (Myra et al 2006 Phys. Plasmas 13 112502, Carralero et al , ASDEX Upgrade Team, JET Contributors and EUROfusion MST1 Team 2015 Phys. Rev. Let . 115 215002), showing that in TCV the increase of ##IMG## [http://ej.iop.org/images/0029-5515/57/11/116014/nfaa7db3ieqn002.gif] {$\Lambda_{\rm div}$} is not sufficient to guarantee the SOL upstream profile flattening

Study of Z scaling of runaway electron plateau final loss energy deposition into wall of DIII-D

Controlled runaway electron (RE) plateau-wall strikes with different initial impurity levels are used to study the effect of background plasma ion charge Z (resistivity) on RE-wall loss dynamics. It is found that Joule heating (magnetic to kinetic energy conversion) during the final loss does not go up monotonically with increasing Z but peaks at intermediate Z similar to 6. Joule heating and overall time scales of the RE final loss are found to be reasonably well-described by a basic 0D coupled-circuit model, with only the loss time as a free parameter. This loss time is found to be fairly well correlated with the avalanche time, possibly suggesting that the RE final loss rate is limited by the avalanche rate. First attempts at measuring total energy deposition to the vessel walls by REs during the final loss are made. At higher plasma impurity levels Z > 5, energy deposition to the wall appears to be consistent with modeling, at least within the large uncertainties of the measurement. At low impurity levels Z < 5, however, local energy deposition appears around 5-20x less than expected, suggesting that the RE energy dissipation at low Z is not fully understood. Published by AIP Publishing.This work was supported in part by the U.S. Department of Energy under Nos. DE-FG02-07ER54917, DE-FC02-04ER54698, DE-AC05-00OR22725, DE-AC52-07NA27344, and DE-AC05-06OR23100 and in part by the Spanish Direccion General de Investigacion Cientifica y Tecnica under Projects ENE2012–31753 and ENE2015-66444R (MINECO/FEDERE, UE). DIII-D data shown in this paper can be obtained in digital format by following the links at https://fusion.gat.com/global/D3D_DMPPublicad

OEDGE modeling for the planned tungsten ring experiment on DIII-D

The OEDGE code is used to model tungsten erosion and transport for experiments with toroidal rings of high-Z metal tiles in the DIII-D tokamak. Such modeling is needed for both experimental and diagnostic design to have estimates of the expected core and edge tungsten density and to understand the various factors contributing to the uncertainties in these calculations. OEDGE simulations are performed using the planned experimental magnetic geometries and plasma conditions typical of both L-mode and inter-ELM H-mode discharges in DIII-D. OEDGE plasma reconstruction based on specific representative discharges for similar geometries is used to determine the plasma conditions applied to tungsten plasma impurity simulations. A new model for tungsten erosion in OEDGE was developed which imports charge-state resolved carbon impurity fluxes and impact energies from a separate OEDGE run which models the carbon production, transport and deposition for the same plasma conditions as the tungsten simulations. These values are then used to calculate the gross tungsten physical sputtering due to carbon plasma impurities which is then added to any sputtering by deuterium ions; tungsten self-sputtering is also included. The code results are found to be dependent on the following factors: divertor geometry and closure, the choice of cross-field anomalous transport coefficients, divertor plasma conditions (affecting both tungsten source strength and transport), the choice of tungsten atomic physics data used in the model (in particular ionization rate for W-atoms), and the model of the carbon flux and energy used for calculating the tungsten source due to sputtering. Core tungsten density is found to be of order 1015m−3 (excluding effects of any core transport barrier and with significant variability depending on the other factors mentioned) with density decaying into the scrape off layer. For the typical core density in the plasma conditions examined of 2 to 4 ×1019m−3, this represents a concentration on the order of 5 × 10−5