I. Landman

The latest results from ELM-simulation experiments in plasma accelerators

Recent results of ELM-simulation experiments with quasi-stationary plasma accelerators (QSPAs) Kh-50 (Kharkov, Ukraine) and QSPA-T (Troitsk, Russia) as well as experiments in the pulsed plasma gun MK-200UG (Troitsk, Russia) are discussed. Primary attention in Troitsk experiments has been focused on investigating the carbon-fibre composite (CFC) and tungsten erosion mechanisms, their onset conditions and the contribution of various erosion mechanisms (including droplet splashing) to the resultant surface damage at varying plasma heat flux. The obtained results are used for validating the numerical codes PEGASUS and MEMOS developed in FZK. Crack patterns and residual stresses in tungsten targets under repetitive edge localized mode (ELM)-like plasma pulses are studied in simulation experiments with QSPA Kh-50. Statistical processing of the experimental results on crack patterns after different numbers of QSPA Kh-50 exposures as well as those on the dependence of cracking on the heat load and surface temperature is performed.

ITER transient consequences for material damage: modelling versus experiments

Carbon-fibre composite (CFC) and tungsten macrobrush armours are foreseen as PFC for the ITER divertor. In ITER the main mechanisms of metallic armour damage remain surface melting and melt motion erosion. In the case of CFC armour, due to rather different heat conductivities of CFC fibres a noticeable erosion of the PAN bundles may occur at rather small heat loads. Experiments carried out in the plasma gun facilities QSPA-T for the ITER like edge localized mode (ELM) heat load also demonstrated significant erosion of the frontal and lateral brush edges. Numerical simulations of the CFC and tungsten (W) macrobrush target damage accounting for the heat loads at the face and lateral brush edges were carried out for QSPA-T conditions using the three-dimensional (3D) code PHEMOBRID. The modelling results of CFC damage are in a good qualitative and quantitative agreement with the experiments. Estimation of the droplet splashing caused by the Kelvin–Helmholtz (KH) instability was performed.

Macroscopic erosion of divertor and first wall armour in future tokamaks

Abstract Sputtering, evaporation and macroscopic erosion determine the lifetime of the ‘in vessel’ armour materials CFC, tungsten and beryllium presently under discussion for future tokamaks. For CFC armour macroscopic erosion means brittle destruction and dust formation whereas for metallic armour melt layer erosion by melt motion and droplet splashing. Available results on macroscopic erosion from hot plasma and e-beam simulation experiments and from tokamaks are critically evaluated and a comprehensive discussion of experimental and numerical macroscopic erosion and its extrapolation to future tokamaks is given. Shielding of divertor armour materials by their own vapor exists during plasma disruptions. The evolving plasma shield protects the armour from high heat loads, absorbs the incoming energy and reradiates it volumetrically thus reducing drastically the deposited energy. As a result, vertical target erosion by vaporization turns out to be of the order of a few microns per disruption event and macroscopic erosion becomes the dominant erosion source.

Vertical target and FW erosion during off-normal events and impurity production and transport during ELMs typical for ITER-FEAT

During off-normal events besides evaporation macroscopic erosion might occur. In carbon-based materials (CBMs) macroscopic erosion means brittle destruction and dust formation and in metals melt layer motion and droplet splashing. Dust, melt flow, droplet splashing and redeposited evaporated material produce complex layers with considerable surface roughness and drastically reduced heat conductivity. Moreover flaking and easy levitation of particles might occur. Subsequent ELMs when depositing their energy into such layers might enhance impurity production because of increased heating of the hot spots. For vertical targets and for first walls (FWs) macroscopic erosion is analyzed. For a simplified hot spot scenario a first estimation on maximum tolerable ELM energy is given.

Residual stresses in tungsten under exposures with ITER ELM-like plasma loads

The residual stresses in tungsten targets after repetitive ITER ELM-like plasma pulses have been studied in recent simulation experiments with the quasi-stationary plasma accelerator (QSPA) Kh-50. The experiments were performed for targets preheated at 650 °C, which is above the ductile-to-brittle transition temperature (DBTT), and at room temperature (RT). The number of exposures was up to 400 pulses of duration 0.25 ms, and the magnitude of heat load was below and above the melting threshold.After plasma impacts, the residual stress was analyzed with the x-ray diffraction (XRD) technique using the sin2 ψ method. Symmetrical tensile stresses were measured in a thin subsurface layer. The largest stress (up to 750–850 MPa) was obtained after exposures of RT targets to plasma loads below the melting threshold. Preheating of the target resulted in some decrease of residual stress.

Plasma/Surface Interaction in ITER Tokamak Disruption Simulation Experiments

In evaluating the lifetime of plasma-facing components for the International Thermonuclear Experimental Reactor (ITER) against nonnormal high heat loads, credit is taken from the existence of a plasma shield that protects the target from excessive evaporation. Formation and physical properties of plasma shields are studied at the dual plasma gun facility, 2MK-200, under conditions simulating ITER hard disruptions and edge-localized modes (ELMs). The experimental results are used for validation of the theoretical modeling of the plasma/surface interaction. The important features of the non-local thermodynamic equilibrium plasma shield, such as temperature and density distribution, its evolution, the conversion efficiency of the energy of the plasma stream into total and soft X-ray radiation from highly ionized evaporated target material, and the energy balance in the plasma shield, are reproduced quite well. Thus, realistic modeling of ITER disruptive plasma/wall interaction is now possible. Because of the rather small target erosion in the simulation experiments, material erosion for ITER typical disruptions and ELMs cannot be evaluated from these simulation experiments. This requires additional simulation experiments with hot plasma streams of longer pulse duration and a separate numerical analysis, which can now be performed with validated theoretical models.

Experimental study of divertor plasma-facing components damage under a combination of pulsed and quasi-stationary heat loads relevant to expected transient events at ITER

This paper concerns the experimental study of damage of ITER divertor plasma-facing components (PFCs) under a combination of pulsed plasma heat loads (representative of controlled ITER type I edge-localized modes (ELMs)) and quasi-stationary heat loads (representative of the high heat flux (HHF) thermal fatigue expected during ITER normal operations and slow transient events). The PFCs tungsten armor damage under pulsed plasma exposure was driven by (i) the melt layer motion, which leads to bridges formation between neighboring tiles and (ii) the W brittle failure giving rise to a stable crack pattern on the exposed surface. The crack width reaches a saturation value that does not exceed some tens of micrometers after several hundreds of ELM-like pulses. HHF thermal fatigue tests have shown (i) a peeling-off of the re-solidified material due to its brittle failure and (ii) a significant widening (up to 10 times) of the cracks and the formation of additional cracks.

Estimation of the dust production rate from the tungsten armour after repetitive ELM-like heat loads

Experimental simulations for the erosion rate of tungsten targets under ITER edge-localized mode (ELM)-like surface heat loads of 0.75 MJ m−2 causing surface melting and of 0.45 MJ m−2 without melting have been performed in the QSPA-Kh50 plasma accelerator. Analytical considerations allow us to conclude that for both energy deposition values the erosion mechanism is solid dust ejection during surface cracking under the action of thermo-stress. Tungsten influx into the ITER containment of NW~5×1018 W per medium size ELM of 0.75 MJ m−2 and 0.25 ms time duration has been estimated. The radiation cooling power of Prad=150–300 MW due to such influx of tungsten is intolerable: it should cool the ITER core to 1 keV within a few seconds.

Experiments and modeling of droplet emission from tungsten under transient heat loads

Tungsten in the form of macrobrush is foreseen as one of the candidate materials for the ITER divertor. Melting of tungsten, the melt motion and melt splashing are expected to be the main mechanisms of damage determining the lifetime of plasma-facing components. Experimental investigations of droplet emission from the W melt layer for ELM-like heat loads were carried out at the plasma gun facility QSPA-T. Distribution functions of droplet velocity and droplet sizes for different heat loads were determined.In the paper, the main physical mechanism (the Kelvin?Helmholtz instability) of the melt splashing under the heat loads being applied at QSPA-T and those anticipated after the ITER transients is analyzed. Analytical distribution functions for droplet sizes and droplet velocities are defined. Numerical simulations demonstrated a reasonable agreement with the experimental data on the droplet sizes and droplet velocities and allowed projections of the melt splashing at ITER conditions.

Plasma shield formation and divertor plate erosion for ITER tokamak plasma disruptions

Abstract During plasma disruptions and ELMs a plasma shield from vaporized target material is formed in front of the divertor plates within a few microseconds which protects the divertor against further excessive evaporation. For calculations of plasma shield formation and material erosion the one dimensional (1 D) radiation magnetohydrodynamics (MHD) code FOREV-1 with 1 1 2 D MHD model was developed. Experimental results on plasma shield formation obtained at the 2MK-200 disruption simulation facility were used to check the adequacy of the physical models used in FOREV-1. It was demonstrated that the theoretical models adequately describe the complex physical processes during plasma shield formation. Plasma shield formation and material erosion in a tokamak is a 2 D problem requiring a 2 1 2 D MHD model. For modeling of this situation the 2 D code FOREV-2 is under development. First results are presented here.