V.S. Koidan

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.

Plasma impact on materials damaged by high-energy ions

We present a short review of experimental research carried out at the NRC Kurchatov Institute over recent years on the behavior of plasma-facing materials (PFMs) when a high level of radiation damage in plasma. Neutron-induced damage was modeled with accelerated ions (in the MeV range) and covered a 1–80 dpa interval. Irradiated carbon materials and tungsten were exposed to deuterium steady-state plasma at deuterium ion energies of 100–250 eV. The work focused on the damaging effect on erosion and on deuterium retention in irradiated materials. The influence of displacement damage was found on the erosion of carbon materials after their bombardment with C+ ions. Changes in deuterium retention were observed on tungsten damaged by 3–4 MeV helium ions. The experiments and results show the efficiency of the method for investigating plasma influence on PFMs for fusion applications taking into account the effect of accumulated radiation damage.

Study of the impact of steady-state plasma on radiation-damaged tungsten

Experimental studies of tungsten (as a candidate plasma-facing material for a fusion reactor) whose properties will degrade as a result of its contact with near-wall plasma and irradiation with neutrons are performed. The effect of a high level of radiation damages (1–100 displacements per atom) on deuterium accumulation and erosion caused by tungsten irradiation with deuterium plasma was studied. Radiation damages are obtained as a result of the irradiation of tungsten samples with high-energy ions in an accelerator (He+2, C+3, 4–10 MeV). Then the samples are exposed to steady-state deuterium plasma at the LENTA facility (National Research Centre Kurchatov Institute). The effects of the erosion of tungsten and accumulation of hydrogen isotopes in it are studied. Modification of the surface microstructure and radiation swelling is observed. The helium and deuterium concentrations were measured using the methods of nuclear elastic backscattering and elastic recoil detection analysis. An increased accumulation of deuterium in the damaged layer to a depth of about 5 μm is revealed.

Accumulation of deuterium in radiation-damaged tungsten

Experimental studies have been performed to determine the effect of high-level radiation damage on the accumulation of deuterium and erosion of tungsten samples exposed to deuterium plasma. Tungsten samples were exposed first to fast helium ions having an energy of 3–4 MeV (providing from one to ten displacements per atom) and then to deuterium plasma up to a dose of 1025 ion/m2. The effects of deformation and modification of the surface microstructure have been observed. The concentrations of helium and deuterium have been measured by the methods of elastic nuclear proton backscattering and nuclear recoil detection of helium ions. A high concentration of deuterium in the damaged layer of a tungsten sample has been measured, and helium has been detected in a layer ∼5 μm thick. The proposed method shows promise for determining the lifetime of materials used in fusion reactors and measuring the concentration of tritium accumulated in these materials.

Analysis of a Tungsten Surface Irradiated by Fast Ions and Deuterium Plasma

The effects occurring on the surface of tungsten under irradiation with fast ions with an energy in the megaelectrolvolt range and with high fluxes of hydrogen (deuterium) plasma are considered. These effects are radiation damage of the surface layer of the material, its erosion and deuterium retention in it. Irradiation with helium 4He2+ (3.2–4.0 MeV) and carbon 12C3+ (10 MeV) ions is performed using a cyclotron at the National Research Center Kurchatov Institute. The thickness of the damaged layer is 3.5–6 μm. The irradiated samples are exposed to steady-state deuterium plasma using a LENTA linear plasma facility to reach a plasma ion fluence of 1021–1022 cm−2. Tungsten erosion and modification of the structure of the damaged layer are analyzed at a plasma-ion energy of 250 eV. Deuterium retention in the damaged layer is studied by elastic recoil detection analysis. The deuterium concentration and its penetration depth into the material are measured. The data obtained for different kinds of fast ions used in the work are compared.