Abstract
Plasma-facing materials for next generation fusion devices, like ITER and DEMO, will be submitted to intense fluxes of light elements, notably He and H isotopes (HI). Our study focuses on tritium (T) retention on a wide range of samples: first, different types of materials were investigated to distinguish the impact of the pristine original structure on the retention, from W-coated samples to ITER-grade pure samples submitted to various annealing and manufacturing procedures, along with monocrystalline for reference. Then, He and He-D irradiated samples were studied to investigate the impact on He-damages such nano-bubbles (exposures in LHD or PSI-2) on T retention. We exposed all the samples to tritium gas-loading using a gentle technique preventing any introduction of new damage in the material. Tritium desorption is measured by Liquid Scintillation counting (LSC) at ambient and high temperatures (800 °C). The remaining T inventory is then measured by sample full dissolution and LSC. Results on T inventory on He exposed samples highlighted that in all cases, tritium desorption a gas (HT) increases significantly due to the formation of He damages. Up to 1.8 times more T can be trapped in the material through a competition of various mechanisms, but the major part of the inventory desorbs at room temperature, and so will most likely not take part to the long-term trapped inventory for safety and operational perspectives. Unfortunately, investigation of as industrial (used for the making of plasma-facing materials) highlighted a strong impact of the pre existing defects on T retention: up to 2.5 times more T is trapped in as received W compared to annealed and polish W, and desorbs only at 800 °C, meaning ideal material studies may underestimate T inventory for tokamak relevant conditions.
Highlights
The distinctive properties of tungsten (W) such as its high sputtering threshold, low induced activation and high melting point [1] drew the interest of the fusion community, making it a high-profile candidate as a first-wall material for current and future tokamaks
He content in W irradiated in various conditions Transmission Electron Microscopy (TEM) provided precious insight on W morphology changes and highlighted the formation of He bubbles in the material, but could not allow the detection of He contents in the various samples; that is why Elastic Recoil detection Analysis (ERDA) was performed on 4 He exposed samples, plus one pristine W sample for reference
All He exposed samples, no matter the conditions, exhibited a drastic increase in tritium gas (HT) desorbed at room temperature compared to the pristine reference
Summary
The distinctive properties of tungsten (W) such as its high sputtering threshold, low induced activation and high melting point [1] drew the interest of the fusion community, making it a high-profile candidate as a first-wall material for current and future tokamaks. W morphology changes triggered by He must be taken into account when considering the material hydrogen inventory [4,5] This property is crucial for the generation of fusion machines, T being radioactive: its use as the fusion reaction fuel imposes operational and safety limits for quantities trapped in the plasma-facing materials. The WHIrr (W under Helium Irradiation, cf Fig. 1) project aims at evaluating and understanding He impact on W and its consequences for the material properties [6], using various irradiation techniques and controlled irradiation temperature (200 to 800 °C) It includes in situ W samples exposures in fusion devices such as the Large Helical Device (LHD), allowing in situ irradiation conditions (in terms of energy and flux distributions) with controlled temperature.
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