Tritium retention in W plasma-facing materials: Impact of the material structure and helium irradiation - Aix-Marseille Université Access content directly
Journal Articles Nuclear Materials and Energy Year : 2019

Tritium retention in W plasma-facing materials: Impact of the material structure and helium irradiation

E. Bernard
S. Garcia-Argote
  • Function : Author
C. Corr
  • Function : Author
S. Markelj
N. Yoshida
  • Function : Author

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 W samples: first, different types of W materials were investigated to distinguish the impact of the pristine original structure on the retention, from W-coated samples to ITER-grade pure W samples submitted to various annealing and manufacturing procedures, along with monocrystalline W for reference. Then, He and He-D irradiated W samples were studied to investigate the impact on He-damages such as 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 as 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 received" industrial W (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 W material studies may underestimate T inventory for tokamak relevant conditions.
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hal-02086134 , version 1 (01-04-2019)

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E. Bernard, R. Sakamoto, E. Hodille, A. Kreter, E. Autissier, et al.. Tritium retention in W plasma-facing materials: Impact of the material structure and helium irradiation. Nuclear Materials and Energy, 2019, 19, pp.403-410. ⟨10.1016/j.nme.2019.03.005⟩. ⟨hal-02086134⟩
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