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

E. Bernard*, R. Sakamoto, E. Hodille, A. Kreter, E. Autissier, M. F. Barthe, P. Desgardin, T. Schwarz-Selinger, V. Burwitz, S. Feuillastre, S. Garcia-Argote, G. Pieters, B. Rousseau, M. Ialovega, R. Bisson, F. Ghiorghiu, C. Corr, M. Thompson, R. Doerner, S. MarkeljH. Yamada, N. Yoshida, C. Grisolia

*Corresponding author for this work

    Research output: Contribution to journalArticlepeer-review

    20 Citations (Scopus)


    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.

    Original languageEnglish
    Pages (from-to)403-410
    Number of pages8
    JournalNuclear Materials and Energy
    Publication statusPublished - May 2019


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