Tritium retention in W plasma-facing materials: Impact of the material structure and helium irradiation. (May 2019)

• Record Type: Journal Article
• Authors: Bernard, E.; Sakamoto, R.; Hodille, E.; Kreter, A.; Autissier, E.; Barthe, M.-F.; Desgardin, P.; Schwarz-Selinger, T.; Burwitz, V.; Feuillastre, S.; Garcia-Argote, S.; Pieters, G.; Rousseau, B.; Ialovega, M.; Bisson, R.; Ghiorghiu, F.; Corr, C.; Thompson, M.; Doerner, R.; Markelj, S.; Yamada, H.; Yoshida, N.; Grisolia, C.
• Is Part Of: Nuclear materials and energy. Volume 19(2019)
• Page Start: 403
• Page End: 410
• Publication Date: 2019-05
• Subjects: Tungsten -- Helium -- Tritium inventory -- Plasma-wall interactions; Nuclear energy -- Periodicals; Nuclear fuels -- Periodicals; Nuclear reactors -- Materials -- Periodicals; Radioactive substances -- Periodicals; 621.4833
• DOI: 10.1016/j.nme.2019.03.005
• Languages: English
• ISSNs: 2352-1791

Abstract

Highlights: Major impact of preexisting defects in W on T inventory (up to 2.5 times). Manufacturing process and/or sample preparation are crucial. He irradiation can drastically increase T inventory in W (up to 1.8 times). He irradiation type plays a key role in the trapping sites creation for T. All he irradiations create a specific trap for T desorbing as HT at 20 °C. 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.