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

E. Bernard (1) , R. Sakamoto (2) , E. Hodille (3) , A. Kreter (4) , E. Autissier (5) , M.-F. Barthe (5) , P. Desgardin (5) , T. Schwarz-Selinger (6) , V. Burwitz (6) , S. Feuillastre (7) , S. Garcia-Argote , G. Pieters (8) , B. Rousseau (9) , M. Ialovega (10) , R. Bisson (11) , F. Ghiorghiu (11) , C. Corr (12) , M. Thompson (12) , R. Doerner (13) , S. Markelj (14) , H. Yamada (15) , N. Yoshida (16) , C. Grisolia (10)

1 ST-CROLLES - STMicroelectronics [Crolles]
2 NIFS - National Institute for Fusion Science
3 Helsingin yliopisto = Helsingfors universitet = University of Helsinki
4 Forschungszentrum Jülich GmbH | Centre de recherche de Juliers
5 CEMHTI - Conditions Extrêmes et Matériaux : Haute Température et Irradiation
6 IPP - Max-Planck-Institut für Plasmaphysik [Garching]
7 ICBMS - Institut de Chimie et Biochimie Moléculaires et Supramoléculaires
8 CEA - Commissariat à l'énergie atomique et aux énergies alternatives
9 III-V Lab - Alcatel-Thalès III-V lab
10 IRFM - Institut de Recherche sur la Fusion par confinement Magnétique
11 PIIM - Physique des interactions ioniques et moléculaires
12 University of Canberra
13 UC San Diego - University of California [San Diego]
14 IJS - Jozef Stefan Institute [Ljubljana]
15 NIFS - Nastional Institute for Fusion Science
16 Kyushu University

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.

Keywords: Plasma-wall interactions; Tritium inventory; Tungsten; Helium

Domains: Physics [physics]; Physics [physics]; Chemical Physics [physics.chem-ph]

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⟩