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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 University of Helsinki
4 Forschungszentrum Jülich GmbH
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 [Fukuoka]
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.
Keywords : Plasma-wall interactions Tungsten Helium Tritium inventory
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UNIV-AMU | CNRS | DSM-IRFM | CEA | PIIM | UNIV-LYON1 | INC-CNRS | INSA-LYON | UNIV-ORLEANS | CEA-DRF | CEA-CAD | INSA-GROUPE | CEMHTI | AMIDEX | UDL

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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, Elsevier, 2019, 19, pp.403-410. ⟨10.1016/j.nme.2019.03.005⟩. ⟨hal-02086134⟩

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