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

Show full item record



Bernard , E , Sakamoto , R , Hodille , E , Kreter , A , Autissier , E , Barthe , M-F , Desgardin , P , Schwarz-Selinger , T , Burwitz , V , Feuillastre , S , Pieters , G , Rousseau , B , Iolavega , M , Bisson , R , Ghiorghiu , F , Corr , C , Thompson , M , Doerner , R , Markelj , S , Yamada , H , Yoshida , N & Grisolia , C 2019 , ' Tritium retention in W plasma-facing materials : Impact of the material structure and helium irradiation ' , Nuclear Materials and Energy , vol. 19 , pp. 403-410 .

Title: Tritium retention in W plasma-facing materials : Impact of the material structure and helium irradiation
Author: Bernard, E; Sakamoto, R; Hodille, E.; Kreter, A; Autissier, E; Barthe, M-F; Desgardin, P; Schwarz-Selinger, T; Burwitz, V; Feuillastre, S; Pieters, G; Rousseau, B; Iolavega, M; Bisson, R; Ghiorghiu, F; Corr, C; Thompson, M; Doerner, R; Markelj, S; Yamada, H; Yoshida, N; Grisolia, C
Contributor organization: Materials Physics
Department of Physics
Date: 2019-05
Language: eng
Number of pages: 8
Belongs to series: Nuclear Materials and Energy
ISSN: 2352-1791
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 degrees 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 degrees C, meaning ideal W material studies may underestimate T inventory for tokamak relevant conditions.
Description: This article has an erratum: DOI 10.1016/j.nme.2020.100729
Subject: 114 Physical sciences
plasma-wall interactions
Tritium inventory
Peer reviewed: Yes
Rights: cc_by
Usage restriction: openAccess
Self-archived version: publishedVersion

Files in this item

Total number of downloads: Loading...

Files Size Format View
1_s2.0_S2352179118302424_main.pdf 1.568Mb PDF View/Open

This item appears in the following Collection(s)

Show full item record