Production of ITER-relevant Be-containing laboratory samples for fuel retention investigations

Antti Hakola (Corresponding author), Kalle Heinola, Jari Likonen, Cristian Lungu, Corneliu Porosnicu, E Alves, R. Mateus, Iva Bogdanović Radović, Z. Siketić, V. Nemanic

    Research output: Chapter in Book/Report/Conference proceedingConference abstract in proceedingsScientific


    Since 2014, a large project has been running under the EUROfusion Consortium to produce ITER-relevant test samples for fuel-retention studies. The strategy is to deposit mixed, Be-containing coatings at the National Institute for Laser, Plasma and Radiation Physics in Romania and distribute the samples for analyses and/or ion-implantation in partner laboratories. The composition, thickness, and surface structure of the deposits have been varied to study their influence on the efficiency of D retention and to make predictions for ITER. For benchmarking purposes, samples resembling the co-deposited layers on the inner divertor of JET during its ITER-Like Wall (ILW) campaigns [1] have also been produced. The properties of the samples have been determined using a variety of surface-analysis tools including Rutherford Backscattering Spectroscopy (RBS), Nuclear Reaction Analysis (NRA), Time-of-flight Elastic Recoil Detection Analysis (TOF-ERDA), Secondary Ion Mass Spectrometry (SIMS), Thermal Desoprtion Spectroscopy (TDS), and Laser-Induced Breakdown Spectroscopy (LIBS).

    The focus has been put on D-doped Be-O, Be-W, Be-C-O (in the case of JET-ILW comparison) coatings with different surface morphologies in the nanoscale and thicknesses ranging from about 0.1 um to some 15 um. The relative D content of the samples can routinely be increased to 5-10 at.%, and in some coating types even 40 at.% has been reached. This is a good starting point to fabricate co-deposits with seeding gases and/or helium in the future. The Be-O-C-D coatings most closely resembling the JET-ILW co-deposits (O and C content 5-10 at.%, D content ~5 at.%, thickness ~15 um) show very similar release behavior of D as real JET samples, indicating that the laboratory samples well mimic the structure of co-deposits in fusion reactors.

    The surface analyses have revealed that increasing the O content from a few to 50 at.% in otherwise identical samples lowers the amount of D that can be retained, by up to a factor of 10. Implantation with D+ ions results in similar retention behavior, though with more peaked D profiles, than direct doping during the deposition phase. The data also indicate that more D can accumulate in the sample if the thickness of the coating is increased, the surface becomes more modified and rough, or, in the case of mixed Be-W deposits, the relative Be fraction increases.

    [1] K. Heinola et al., Experience on divertor fuel retention after two ITER-Like Wall campaigns, Phys. Scr. (accepted).
    Original languageEnglish
    Title of host publication23rd PSI Conference Princeton USA
    Subtitle of host publicationBook of Abstracts
    PublisherPrinceton University
    Number of pages1
    Publication statusPublished - Jun 2018
    MoE publication typeNot Eligible
    Event23rd International Conference on Plasma Surface Interactions in Controlled Fusion Devices - Princeton University, Princeton, United States
    Duration: 17 Jun 201822 Jun 2018


    Conference23rd International Conference on Plasma Surface Interactions in Controlled Fusion Devices
    Abbreviated titlePSI-23
    Country/TerritoryUnited States
    Internet address


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