Tungsten and Beryllium Armour Development for the JET ITER-like Wall Project

J. Likonen, H. Maier, et al., JET-EFDA contributors

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

    Abstract

    The operational behaviour and the interplay of the ITER plasma facing material choice has never been investigated in a tokamak experiment. This motivated the ITER-like Wall Project at JET in which the present main chamber CFC tiles will be exchanged with Be tiles and in parallel a fully tungsten-clad divertor will be prepared. Among the scientific objectives of the ITER-like Wall project are general questions of plasma operation with a low melting Be wall, compatibility of all envisaged ITER scenarios with a W divertor, tritium retention and removal and mixed materials effects, erosion behaviour and lifetime investigations. For the tungsten divertor, two R&D programs were initiated: Forschungszentrum Jülich, Germany, developed a conceptual design for a bulk W horizontal target plate, based on an assembly of tungsten blades. For all other divertor parts five Euratom Fusion Associations performed R&D to provide the technology to coat the 2-directional CFC material used at JET with thin tungsten coatings. In addition, beryllium coatings for the first wall inconel steel are developed. In a first screening, the tungsten coated CFC tiles were subjected to heat loads with power densities ranging from 6 MW/m2 to 22 MW/m2 with surface temperatures exceeding 2000oC. In a second step, a selection of coatings was exposed to cyclic heat loading for 200 pulses at 10 MW/m2 for 5 s corresponding to surface temperatures of about 1600oC. All coatings tested developed cracks perpendicular to the CFC fibres due to the stronger contraction of the coating upon cool-down after the heat pulses. For the bulk tungsten, a design with an an assembly of tungsten blades was developed. To minimise electromagnetic forces the design consists of stacks of tungsten blades of 6 mm width that are insulated in toroidal direction. High heat flux tests of a test module were performed on the electron beam facility JUDITH at a nominal power and duration of (7 MW/m2, 10 s) for 100 pulses and finally with increasing power loads leading to surface temperatures in excess of 3000oC. No macroscopic failure occurred during the test while SEM showed the development of microcracks at grain boundaries.
    Original languageEnglish
    Title of host publicationFusion Energy 2006
    Subtitle of host publicationProceedings of the 21st IAEA Conference
    PublisherInternational Atomic Energy Agency IAEA
    ISBN (Electronic)92-0-100907-0
    Publication statusPublished - 2006
    MoE publication typeB3 Non-refereed article in conference proceedings
    Event21st IAEA Fusion Energy Conference - Chengdu, China
    Duration: 16 Oct 200621 Oct 2006

    Publication series

    SeriesIAEA Conference Proceedings
    NumberIAEA-CN-149

    Conference

    Conference21st IAEA Fusion Energy Conference
    CountryChina
    CityChengdu
    Period16/10/0621/10/06

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