Power handling and vapor shielding of pre-filled lithium divertor targets in Magnum-PSI

P. Rindt (Corresponding author), T.W. Morgan, G. van Eden, M.A. Jaworski, N.J. Lopes Cardozo

Onderzoeksoutput: Bijdrage aan tijdschriftTijdschriftartikelAcademicpeer review

1 Citaat (Scopus)

Uittreksel

To develop realistic liquid lithium divertors for future fusion reactors, this paper aims to improve the understanding of their power handling capabilities. A liquid lithium divertor target prototype, designed to facilitate liquid metal experiments in tokamaks, was tested in Magnum-PSI. The target has an internal reservoir pre-filled with lithium and aims to passively re-supply the textured plasma facing surface during operation. To assess the power handling capability the target was exposed to helium plasmas with increasing power flux density in the linear plasma device Magnum-PSI. The temperature response of lithium targets was recorded via an infrared camera, and compared to finite element method modeling taking into account dissipation via lithium in the plasma. It was found that the target works as intended and can take up to 9 1 MW m−2 for 10 s before the mesh layer was damaged, and could continue operating at higher power densities even after being damaged. The total lifetime of the targets was up to 100 s. Overall the targets are found suitable for use in tokamak experiments. Additionally, a central surface temperature evolution indicative of vapor shielding was observed on intact targets. Predicting the target temperature (and consequently the evaporation rates and thermal stresses) is considered very relevant for the design of lithium divertor targets for DEMO. The observed temperature response could indeed be replicated through modeling, which showed that a significant power fraction was dissipated by the lithium in the plasma.
TaalEngels
Artikelnummer056003
Aantal pagina's13
TijdschriftNuclear Fusion
Volume59
Nummer van het tijdschrift5
DOI's
StatusGepubliceerd - mei 2019

Trefwoorden

    Citeer dit

    @article{91595cdcf02243f1b36a30d321e445f7,
    title = "Power handling and vapor shielding of pre-filled lithium divertor targets in Magnum-PSI",
    abstract = "To develop realistic liquid lithium divertors for future fusion reactors, this paper aims to improve the understanding of their power handling capabilities. A liquid lithium divertor target prototype, designed to facilitate liquid metal experiments in tokamaks, was tested in Magnum-PSI. The target has an internal reservoir pre-filled with lithium and aims to passively re-supply the textured plasma facing surface during operation. To assess the power handling capability the target was exposed to helium plasmas with increasing power flux density in the linear plasma device Magnum-PSI. The temperature response of lithium targets was recorded via an infrared camera, and compared to finite element method modeling taking into account dissipation via lithium in the plasma. It was found that the target works as intended and can take up to 9 1 MW m−2 for 10 s before the mesh layer was damaged, and could continue operating at higher power densities even after being damaged. The total lifetime of the targets was up to 100 s. Overall the targets are found suitable for use in tokamak experiments. Additionally, a central surface temperature evolution indicative of vapor shielding was observed on intact targets. Predicting the target temperature (and consequently the evaporation rates and thermal stresses) is considered very relevant for the design of lithium divertor targets for DEMO. The observed temperature response could indeed be replicated through modeling, which showed that a significant power fraction was dissipated by the lithium in the plasma.",
    keywords = "fusion, divertor, lithium, power handling, Magnum-PSI, prototype testing",
    author = "P. Rindt and T.W. Morgan and {van Eden}, G. and M.A. Jaworski and Cardozo, {N.J. Lopes}",
    year = "2019",
    month = "5",
    doi = "10.1088/1741-4326/ab0560",
    language = "English",
    volume = "59",
    journal = "Nuclear Fusion",
    issn = "0029-5515",
    publisher = "Institute of Physics",
    number = "5",

    }

    Power handling and vapor shielding of pre-filled lithium divertor targets in Magnum-PSI. / Rindt, P. (Corresponding author); Morgan, T.W.; van Eden, G.; Jaworski, M.A.; Cardozo, N.J. Lopes.

    In: Nuclear Fusion, Vol. 59, Nr. 5, 056003, 05.2019.

    Onderzoeksoutput: Bijdrage aan tijdschriftTijdschriftartikelAcademicpeer review

    TY - JOUR

    T1 - Power handling and vapor shielding of pre-filled lithium divertor targets in Magnum-PSI

    AU - Rindt,P.

    AU - Morgan,T.W.

    AU - van Eden,G.

    AU - Jaworski,M.A.

    AU - Cardozo,N.J. Lopes

    PY - 2019/5

    Y1 - 2019/5

    N2 - To develop realistic liquid lithium divertors for future fusion reactors, this paper aims to improve the understanding of their power handling capabilities. A liquid lithium divertor target prototype, designed to facilitate liquid metal experiments in tokamaks, was tested in Magnum-PSI. The target has an internal reservoir pre-filled with lithium and aims to passively re-supply the textured plasma facing surface during operation. To assess the power handling capability the target was exposed to helium plasmas with increasing power flux density in the linear plasma device Magnum-PSI. The temperature response of lithium targets was recorded via an infrared camera, and compared to finite element method modeling taking into account dissipation via lithium in the plasma. It was found that the target works as intended and can take up to 9 1 MW m−2 for 10 s before the mesh layer was damaged, and could continue operating at higher power densities even after being damaged. The total lifetime of the targets was up to 100 s. Overall the targets are found suitable for use in tokamak experiments. Additionally, a central surface temperature evolution indicative of vapor shielding was observed on intact targets. Predicting the target temperature (and consequently the evaporation rates and thermal stresses) is considered very relevant for the design of lithium divertor targets for DEMO. The observed temperature response could indeed be replicated through modeling, which showed that a significant power fraction was dissipated by the lithium in the plasma.

    AB - To develop realistic liquid lithium divertors for future fusion reactors, this paper aims to improve the understanding of their power handling capabilities. A liquid lithium divertor target prototype, designed to facilitate liquid metal experiments in tokamaks, was tested in Magnum-PSI. The target has an internal reservoir pre-filled with lithium and aims to passively re-supply the textured plasma facing surface during operation. To assess the power handling capability the target was exposed to helium plasmas with increasing power flux density in the linear plasma device Magnum-PSI. The temperature response of lithium targets was recorded via an infrared camera, and compared to finite element method modeling taking into account dissipation via lithium in the plasma. It was found that the target works as intended and can take up to 9 1 MW m−2 for 10 s before the mesh layer was damaged, and could continue operating at higher power densities even after being damaged. The total lifetime of the targets was up to 100 s. Overall the targets are found suitable for use in tokamak experiments. Additionally, a central surface temperature evolution indicative of vapor shielding was observed on intact targets. Predicting the target temperature (and consequently the evaporation rates and thermal stresses) is considered very relevant for the design of lithium divertor targets for DEMO. The observed temperature response could indeed be replicated through modeling, which showed that a significant power fraction was dissipated by the lithium in the plasma.

    KW - fusion

    KW - divertor

    KW - lithium

    KW - power handling

    KW - Magnum-PSI

    KW - prototype testing

    U2 - 10.1088/1741-4326/ab0560

    DO - 10.1088/1741-4326/ab0560

    M3 - Article

    VL - 59

    JO - Nuclear Fusion

    T2 - Nuclear Fusion

    JF - Nuclear Fusion

    SN - 0029-5515

    IS - 5

    M1 - 056003

    ER -