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The Pb-17Li/water-cooled blanket is one of the concepts being developed in Europe for testing in NET (Next European Torus). This paper describes the impact of the latest experimental results on the blanket design. The points considered are: - breeder operating temperature and thermomechanical design: experiments on corrosion with steel 316L and liquid metal embrittlement tests provide upper and lower limits for the breeder operating temperature (280-400 C); - tritium recovery from the breeder and permeation rate to the coolant: Ispra measurements indicate that solubility and diffusivity of hydrogen in Pb-17Li are lower, compared with the previous values used in blanket tritium analyses. The impact of these results on the design of the tritium recovery system is discussed; - accident analyses: the experiments in progress at Ispra on the Pb-17Li/water interaction are reviewed and their application to a coolant pipe break accident is shown.

Additional information

Authors: CASINI G, JRC Ispra (IT)
Bibliographic Reference: Article: Nuclear Engineering and Design/Fusion, Vol. 8 (1988) pp. 139-143
Record Number: 199010517 / Last updated on: 1994-12-01
Original language: en
Available languages: en