Tritium Migration in HCLL and WCLL Blankets: Impact of Tritium Solubility in Liquid Pb-17Li

Abstract

The next generation of fusion power plants (DEMO) should rely on a breeding blanket (BB) able to efficiently convert the neutrons kinetic energy into heat, to ensure the tritium self-sufficiency and to adequately shield the toroidal field coils from neutrons and gamma rays. The eutectic lithium-lead alloy is a consolidate liquid blanket material, which simultaneously includes the breeder (Li) and the neutron multiplier (Pb). The assessment of the tritium inventory inside the blanket and its environmental release requires knowledge of the hydrogen isotopes interactions with blanket materials, in particular the hydrogen solubility in lithium-lead, which is defined by means of the hydrogen Sievert's constant. Several experiments, aiming to determine the hydrogen isotopes solubility in lithium-lead, have been performed in the past giving values of the temperature-dependent Sieverts' constant, K , distributed in a wide range (covering about two orders of magnitude on the Arrhenius plot). Starting from a literature review of the K values, this paper provides a parametric analysis for the most influencing parameters related to tritium migration in helium-cooled and water-cooled lead-lithium DEMO BBs. This analysis has been performed using the computational code FUS-TPC for several operative scenarios and considering the different K values. This paper demonstrates that the tritium Sieverts' constant in Pb-17Li has a great impact on the assessment of tritium losses, whose value can spread in more than one order of magnitude. Furthermore, the analysis suggests suitable permeation reduction factors to be adopted in the different scenarios as well as the need of addressing new accurate experiments on the solubility constant.