NEUTRONICS ASSESSMENT OF MOLTEN SALT BREEDING BLANKET DESIGN OPTIONS

摘要

Neutronics assessment has been performed for molten salt breeding blanket design options that can be utilized in fusion power plants. The concepts evaluated are a self-cooled Flinabe blanket with Be multiplier and dual-coolant blankets with He-cooled FW and structure. Three different molten salts were considered including the high melting point Flibe, a low melting point Flibe, and Flinabe. The same TBR can be achieved with a thinner self-cooled blanket compared to the dual-coolant blanket. A thicker Be zone is required in designs with Flinabe. The overall TBR will be ～1.07 based on 3-D calculations without breeding in the divertor region. Using Be yields higher blanket energy multiplication than obtainable with Pb. A modest amount of tritium is produced in the Be (～3 kg) over the blanket lifetime of ～3 FPY. Using He gas in the dual-coolant blanket results in about a factor of 2 lower blanket shielding effectiveness. We show that it is possible to ensure that the shield is a lifetime component, the vacuum vessel is reweldable, and the magnets are adequately shielded. We conclude that molten salt blankets can be designed for fusion power plants with neutronics requirements such as adequate tritium breeding and shielding being satisfied.