Helium-3 blankets for tritium breeding in near-term fusion devices

Abstract

The purpose of this paper is to propose a novel blanket concept for in-situ tritium breeding in a near-term device such as ITER. In this concept terrestial supplies of helium-3, rather than lithium, are used for tritium breeding. In order to assess the key characteristics of this concept, a reference configuration was adopted based on minor modifications to the helium-cooled blanket concepts considered in the Blanket Comparison and Selection Study. The chosen configuration assumes a ferritic steel for structure and cladding and beryllium for neutron multiplication. The helium-3 is contained in a loop separate from the helium-4 coolant loop and flows within the beryllium. The helium-3 blanket exhibits good tritium breeding potential and low tritium inventories and leakage rates. The helium-3 requirements are 25–50 kg of inventory and 3.4–8.5 kg makeup/yr, 95% of which is due to burnup. It is estimated that there is sufficient helium-3 from decay of present military tritium stockpiles to meet this requirement.