As mentioned previously, the thorium cycle can reach positive breeding rates in a thermal spectrum. However, in order to solve the problem of the poison­ing by neutron captures on fission products and 233Pa, liquid fuels have to be used since they allow online extraction of selected nuclei. This is made pos­sible by the use of molten salt reactors (MSRs).

The concept of molten salt reactors has been extensively studied in the 1950s, 1960s and 1970s, in particular in the USA, where a demonstrator was operated for a few years at the Oak Ridge National Laboratory. The design of the molten salt breeder reactor (MSBR) was a cylindrical assembly of high-density graphite hexagons, each hollowed out to accommodate a

channel for molten salt circulation. Above and under the core salt tanks and graphite axial reflectors are located (see figure 11.8). The fuel circulates from the bottom to the top at a velocity of around 2 m/s. The fraction of the fuel which is outside the core (heat exchanger, reprocessing unit) is about 30%. Of course, a significant part of the delayed neutrons will be emitted outside the core, which reduces the proportion of delayed neutrons that can be used to control the reactor.

The use of a liquid fuel allows continuous feeding of the core with fertile or fissile nuclei, in order to maintain the reactivity and to keep the chemical composition of the salt constant. It saves having large reactivity reserves at the beginning of the irradiation, and improves the neutron balance. Different extraction methods can be considered. One of them is based on a liquid — liquid extraction process, which consists of exchanging thorium and lithium dissolved in molten bismuth for the constituents to be removed from the salt. Chemically, the fission products are more or less similar to thorium, and their extraction efficiencies can vary: they are around 20% for halogens and rare earths, around 5% for Zr and semi-noble metals and 1% for alkaline elements. These values have to be confirmed at an industrial scale, but seem accessible within a few tens of years. Figure 11.9 shows a simplified schematic of the materials reprocessing principle.

The reprocessing time is a free parameter which can be adapted to the performances one tries to reach. A fast reprocessing rate will correspond to an optimal breeding rate. The carrier salt can be LiF-BeF2-(HN)F4, where HN denotes heavy nuclei. The proportions are around 70:17.5: 12.5mol% respectively.

The main advantage of the thermal spectrum is the fissile inventory, which is greatly reduced compared with the fast-spectrum case. A 1 GWe reactor needs around 67 tons of Th and only 1.1 tons of U. Table