The design of molten salt reactors is similar to those for other thermal reactors, including similar neutron fluxes and reactivity temperature coefficients (Rosenthal et al., 1972). In particular, the graphite-moderated MSR has much in common with the HTR: a graphite moderator at an average temperature of 600-700 °C, Th-233U fuel cycle, and similar fuel-moderator ratios. However, there are some important differences. Table 13.9 shows some relevant parameters for MSR, using 233U or 235U as the fissile material. The main parameter shown in this table is the critical fuel concentration; to obtain the same reactivity only half as much 233U is needed as 235U, because of the superior neutronic characteristics of 233 U (Rosenthal et al., 1972). That is an important feature to take into account for the design of reactors that use Th as fertile material.

7LiF-BeF2 (66/34 in mol%) salt is used as the fuel carrier for the moderated (thermal) molten salt thorium breeder, producing 7LiF-BeF2-ThF4-UF4 as fuel salt. All of the alternatives to this salt reduce the breeding capacity of the reactor

(Renault et al., 2009). SFR systems use a molten salt actinide burner. A carrier salt with good solubility for actinide trifluorides is needed; something which can be achieved using 7LiF-NaF-(KF) as the solvent or 7LiF-(NaF)-BeF2 melt. An interesting alternative is the use of Pu and MAs as start-up fuel for the Th cycle in the MSR, leading to 7LiF-NaF-ThF4 carrier salt (Renault et al., 2009). A single stream Li, Na, Be/F molten salt actinide recycler and transmuter (MOSART) fast spectrum system fuelled by combinations of Pu with MA trifluorides (AnF3) from UOX and MOX has been developed in Russia (Ignatiev et al., 2005). A modified two-fluid Th-U molten salt system (based on MOSART) has also been developed more recently (Feynberg and Ignatiev, 2010). A comparison between MOSART, MSBR and MSFR is provided in Table 13.10 . In order to summarize different potential combinations, some reference salts and fuel compositions are reported in Table 13.11.

Salt processing relies on both on-line and batch processes to maintain smooth reactor operation while minimizing losses to waste streams. The removal of lanthanides is necessary because of their low solubility in the molten salt and their adverse effect on reactivity through neutron capture (Renault et al., 2009). A potential processing scheme is shown in Fig. 13.14 . The main innovation is stages 2 and 3, which combine chemical and electrochemical extraction methods with the back extraction of actinides and lanthanides. This allows fuel processing to occur with no variation in effluent volume, while reducing the fuel processing balance to just one reaction: 2LnF. + 3H. O(g) = Ln. O3 + 6HFfe). An effective method for actinide/lanthanide separation is still needed (Delpech et al., 2008). A practicable fuel clean-up rate is 40 l per day, corresponding to the

processing of 100 kg heavy nuclei per day. This value is almost two orders of magnitude lower than that needed by the reference MSBR scheme (Renault et al., 2009).