This section only covers the key aspects of partitioning and transmutation technology relevant to this chapter (see Chapter 17 for further information). The type of fuel form is dependent not only on the kind of reactor (thermal or fast), but also on the type of cycle (heterogeneous or homogeneous) and the nature of the envisaged reprocessing. Different options have been considered for transmutation of advanced fuels (Ogawa et al., 2005; Warin and Boullis, 2008). The main issue is the very different chemical behaviour of the various actinides. The ‘classic’ fuel forms like oxides, successfully used up to now for U — and (U, Pu)-based fuels, are not directly applicable to MA-bearing fuels. In order to obtain a high level of transmutation, the fuel should be irradiated up to very high burn-ups, and it should be particularly resistant to radiation damage. Key points for the choice of a transmutation fuel are also thermal conductivity and density. Carbides and nitrides meet these requirements. However, carbides and nitrides of Am are volatile and extremely difficult to treat, while there is still limited knowledge about the behaviour of carbides and nitrides of Cm (Bomboni, 2009). Another important aspect to consider is the production of He in MA-based fuels due to the transmutation of 241Am, which could cause an excessive swelling rate. These problems have led to research on U, Pu or MA-bearing oxides, nitrides (Arai et at., 2008), carbides (NEA, 2005) and metals (Pasamehmetoglu, 2008) for thermal or fast reactors.

Another option is that of fuel dispersed in a matrix. Inert matrix fuel (IMF) consists of a neutron-transparent matrix (generally with good thermal conductivity) and a fissile phase that is either dissolved in the matrix or incorporated as macroscopic inclusions. The matrix plays a crucial role in diluting the fissile phase to the volumetric concentrations required for reactor control. The same role is played by 238U in conventional low enriched uranium (LEU) or MOX fuel. The key difference is that replacing fertile 238U with a neutron-transparent matrix eliminates plutonium breeding as a result of neutron capture. Oxides, metals, carbides and nitrides have been found to be suitable matrix materials. As an example, silicon carbide (SiC) and yttrium-stabilized zirconia (YSZ) are two typical IM materials for He-cooled high-temperature reactor technology (IAEA, 2006).

The concept of a dedicated, moderated assembly for MA burning in FRs is also relevant to transmutation. In order to better exploit the high flux of FRs, introducing a MA-bearing fuel mixed with moderating material in certain zones of the core can maximize the transmutation rate of these nuclides. Many different materials have been considered as moderators for these dedicated assemblies: hydrides are a particularly promising option, since they are very good moderators because of their H content. However, one disadvantage is their relatively ‘low’ (at least as far as high-temperature reactors are concerned) operating temperatures (<800 °C).

As far as the reprocessing technology of advanced fuels is concerned, a few processes, developed at the laboratory scale, have already reached the target efficiency (see Table 13.6 for some examples of processes developed in Europe). Pyrochemical methods, for example, are a very attractive option for high burn-up fuels, since they are based on fuel dissolution in molten salts from which individual actinides are selectively precipitated by electrorefining. A few aqueous processes also show promise, but these are based on organic molecules that are vulnerable to radiolysis, while molten salts are highly stable in all possible conditions.

Whereas aqueous processes require mainly oxide fuels, pyrochemical processes allow the treatment of any fuel form and the recovery of virtually every kind of element. Nevertheless, they are currently characterized by low recovery efficiencies, very corrosive reagents and high secondary losses (NEA, 2006a).

AIROX-like processes, in which UO2 fuel is powdered by oxidation to U3O8 and re-enriched without separation of either solid fission products or HMs, could offer a potential alternative (see Fig. 13.3). Using these processes, it should be possible to decouple deployment of FRs from development of expensive commercial reprocessing of LWR SNF (Greenspan, 2007; Feinroth et al., 1993). The carbon dioxide oxidation (CARDIO) process is another innovative reprocessing method (Greenspan, 2007). It is a dry process for UC spent fuel. The process can be subdivided into three stages:

1 UC spent fuel can be converted into UO2 via UC + 3CO2 ^ UO2 + 4CO at T>670°C

2 Applying the AIROX process it is then possible to remove volatile fission products

3 Applying the carbothermic reduction of oxide fuel in a high-purity inert atmosphere (UO2 + 3C ^ UC + 2CO) it is possible to produce UC fuel again.