The VHTR is a thermal neutron system originally designed to operate a once — through fuel cycle with low-enriched uranium fuel and very high fuel burn-up. However, the system’s flexibility will allow it to adopt closed fuel cycles using thorium fuel and ‘burner’ cores that can efficiently transmute Pu (Abram and Ion, 2008; Generation IV International Forum, 2009). It will be possible to use VHTR with Pu fuel and for MA incineration or transmutation, due to the high burn-up capabilities of the coated particle fuel, though the build-up of even — numbered Pu isotopes is an issue that will need addressing. These features can also be used in symbiosis with other reactor types to reduce MA content and decay heat, which effect repository design (see 13.4.3). The deep-burn potential of VHTR avoids multi-recycling of spent fuel. It is especially attractive if it can be shown that ultra-high burn-up coated particles are still able to maintain their barrier function under disposal conditions (Brossard et al., 2009). VHTR could also potentially use Th as a fuel, as shown by the experience of HTR (Mazzini et al., 2009). Using a Th-based fuel cycle in a HTR with Pu as a driver would (see Fig. 13.8) increase the efficiency of TRU fission and achieve higher fuel burn-ups. Another advantage is that Th is about three times more abundant than U.

13.8 Mass of actinides at EOC starting from 1 g of Pu (Cerullo et al., 2005).

The VHTR core can be constructed from one of two basic designs (see Fig. 13.6):

1 the prismatic block type

2 the pebble bed type

From core configurations (deterministic vs. stochastic) and refuelling schemes (batch-wise vs. continuous) points of view, the pebble-bed and prismatic fuel design are quite different (Lomonaco, 2003). Just to give an example, the pebble-bed core
configuration needs specific design tools, as underlined in Bomboni et al. (2009a, 2010, 2012).

Anyway these designs have a number of common features. These include a UO2 kernel surrounded by successive layers of porous graphite, dense pyrocarbon (PyC), silicon carbide (SiC) then pyrocarbon (PyC). This could be enhanced through the use of a UCO kernel or ZrC coating. These coatings have the potential to provide improved burn-up capability, minimized fission product release and increased resistance to core heat-up accidents, even above 1600 °C, which is considered the maximum operating temperature for TRISO fuel (Brossard et al., 2009) (see Fig. 13.9). Coating of fuel particles could be achieved by chemical vapour deposition.

Empirical formulations exist for HTR fuel but little is known about how different process parameters, e. g. gas composition and temperature, would affect the properties and the performance of the resulting fuel (Abram and Ion, 2008). There is also a need to undertake more research on manufacture, characterization and irradiation performance. Irradiation tests are necessary for the fabrication process, fuel design and fission product transport, as well as for post-irradiation and safety testing. Fuel performance must also be assessed for both normal operating and accident scenarios. A key requirement of the fuel is its ability to retain fission products in the fuel particles under a range of accident scenarios with temperatures up to 1600 °C. Although very good irradiation performance has been demonstrated under HTR conditions, the behaviour of

13.9 I nfluence of temperature on TRISO fuel failure fraction. Brossard et al., 2009.

the coated particles under irradiation is not fully understood (Abram and Ion, 2008).

There are a number of possible methods for dealing with spent fuel:

• direct disposal of coated particles and graphite moderator

• separation of coated particles and moderator, with separate treatment of both fractions

• separation of kernels from coatings and reprocessing kernels for recycling in VHTR systems (or other reactors)

Research on long-term repository/direct disposal is currently focused on the potential of SiC coatings. It is believed that the coating may act as a miniature containment vessel to retain fission products during the repository post-closure period but more information are needed about the long-term integrity of these layers (Brossard et al., 2009). The main focus of research in reprocessing fuel is particle kernel dissolution (Brossard et al., 2009). Other research has concentrated on technologies to separate the highly active graphite fractions from those of low activity and to evaluate the feasibility of reusing the graphite (Brossard et al., 2009).