Perhaps the easiest material to immobilize, at least technologically, is Pu. It is possible to vitrify Pu in a variety of glass compositions and immobilize it in various ceramic-based hosts. In the latter half of the 1990s, a number of scoping studies were published (e. g., Matzke and van Geel, 1996; Gray and Kan, 1996; Wicks et al., 1996), which outlined a number of potential hosts and demonstrated practically that it was indeed possible to immobilize Pu in a variety of glass and ceramic hosts.

The largest of these studies was performed on behalf of the US DOE which identified 72 possible options (Gray, 1996a) of which five were studied in depth, three involving vitrification and two immobilization in a ceramic host. For the vitrification processes lanthanum borosilicate glass based on the original ‘Loeffler’ optical glass composition was selected as the host in preference to lead iron phosphate or alkali tin silicate compositions, which had also been considered as potential hosts (Gray, 1996b). Studies of boro — silicate glass developed for vitrifying waste arising from reprocessing nuclear fuel elements have shown that this glass can also be used, but suffers from having a low actinide solubility, i. e. <3 mass% Pu, which compares unfavourably with lanthanum borosilicate glass, which can incorporate in excess of 10 mass% PuO2 (Meaker et al., 1997; Peeler et al., 1997). Two of the vitrification proposals were similar in that the waste plutonium and any scrap material, pre-treated where necessary, would be vitrified in the lan­thanum borosilicate glass. However, while one proposal (Gray, 1996b) required a new facility to be built, the second (Gray, 1996c) made use of an existing facility at SRS. In both proposals the glass would be poured into stainless steel HLW canisters, whereas in the third proposal (Gray, 1996d) it would be cast into small metal cans, 20 of which would be carefully posi­tioned within a HLW canister. The size of HLW canisters, 0.6 m diameter x 3 m high and weighing 2,000 kg, mitigates against theft, but additional secu­rity was proposed for the first two proposals by spiking the glass with suf­ficient 137Cs to maintain a y-radiation field above 1 Gy/hr for 30-60 years. A slightly different approach was proposed for the third option in that the voidage surrounding the cans would be filled with conventional borosilicate glass containing either 137Cs or HLW. Compositions of a selection of the glasses suggested for immobilizing surplus Pu are given in Table 25.6. Only a few glasses containing radioactive constituents have been prepared, most candidate compositions were made containing non-radioactive rare earth oxides as surrogates for Pu.

Phosphate glasses have also been investigated as they tend to have a higher solubility for actinides than silicate glasses and have been used in Russia as an alternative to borosilicate glass for the immobilization of HLW. Initially they suffered from poor durability and were highly corrosive in the molten state, so they found less favour than borosilicate glasses for which non-active processing technology existed. However, their durability has improved with the development of sodium aluminium phosphate (Minaev et al., 2004; Donald et al., 2006), iron phosphate (Day et al., 1998; Mogus-Milancovic et al., 1997) and lead iron phosphate glasses (Sales and Boatner, 1988) and they now have durabilities which match or exceed the standard borosilicate glass. Further information on phosphate-based glasses for immobilizing wastes is given elsewhere (e. g., Donald, 2010; Jantzen, 2011).

The introduction of CCIM technology largely addresses the problem of refractory liner corrosion by containing the melt within a solid skull of glass produced by cooling the outside of the furnace. Corrosion of glass contact refractories used in the vitrification of RAW has been reviewed in, e. g., Bingham et al. (2011).

Although the majority of the effort went into investigating vitrification, there was significant effort put into the ceramic option employing a Synroc — based composition which was subsequently selected as the most suitable choice, with disposition to be carried out in a similar manner to that of the vitrification route, i. e., ceramic pellets encapsulated in HLW glass in steel containers. Synroc (synthetic rock) is the generic name for a group of ceramics containing varying proportions of minerals found in nature, includ­ing hollandite (BaAl2Ti2O6), perovskite (CaTiO3), zirconolite (CaZrTi2O7) and rutile (TiO2) (Ringwood et al., 1979). Synroc-D was initially developed for defence requirements and consists primarily of perovskite, zirconolite, nepheline (NaAlSiO. ) and spinel (MgAl. O4), together with a continuous intergranular glassy phase, whilst the ceramic developed for immobilizing Pu consisted mainly of zirconolite, the primary phase for incorporating actinide elements. Waste forms with Pu loadings in excess of 10 mass% have been reported to exhibit excellent durability (Jostsons et al., 1995). Mono­lithic waste forms can be produced from a mixture of waste and ceramic precursor powder using conventional ceramic processing techniques, i. e., hot pressing (HP), hot isostatic pressing (HIP) and cold-pressing followed by sintering (CPS). A HP process developed at the Australian Nuclear

Glass BaO/ SrO І-З2О3 B2O3 ai2o3 Si02 PbO p2o5 Gd203 Nd203 Pu02 Na20 Others Beference Lan-14 3.5 30.4 4.7 15.0 27.0 11.3 8.2 CeO Bamsey et al. (1994) Lan-17 3.3 16.2 3.9 22.5 22.5 8.8 20.6 8.3 CeO Bamsey et al. (1994) LaAIBSi 4.4 23.2 4.3 9.5 26.3 11.1 15.0 6.1 Sm203 0.2 Zr02 Bibler et al. (1996) LaAIBSi 2.5 12.4 11.7 21.5 29.1 8.6 12.8 1.3 Zr02 Vienna et al. (1996) LaAIBSi 2.4 8.6 11.4 20.8 28.2 7.5 8.6 11.4 Macfarlane (1998) LaAIBSi 2.5 20.3 13.0 10.0 20.0 11.7 15.4 Macfarlane (1998) FEP 1 62.0 38.0 Fe203 M og us-M і I a ncovic et al. (1997) FEP 2 85.0 15.0 Fe203 M og us-M і I a ncovic et al. (1997) NaAIPI 2.0 19.0 39.0 40.0 Donald et al. (2006) MMW 17.9 6.3 50.4 12.5 8.9 4.1 Li20 Harrison et al. (2008) LaBS 2.5 19.0 13.0 10.0 20.0 13.5 15.0 7.0 Hf02 Fox et al. (2008) Frit X

Science and Technology Organisation (ANSTO) utilizes a stainless steel collapsible bellows can into which the mixture is placed. After evacuating and sealing the can, it is cold pressed to approximately two-thirds of its original length before being hot-pressed. Using simulated wastes, ANSTO have demonstrated this process on an industrial scale by successfully pro­ducing samples up to 436 mm in diameter.

More recently, both these potential options have been dropped in favour of use of surplus Pu as a mixed oxide (MOX) fuel, in line with the Russian view of Pu as a strategic material rather than a waste, for use in either Pu breeder reactors or light water power reactors (Gong et al., 2001; IPFM, 2009). France and Germany started bilateral programmes with Russia in 1992 which demonstrated the feasibility of recycling weapons grade Pu in Russian VVER 1000 and BN 600 reactors (Seyve et al., 1999 ).

The UK government has not declared any weapons grade plutonium to be surplus, but began a public consultation in 2011 (DECC, 2011) into the long-term management of the large stock of UK-owned civilian pluto­nium, 114.8 te at December 2010 (www. hse. gov) . Studies funded by the Nuclear Decommissioning Authority (NDA) into a variety of topics includ­ing re-use as MOX fuel, the preferred option, and immobilization will permit decisions to be made on the management of the stocks. Immobiliza­tion of civilian stocks declared surplus or unsuitable for re-use in ceramic and vitreous waste forms is being investigated (Harrison et al., 2008) and will provide a significant read-over to weapons grade plutonium should a future need arise.

Other alternative disposition options have been suggested, including the use of some surplus Pu to produce 99Mo by irradiation of 239Pu for medical applications (Mushtaq, 2011), but these must generally be viewed as only suitable for dealing with very small quantities.