Immobilisation options

Choosing a suitable waste form for nuclear waste immobilisation is difficult and durability is not the sole criterion. In any immobilisation process where radioactive materials are used, the process and operational conditions can become complicated, particularly if operated remotely and equipment maintenance is required. Therefore priority is given to reliable, simple, rugged technologies and equipment, which may have advantages over complex or sensitive equipment.

A variety of matrix materials and techniques is available for immobilisa­tion (NRC, 2011). The choice of the immobilisation technology depends on the physical and chemical nature of the waste and the acceptance criteria for the long-term storage and disposal facility to which the waste will be consigned. A host of regulatory, process and product requirements has led to the investigation and adoption of a variety of matrices and technologies
for waste immobilisation. The main immobilisation technologies that are available commercially and have been demonstrated to be viable are cementation, bituminisation and vitrification.

Immobilisation can be simply physically surrounding the waste in a barrier material (largely the case in cementation) or chemically incorporating it into the structure of a host material (largely the case in vitrification).

Подпись: (a) 1.12 ( a) 500 litre steel drums containing cemented ILW and (b) vitrified waste in 2 m tall steel containers. image17

Cementation uses hydraulic cements to physically surround solid ILW that is contained in steel drums (Fig. 1.12a). Ordinary Portland cement (OPC) is the most common type of cement used for immobilising liquid and wet solid wastes worldwide. Several OPC-based mixtures are currently used to improve the characteristics of waste forms and overcome the incom­patibility problems associated with the chemical composition of certain types of radioactive waste. Composite cement systems may use additional powders as well as OPC such as blast furnace slag (BFS) and pulverised fuel ash (PFA). These offer cost reduction, energy saving and potentially superior long-term performance. As well as the waste form matrix, OPCs will be used in structural components of any GDF (such as walls and floors) and are potential backfill materials, so an understanding of their durability in an underground environment even without waste is important.

Embedding radioactive waste in bitumen has been used in immobilisa­tion since the 1960s and the total volume of RAW immobilised in bitumen currently exceeds 200,000 m3. In the bituminisation process, radioactive wastes are embedded in molten bitumen and encapsulated when the bitumen cools. Bituminisation combines heated bitumen and a concentrate of the waste material in either a heated thin film evaporator or extruder containing screws that mix the bitumen and waste. The waste is usually in the form of a slurry, for example salt aqueous concentrates or wet ion exchange resins. Water is evaporated from the mixture to about 0.5% mois­ture, intermixed with bitumen so that the final product is a homogeneous mixture of solids and bitumen, termed bitumen compound. Its retention properties usually exceed those of cements at higher waste loadings. Bitu- minisation is particularly suitable for water-soluble RAW such as bottom residues from evaporation treatment and spent organic ion exchangers. However, a drawback of bitumen is its potential fire hazard. The possibility of combustion in the case of an accidental fire has led to certain restrictions on the use of bitumen as an immobilising matrix.

Vitrification is an attractive immobilisation technique because of the small volume of the resulting waste form (Fig. 1.12b), the large number of elements which can be incorporated in it and its high durability. The high chemical resistance of glass allows it to remain stable in corrosive environments for long periods. Waste vitrification technology is a compro­mise between the desired durability of the final waste form and its process­ing efficiency (Ojovan and Lee, 2007). The most durable materials would require very high processing temperatures (>1500°C) which cannot be used because at high temperatures waste radionuclides occur in volatile species, generating large amounts of secondary wastes and diminishing the immobilisation efficiency. The most common glasses used in vitrification of nuclear waste are borosilicates and phosphates. Vitrification has been used for nuclear waste immobilisation for more than 40 years in France, Germany, Belgium, Russia, UK, Japan and the US. The total produc­tion of all vitrification plants by the end of 2000 was approximately 10,000 tonnes of radioactive glass in roughly 20,000 canisters. Vitrification is also currently used for immobilisation of low and intermediate level waste (LILW).

The highest degree of volume reduction and safety is achieved through vitrification, although this is the most complex and expensive method, requiring a relatively high initial capital investment. However, difficult legacy waste streams are known for which current technology is inadequate, so that new approaches must be developed. These comprise development of new waste forms such as crystalline ceramic and composite radionuclide hosts as well as of new immobilising technologies such as thermochemical and in-situ methods. New approaches aim also to create geochemically

stable materials in equilibrium with the disposal environment to ensure a safer nuclear waste disposal scenario.

Glass composite materials (GCMs) are used to immobilise glass — immiscible waste components such as sulphates, chlorides, molybdates and refractory materials requiring unacceptably high melting temperatures. GCMs comprise both vitreous and crystalline components (Lee et al., 2006). Depending on the intended application, the major component may be a crystalline phase with a vitreous phase acting as a bonding agent, or, alter­natively, the vitreous phase may be the major component, with particles of a crystalline phase dispersed in the glass matrix (see Fig. 1.13).

GCMs may be produced by dispersing both melted materials and fine crystalline particles in a glass melt and may be used to immobilise long-lived radionuclides (such as actinide species) by incorporating them into the more durable crystalline phases, whereas the short-lived radionuclides may be accommodated in the less durable vitreous phase. GCMs may also be glass ceramics where a glass is crystallised in a separate heat treatment step (Caurant et al., 2009), The French have developed a U-Mo GCM to immobilise Mo-rich HLW. Another example is the GCM developed to immobilise sulphur-enriched waste streams in Russia, containing


1.13 Examples of GCM microstructures.

conventional borosilicate glass vitreous phase with uniformly distributed particles comprising up to 15% by volume of yellow phase.

GCMs are being developed in many countries to immobilise their diffi­cult wastes. For example, the UK’s most hazardous wastes are those in the legacy ponds and silos (LP&S) at Sellafield. A number of novel thermal technologies are being examined to immobilise the complex, often ill — defined and heterogeneous wastes found in the LP&S. These include pyrol­ysis steam reforming, plasma vitrification and Joule heating in container melting (JHCM). In the latter process, mixed solids and sludge wastes are placed in a concrete lined steel container with embedded graphite elec­trodes in the corner (Fig. 1.14) and melted to produce a stable solid.

While JHCM can successfully convert reactive material (e. g., metals, sludges and organics) to more stable forms, the variable nature of the wastes makes control of process and product difficult, and it is difficult to charac­terise both the heterogeneous waste and product. Much R&D is needed including durability testing of the products of these technologies. However, their use has seriously reduced the hazard from the original wastes and pragmatic engineering approaches such as these are needed even if the resulting waste form is not as perfect as ultimately desirable.

Single-phase ceramics such as zircon (ZrSiO4) can potentially host a large number of nuclides and can be used as a monophasic waste form. However, monophase ceramics are difficult to fabricate and polyphase compositions are more common. The composition of the polyphase ceramic can host multiple radionuclides and be tailored to that of the waste composition to achieve complete and reliable immobilisation of the waste constituents. The most famous polyphase ceramic for nuclear waste immobilisation is Synroc. Synroc is short for ‘Synthetic Rock’, invented in 1978 by T. Ringwood of the Australian National University. Synroc is made of geochemically stable natural titanate minerals which have immobilised uranium and thorium for billions of years. U/Th-containing natural analogues of the basic constituent


(a) Before (b) During (c) After

1.14 ( a) In-container setup, (b) during heating, (c) resulting stable solid product.

of Synroc — zirconolites from Sri Lanka dating back 550 million years while amorphised — have nonetheless withstood the alteration processes of their natural environment.

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