Vitrified waste

Deglor [Detoxification and glassification of residues] A process for vitrifying wastes, such as fly ash, by heating to 1,400°C in an electric furnace. Some of the heat is provided by radiant heaters, some by passage of electricity through the melt. Developed and piloted by ABB in Switzerland from the 1980s, commercialized in Japan in 1996. 

According to Robert T. Mueller of the State of New Jersey Department of Environmental Protection (NJDEP), demonstrations conducted by the vendor have supported the premise that the cost to vitrify waste into nonleachable materials is substantially lower than the excavation and relocation of these materials to regulated landfills as a means of permanent disposal (D17164M,... 

The preparation, composition, structure and leaching characteristics of a crystalline, ceramic radioactive waste form have been discussed, and where applicable, compared with vitrified waste forms. The inorganic ion exchange materials used such as sodium titanate were prepared from the corresponding metal alkoxide. The alkoxides were reacted in methanol with a base containing the desired exchangeable cation and the final powder form was produced by hydrolysis in an acetone-water mixture followed by vacuum drying the precipitate at ambient temperature. 

The leaching characteristics of the ceramic waste were compared with some vitrified waste forms under identical experimental conditions. The results are complex due to different leaching rates for various elements and the attack on the glass surface which results in the formation of a surface film which periodically sloughs off, however, some general comparison can he made. 

The use of inorganic ion exchangers to solidify liquid radioactive waste followed by pressure sintering to produce a ceramic waste form appears to be a viable alternative to calcina-tion/vitrification processes. Both the process and waste form are relatively insensitive to changes in the composition of the waste feed. The stability of the ceramic waste form has been shown to be superior to vitrified wastes in leaching studies at elevated temperatures. Further studies on the effects of radiation and associated transmutation and the influence of temperature regimes associated with potential geologic repositories are needed for a more definitive comparison of crystalline and amorphous waste forms. 

Figure 16.13 Schematic diagram of the final steps in putting vitrified waste into a geologic repository. (Figure also appears in color figure section.)...

WASTOXHAS is the acronym for WASte ecoTOXic Hazard Assessment Scheme. This method was developed to ensure that unacceptable adverse effects would not arise from landfilled or re-used waste disposal. It is dedicated to assess the long-term leaching hazardous impact of any solid waste containing potentially hazardous substances (e.g., bulk, stabilized, solidified, or vitrified wastes as well as contaminated soils or sediments intended for soil disposal). 

It is considered a form of HLW because of the uranium, fission products, and transuranics that it contains. HLW includes highly radioactive liquid, calcined or vitrified wastes generated by reprocessing of SF. Both SF and HLW from commercial reactors will be entombed in the geological repository at Yucca Mountain —100 mile (1 mile = 1.609344 km) northwest of Las Vegas, Nevada. Disposal of spent nuclear fuel and HLW in the US is regulated by 40 CFR Part 191 (US EPA, 2001) and 10 CFR Part 60 (US NRC, 2001). It is discussed in more detail in a later section of this chapter. 

The waste container is expected to be a steel box approximately 3 m (10 ft) high, 2.4 m (8 ft) wide, and 7.3 m (24 ft) long. Before waste is placed in it, the container will be lined with insulating board, sand, and a layer of castable refractory. The refractory will be in contact with the waste. A layer of melt-initiating graphite and soil will be placed over the refractory in the bottom of the container. The container will have one or more ports for sampling the vitrified waste after it has cooled. 

Figure 11.9 Long-term leaching curves of a vitrified waste cylinder calculated according to different kinetic models [initial leach rate 2 X 10 g/(cm day)). (From Ewest[ElJ.)...

The conceptual design of the repository (Figure 2) consists of a series of parallel tunnels, where the wastes would be emplaced in boreholes excavated in the floors of the gallery. The centreline distance between adjacent tunnels is 10 m and the centreline distance between adjacent inground boreholes for the wastes is 4.44 m. The depth of each borehole is 4.13 m and the diameter is 2.22 m. The overpack for vitrified wastes would be emplaced into the borehole, and a bentonite buffer material would be compacted around the overpack. The tunnels would also be backfilled with a mixture of gravel and clay. 

The engineered barriers are composed of vitrified waste encapsulated in a canister, metal overpack and buffer material in a geological disposal system of high-level radioactive waste (HLW) adopted by INC (2000). Highly compacted bentonite is considered as the candidate buffer material for the system. Figure I shows the schematic view of the geological disposal system and expected processes after emplacement of the engineered barriers. 

In the near-field of a HLW repository, the coupled thermo -hydro -mechanical and chemical (T-H-M-C) processes will occur, involving the interactive processes among radioactive decay heat from the vitrified waste, infiltration of groundwater into bentonite, swelling pressure of bentonite due... 

Long-term studies performed by the CEA (French Atomic Energy Commission) resulted in the formulation of the so-called R7 and T7 glasses. Characterization studies have verified the main properties of the final products and have established the glass specifications. The vitrified waste specifications were subjected to peer review by an independent commission of nuclear specialists. 

The Vitrification Plant processes HAL into a solid vitrified waste form that is then stored in a dedicated storage facility. The vitrification plant process is described below, considering each of the main cells in the plant in sequential process order ... 

The vitrified waste canister assay system (VCAS) is intended to determine the residual uranium and plutonium content in canisters of vitrified high-level spent-ffiel reprocessing waste prior to the termination of safeguards on this material. It consists of five neutron detectors (two fission chambers, two U chambers and a bare chamber sensitive to thermal neutrons) and one gamma detector (ionization chamber, meant to authenticate the presence of gamma radiation). In contrast to the VWCC, the VCAS uses fission chambers... 

Each of these elements may be used for production of nuclear fuel or other purposes. The recovery efficiency for uranium is reported as 99.87% and for plutonium 99.36%-99.51% (NEA 2012). The extended PUREX includes separation of neptunium and technetium as well as recovery of americium and curium that are also separated from each other by additional extraction stages as given in detail in the flowsheet (NEA 2012). The advanced UREX-i-3 process generates six streams after separation uranium for re-enrichment Pu-U-Np for mixed oxide fuel c for managed disposal Am-Cm to be used as burnable poisons and for transmutation high-heat-generating products (Cs and Sr) and a composite vitrified waste with all other fission products. Some fuel types may require preliminary steps like grinding to enable their dissolution. 

Examples of impact assessments taken from the European Union Everest exercise give an idea of these variations for an hypothetical repository containing vitrified waste in a granite formation. Figure 1 summarizes, for the normal scenario, the variation of potential individual doses due to different radionuclides, when varying hydraulic parameters of the geosphere, retardation coefficients, matrix solubility, and radionuclide solubilities. 

Many of the nuclear waste disposal concepts advocate the use of bentonite as a buffer between the host rock and the metal overpack, which contains HLW, i.e., the spent fuel or the vitrified waste, but little is known about the long-term behavior of bentonite. 

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