The thermal decomposition of nanosized aluminum hydroxides, which are precipitated from solution in direct and inverse mixing of the chemical agents, is studied. The thermal treatment of hydroxides between 370 and 870 K results in their two-stage thermal decomposition. Intermediate decomposition products are amorphous aluminum oxyhydroxide AlO(OH) 1+x and boehmite. Nanosized γ-Al 2 O 3 begins forming even at 570 K. The thermal decomposition of amorphous oxyhydroxide AlO(OH) 1+x ceases when amorphous alumina forms at 770-870 K. This results in powders representing a mixture of amorphous and γ-alumina.

Low-temperature sintering of a tetragonal zirconia solid solution proceeds through the fracture of all agglomerates during pressing of samples from hydroxide powder coprecipitated from an aqueous solution and through the increased reactivity of amorphous zirconium hydroxide and oxide. Thermal treatment at 1100ºC for 1 h produces ceramics with relative density 0.928, grain size 120-135 nm, and pore size 50-75 nm. Sintering is most intensive in the temperature range 950-1150ºC and is less active in the range 800-950ºC.

The behavior of coarse and fine pore channels and closed pores (pore space components) in temperature ranges of intensive (900-1150°C) and less active (1150-1400°C) sintering of a sample pressed from nanosized powder of tetragonal zirconia solid solution is examined. The change in the volume of the sample in the temperature range of intensive sintering is determined by approximately equal decrease in the volume of coarse and fine pore channels, which originate from channels between agglomerates and between aggregates in the powder agglomerates. Closed pores of the sample originating from closed pores of the powder make a minor contribution to the decrease in volume since they contain air. In the temperature range of less active sintering, all fine (200-300 nm) and coarse (0.8-1 μm) pore channels are broken and new closed pores are formed.

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This paper describes the research work carried out at the NSC KIPT to develop and apply a final waste form in the shape of a monolithic solid block for the containment of spent nuclear fuel. To prepare radioactive waste for long-term storage and final deep geological disposal, investigations into the development of methods of immobilizing HLW simulators in protective solid matrices are being conducted at the NSC KIPT. For RBMK spent nuclear fuel it is proposed and justified to encapsulate the spent fuel bundles into monolithic protective blocks, produced with the help of hot isostatic pressing (HIP) of powder materials. In accordance with this approach, as a material for the protective block made up of the glass-ceramic composition prepared by sintering at isostatic pressure, the powder mixture of such natural rocks as granite and clay has been chosen. Concept approach and characterization of waste form, technological operations of manufacturing and performance assessment are presented. The container with spent fuel for long-term storage and final disposal presents a three barrier protective system: ceramic fuel UO2 in cladding tube, material of the glass-ceramic block, material of the sealed metal capsule. Investigations showed that the produced glass-ceramic material is characterized by high stability of chemical and phase compositions, high resistance in water medium, low porosity (compared with the porosity of natural basalt). With the help of mathematical calculations it was shown that the absorbed dose of immobilizing material by RBMK spent fuel irradiation for 1000 years of storage in the geological disposal after 10 years of preliminary cooling will be ∼ 3.108 Gy, that is 2–3 orders of magnitude less than the values corresponding to preserving radiation resistance and functional parameters of glasses and ceramics. The average value of velocity of linear corrosion in water medium of the protective layer made up of the glass-ceramic composition determined experimentally makes up ∼ 15 mm per year. This allows to use glass-ceramic compositions effectively as an engineering barrier in the system of spent fuel geological disposal and to increase the lifetime of the waste container, in particular, up to 3000 years with the layer thickness ∼ 40 mm. The possible release of radionuclides from the waste container during its interim storage in the open air (near-surface storage) is estimated. The calculations are made by taking into account the possible increase of coefficients of radionuclide diffusion from 10−16 to 10−14 m2/c as a result of spent fuel radiation affecting the protective layer. The obtained results showed that the protective barrier (about 40 mm) at the base of the glass-ceramic composition, ensures reliable isolation from the environment against the release of radionuclides from the controlled near-surface long-term storage far up to 1000 years. The relatively limited release of radionuclides will make up about 1% for the period of more than 400 years, and 10% - in 1000 years. During this period of time, the radionuclides 90Sr and 137Cs will completely turn into stable 90Zr and 137Ba and the decay of many transuranium elements will occur. The results from laboratory scale experiments, tests and calculations carried out so far, show that the proposed glass-ceramic materials may be used as basic materials for manufacturing the monolithic protective block in which the spent fuel elements will be embedded with the aim of further long-term storage or final disposal.