B. I. Omel’yanenko

Structure of Borosilicate Glassy Materials with High Concentrations of Sodium, Iron, and Aluminum Oxides

Alkali borosilicate glassy materials, which contain high iron and aluminum oxide concentrations and simulate vitrified high-level wastes of the Savannah River Site (United States), are investigated using X-ray powder diffraction, optical and electron microscopies, and infrared spectroscopy. The materials prepared by induction melting in cold crucibles operating in pilot and industrial facilities at the State Unitary Enterprise “Moscow Research and Production Association Radon” consist of a glass matrix with distributed individual or aggregated crystals of spinel similar in composition to trevorite. The maximum content of the crystalline phase in the glassy material from a “dead volume” of the cold crucible with an industrial size reaches ∼13 vol %. The texture of the glass phase is complex and determined by the direction of flows in cold crucibles under the action of eddy currents, the character of outflow of the glass melt stream during pouring into canisters, and the interaction of the stream with the glass solidified in the canister after preceding pourings. The structure of the anionic motif of the glass phase is predominantly built up of metasilicate chains and boron-oxygen fragments with threefold-coordinated boron.

Murataite as a universal matrix for immobilization of actinides

A new variety of matrices based on synthetic phases whose structure is close to that of murataite (a natural mineral) is proposed for immobilization of nuclear wastes. Murataite is Na, Ca, REE, Zn, and Nb titanate with a structure derived from the fluorite lattice. This very rare mineral was found in alkali pegmatites from Colorado in the United States and the Baikal region in Russia. The synthetic murataite-like phases contain manganese instead of zinc, as well as actinides and zirconium instead of sodium, calcium, and niobium. Varieties with threefold, as in the mineral, and five-, seven-, and eightfold repetition of the lattice relative to the fluorite cell have been established. Correspondingly, the structural varieties M3, M5, M7, and M8 are recognized among the synthetic murataites. A decrease in the contents of actinides, rare earth elements, and zirconium occurs in the series M7-M5-M8-M3, along with enrichment in Ti, Al, Fe, and Ga. Murataite-based ceramics are characterized by high chemical and radiation stability. The rate of U, Th, and Pu leaching with water at 90°C in static and dynamic tests is 10−6–10−5 g/m2 per day. These values are lower than the leaching rate of other actinide confinement matrices, for example, zirconolite-or pyrochlore-based. Murataite is close to other titanates in its radiation resistance. At 25°C, amorphization of its lattice is provided by a radiation dose of 2 × 1018 α decays/g, or 0.2 displacements/atom. Murataite-based matrices are synthesized within a few hours by cold compacting combined with sintering at 1300°C or by melting at 1500–1600°C and subsequent crystallization. The melting technology, including induction smelters with a cold crucible, makes it possible to produce samples with zonal murataite grains. The inner zone of such grains is composed of structural variety M5 or M7; the intermediate zone, of M8; and the outer zone, of M3. The contents of actinides, zirconium, and rare earth elements reach a maximum in the inner zone and drop to a minimum in the outer zone, while the amounts of nonradioactive elements—Ti, Al, Fe, and Ga—vary conversely. The U, Th, and Pu contents in the inner and outer zones differ by three to five times. Such a distribution precludes removal of actinides by interaction of the matrix with solution after its underground disposal. Individual actinides (Np, Pu, Am); the actinide-zirconium-rare earth fraction of high-level radioactive wastes (HLW); Am-Ga residues of weapons plutonium reprocessing with its conversion into U-Pu mixed oxide (MOX) fuel; and other sorts of HLW enriched in actinides, REE, and products of corrosion (Mn, Fe, Al, Zr) can be incorporated into a murataite-based matrix. As much as 350 kg of HLW components can be included in 1 t of such a ceramic. An actinide matrix that is composed of titanates with a pyrochlore structure is its nearest analogue. The advantage of murataite in comparison with pyrochlore consists in its universal character; i.e., a murataite-based matrix can be used for utilization of a wider range of actinide-bearing highly radioactive wastes.

Murataite matrices for actinide wastes

Matrices for actinide wastes, consisting of complex murataite-type oxides, were studied. The samples were prepared by sintering of the oxide charge at 1100–1400°C or by fusion at 1450–1600°C followed by crystallization. Along with the structural analog of the natural mineral (murataite 3C), appreciable role in the samples is played by the phases (hereinafter murataites 5C, 7C, and 8C) consisting of pyrochlore and murataite 3C units. The fraction of wastes in the samples is about 10 wt %, which is close to the values for the pyrochlore matrix for Pu immobilization. In the ceramics prepared from the melt, the murataite grains have a zonal structure. They are built of murataite 3C at edges and murataite 5C (rarely 7C) in the center where the actinide content is maximal. This fact accounts for their high capability to isolate radionuclides. Amorphization of the structure only slightly affects the stability of murataite in solution. The optimal procedure for the industrial production of the matrices is their melting by induction heating in cold crucible (IMCC) and crystallization.

Influence of the content of a surrogate of iron aluminate high-level wastes on the phase composition and structure of glassy materials for their immobilization

Alkali borosilicate glassy materials with high iron and aluminum oxide concentrations, which simulate vitrified high-level wastes from the Savannah River Site (United States) at their content ranging from 50 to 70 wt %, have been investigated using X-ray powder diffraction, optical and electron microscopy, and infrared spectroscopy. Quenched and slowly cooled samples containing 50 wt % wastes are glasses. Samples containing 60 and 70 wt % wastes, which were quenched on a metal slab, are predominantly glasses with an insignificant content of the spinel formed in a trevorite-magnetite solid solution. The slowly cooled samples also contain nepheline, and its amount increases with an increase in the waste content in the glassy materials.

Isolation of long-lived technetium-99 in confinement matrices

Amongst fission products formed in atomic reactors, 99Tc is the most hazardous for the environment because of its long half-life (213000 yr), high content in spent nuclear fuel (SNF) (0.8–1.0 kg per ton of SNF), low sorption ability, and high mobility under aerobic conditions. The bulk of 99Tc (∼200 t) is incorporated into SNF. In the course of SNF reprocessing, this radioisotope is released as a separate fraction or along with actinides. More than 60 t of highly concentrated 99Tc have been accumulated to date. It is evident that isolation of 99Tc from the environment is a matter of great urgency. The immobilization of technetium in a highly stable and poorly soluble matrix is a necessary element in settling this problem. Ceramics composed of titanates with pyrochlore, perovskite, and rutile structures are proposed as matrices able to retain technetium along with actinides. The high chemical stability of these compounds has been corroborated by experiments. The difficulties in production of such matrices are related to the fugacity of Tc and the necessity of converting it into Tc(IV). To overcome this obstacle, self-propagating high-temperature synthesis (SHS), characterized by reductive conditions and a high reaction rate, is proposed. The charge for matrix synthesis consists of reducing agents (metallic powders with a strong affinity to oxygen, e.g., Ti and Zr), oxidants (MoO3, Fe2O3, CuO), and additives (TiO2, ZrO2, Y2O3, CaO, etc.), which taken together with other elements form target phases. Instead of Tc, Mo, close in chemical properties, is used in matrix synthesis as a simulator. Samples of Mo-bearing matrices have been synthesized with SHS; their phase compositions and Mo distribution therein are characterized. It has been shown that up to 40 wt % Mo can be incorporated into the synthesized matrices in the form of metal or structural admixtures in titanates. The titanate-zirconate pyrochlore-based matrices are the most appropriate for the joint immobilization of actinides, REEs, and 99Tc.

Self-propagating high-temperature synthesis and characteristics of cermet matrices for isolation of wastes with long-lived radionuclides

The possibility of preparing by self-propagating high-temperature synthesis (SHS) metal-ceramic (cermet) matrices with simulated wastes of REE-actinide fraction and Tc was examined. The specimens consist of oxide crystalline phases, glass, and melts. In the aluminate composite, the component (Sm) simulating the REE-actinide fraction is in the garnet and glass phases, and in the titanate composite, in the pyrochlore, titanosilicate of perrierite structure, and glass phases. Rhenium (Tc simulator) is incorporated in alloy phases. To evaluate the prospects for radioactive waste immobilization by SHS, it is necessary to synthesize matrices containing actinide isotopes (Am) and Tc and to study their structure and isolation properties.

Behavior of uranium under conditions of interaction of rocks and ores with subsurface water

The behavior of uranium during interaction of subsurface water with crystalline rocks and uranium ores is considered in connection with the problem of safe underground insulation of spent nuclear fuel (SNF). Since subsurface water interacts with crystalline rocks formed at a high temperature, the mineral composition of these rocks and uranium species therein are thermodynamically unstable. Therefore, reactions directed toward the establishment of equilibrium proceed in the water-rock system. At great depths that are characterized by hindered water exchange, where subsurface water acquires near-neutral and reducing properties, the interaction is extremely sluggish and is expressed in the formation of micro- and nanoparticles of secondary minerals. Under such conditions, the slow diffusion redistribution of uranium with enrichment in absorbed forms relative to all other uranium species is realized as well. The products of secondary alteration of Fe- and Ti-bearing minerals serve as the main sorbents of uranium. The rate of alteration of minerals and conversion of uranium species into absorbed forms is slow, and the results of these processes are insignificant, so that the rocks and uranium species therein may be regarded as unaltered. Under reducing conditions, subsurface water is always saturated with uranium. Whether water interacts with rock or uranium ore, the equilibrium uranium concentration in water is only ≤10−8 mol/l. Uraninite ore under such conditions always remains stable irrespective of its age. The stability conditions of uranium ore are quite suitable for safe insulation of SNF, which consists of 95% uraninite (UO2) and is a confinement matrix for all other radionuclides. The disposal of SNF in massifs of crystalline rocks at depths below 500 m, where reducing conditions are predominant, is a reliable guarantee of high SNF stability. Under oxidizing conditions of the upper hydrodynamic zone, the rate of interaction of rocks with subsurface water increases by orders of magnitude and subsurface water is commonly undersaturated with uranium. Uranium absorbed by secondary minerals, particularly by iron hydroxides and leucoxene, is its single stable species under oxidizing conditions. The impact of oxygen-bearing water leads to destruction of uranium ore. This process is realized simultaneously at different hypsometric levels even if the permeability of the medium is variable in both the lateral and vertical directions. As a result, intervals containing uranyl minerals and relics of primary uranium ore are combined in ore-bearing zones with intervals of completely dissolved uranium minerals. A wide halo of elevated uranium contents caused by sorption is always retained at the location of uranium ore entirely destroyed by weathering. Uranium ore commonly finds itself in the aeration zone due to technogenic subsidence of the groundwater table caused by open-pit mining or pumping out of water from underground mines. The capillary and film waters that interact with rocks and ores in this zone are supplemented by free water filtering along fractures when rain falls or snow is thawing. The interaction of uranium ore with capillary water results in oxidation of uraninite, accompanied by loosening of the mineral surface, formation of microfractures, and an increase in solubility with enrichment of capillary water in uranium up to 10−4 mol/l. Secondary U(VI) minerals, first of all, uranyl hydroxides and silicates, replace uraninite, and uranium undergoes local diffusion redistribution with its sorption by secondary minerals of host rocks. The influx of free water facilitates the complete dissolution of primary and secondary uranium minerals, the removal of uranium at the sites of groundwater discharge, and its redeposition under reducing conditions at a greater depth. It is evident that the conditions of the upper hydrodynamic zone and the aeration zone are unfit for long-term insulation of SNF and high-level wastes because, after the failure of containers, the leakage of radionuclides into the environment becomes inevitable.

Glasses for immobilization of low- and intermediate-level radioactive waste

Reprocessing of spent nuclear fuel (SNF) for recovery of fissionable elements is a precondition of long-term development of nuclear energetics. Solution of this problem is hindered by the production of a great amount of liquid waste; 99% of its volume is low- and intermediate-level radioactive waste (LILW). The volume of high-level radioactive waste (HLW), which is characterized by high heat release, does not exceed a fraction of a percent. Solubility of glasses at an elevated temperature makes them unfit for immobilization of HLW, the insulation of which is ensured only by mineral-like matrices. At the same time, glasses are a perfect matrix for LILW, which are distinguished by low heat release. The solubility of borosilicate glass at a low temperature is so low that even a glass with relatively low resistance enables them to retain safety of under-ground LILW depositories without additional engineering barriers. The optimal technology of liquid confinement is their concentration and immobilization in borosilicate glasses, which are disposed in shallow-seated geological repositories. The vitrification of 1 m3 liquid LILW with a salt concentration of ∼300 kg/m3 leaves behind only 0.2 m3 waste, that is, 4–6 times less than by bitumen impregnation and 10 times less than by cementation. Environmental and economic advantages of LILW vitrification result from (1) low solubility of the vitrified LILW in natural water; (2) significant reduction of LILW volume; (3) possibility to dispose the vitrified waste without additional engineering barriers under shallow conditions and in diverse geological media; (4) the strength of glass makes its transportation and storage possible; and finally (5) reliable longterm safety of repositories. When the composition of the glass matrix for LILW is being chosen, attention should be paid to the factors that ensure high technological and economic efficiency of vitrification. The study of vitrified LILW from the Kursk nuclear power plant with high-power channel reactors (HPCR; equivalent Russian acronym, RBMK) and the Kalinin nuclear power plant with pressurized water reactors (PWR; equivalent Russian acronym VVER) after their 14-yr storage in the shallow-seated repository at the MosNPO Radon testing ground has confirmed the safety of repositories ensured by confinement properties of borosilicate matrix. The most efficient vitrification technology is based on cold crucible induction melting. If the content of a chemical element in waste exceeds its solubility in glass, a crystalline phase is formed in the course of vitrification, so that the glass ceramics become a matrix for such waste. Vitrified waste with high Fe; Na and Al; Na, Fe, and Al; Na and B is characterized. The composition of frit and its proportion to waste depends on waste composition. This procedure requires careful laboratory testing.

Influence of the content of radioactive wastes with high concentrations of aluminum, sodium, and iron oxides on the phase composition and structure of glassy materials prepared in a “cold crucible”

The vitrification of a high-level waste surrogate with high concentrations of aluminum, sodium, and iron oxides in a “cold crucible” results in the formation of glassy materials with the phase composition and structure dependent on the ratio between waste oxides and borosilicate glass frit. With an increase in the waste content from ∼50 to ∼66 wt %, the degree of crystallinity of the materials increases from ∼5 to 50 vol %. The main crystalline phase is spinel, and the additional crystalline phase is nepheline. The amount of the nepheline increases with an increase in the waste content in glassy materials and with a decrease in the rate of their cooling in heat-insulated molds. The spinel is the main concentrator of transition elements (Cr, Mn, Fe, Ni, Cu, Zn). A considerable fraction of sodium, aluminum, and silicon transfers into the formed nepheline, which leads to the depletion of the glass phase in oxides of these elements and to a decrease in its chemical durability.

Matrices for isolation of long-lived radionuclides

Safe isolation of highly radioactive waste is a most important condition of successful development of nuclear power engineering. The main source of such waste is facilities that process irradiated nuclear fuel. This fuel is processed for extracting fissionable actinide isotopes for their reuse and reducing waste volume. The principal difficulties of safe isolation of highly radioactive waste are related to the presence of