Publications
273 results found
Ojovan MI, 2003, Effect of gas generation in matrices containing Ra-226 sources, Pages: 609-614, ISSN: 0272-9172
Spent sealed radiation sources containing Ra-226 are immobilised either by the embedding into metal matrix materials or by welding enclosure in metal hermetic capsules. Encapsulation of sources ought provide safe conditions for storage for a long time period. Gaseous products (helium and radon) are generated due to the decay of Ra-226. The gas generation of Ra-226 sources can cause their rupture and consequent leakages of radionuclides. The higher the radioactivity of sources the higher the intensity of gas generation. The pressure of accumulated gas depends on many other factors therefore necessary limitations ought be determined for the encapsulation procedures. The process of gas generation is described herein in order to reveal the conditions of safe immobilisation of radium sources. A simple model describing immobilisation of sources in a metal host is proposed. Both helium and radon generation is considered followed by accumulation of gases near sources. Partial pressures of helium and radon are found depending on time. The diffusion of helium in the metal and decay of Rn-222 are taken into account. It is shown that the pressures of both radon and helium increase until some maximum values, after that the pressures slowly fall down. The maximum radon pressure being not very high is achieved during 65 days after immobilisation whereas helium increases its pressure during an extended period of time lasting for tens of years (137 years in the glass matrix) reaching very high magnitudes for powerful sources (137.5 MPa for a 100 GBq source in a glass capsule). Safe conditions of immobilisation of radium sources can be determined on the base of obtained data.
Lifanov FA, Ojovan MI, Stefanovsky SV, et al., 2003, Cold crucible vitrification of NPP operational waste, Pages: 201-206, ISSN: 0272-9172
Operational radioactive waste is generated during routine operation of nuclear power plants (NPP). This waste must be solidified in order to ensure safe conditions of storage and disposal. Vitrification of NPP operational waste is a relative new solidification option being developed for last years. The vitrification technology comprises a few stages, starting with evaporation of excess water from liquid radioactive waste, followed by batch preparation, glass melting, and ending with vitrified waste blocks and some relative small amounts of secondary waste. Application of induction high frequency cold crucible type melters facilitates the melting process and significantly reduces the generation of secondary waste. Two types of glasses were designed in order to vitrify operational waste depending on the reactor type at the NPP. For the NPP with RBMK-type reactors the glass 16.2Na2O 0.5K2O 15.5CaO 2.5 Al2O3 1.7Fe2O3 7.5B 2O3 48.2SiO2 1.1 Na2SO4 1.2NaCl (5.7 others) was produced. For NPP with WWER reactors the glass 24.0Na2O 1.9K2O 6.2CaO 4.3Al2O3 1.8Fe2O3 9.0B2O3 46.8SiO 2 0.8Na2SO4 0.9NaCl (4.3 others) was produced. The melting temperatures of both glass formulations were 1200-1250 C, specific power consumption was 5.2 ± 0.8 kW h/kg, 137Cs loss was within the range 3 - 4 %. The specific radioactivity of glass reached 7.0 MBq/kg. Glass blocks obtained were studied both in laboratory and field conditions. Long-term studies revealed that vitrified NPP operational waste has the minimal impact onto environment. Since the glass has excellent resistance to corrosion it gives the basic possibility of maximal simplification of engineered barrier systems in a disposal facility. The simplest disposal option for vitrified NPP waste is to locate the packages directly into earthen trenches provided the host rock has the necessary sorption and confinement properties.
Bacon DH, Ojovan MI, McGrail BP, et al., 2003, Vitrified waste corrosion rates from field experiment and reactive transport modeling, Pages: 803-809
The calculation of vitrified waste corrosion rates from field experiment and reactive transport modeling is discussed. The low-activity wastes (LAW) are immobilized in glass and placed in a near-surface disposal system. Determination of the long-term release rates of radionuclides from LAW glasses is important for acceptance of the performance assessment (PA) by regulators and stakeholders. The reactive transport methodology models the physical and chemical processes that are responsible for controlling dissociation behavior of glass.
Karlina OK, Varlakova GA, Dmitriev SA, et al., 2003, Thermochemical conditioning of radioactive waste: Structure and properties of final processed product, Pages: 543-546
Thermochemical processing method is based on utilization of energy of chemical reactions between components of special exothermic mixtures (termed heat base) to melt radioactive waste and form a vitreous melt after cooling of which a durable monolith product is formed in which radionuclides are fixed. Compositions were studied, structure and properties of final products of thermochemical treatment of ash residue from incineration of solid radioactive waste, spent inorganic ion exchangers (e.g. clinoptililite and silica gel), contaminated clay and sand-based soils. Investigations showed that matrix material is dominantly amorphous and on complies to basic requirements to solidified radioactive waste of medium level of activity.
Ojovan MI, Karlina OK, Petrov GA, et al., 2003, Self-sustaining immobilisation processes, Proceedings of the International Conference on Radioactive Waste Management and Environmental Remediation, ICEM, Vol: 3, Pages: 1801-1806
An overview of self-sustaining immobilisation processes is given which describes also new thermochemical and radiogenic heat immobilising schemes based on utilization of both exothermal chemical reactions and radioactive decay energy.
Dmitriev SA, Karlina OK, Klimov VL, et al., 2003, Retention properties of carbide-corundum ceramics containing carbon-14, caesium-137, and strontium-90, Pages: 1293-1296
The system C - Al - TiO2 is of considerable interest for the processing of irradiated reactor graphite waste with the retention of biologic hazardous carbon-14. Investigations of this system were conducted both theoretically and experimentally. Previously, the thermodynamic calculations of the phase composition of resulting end product were performed for a wide variety of components content in the system being investigated. These simulation results have been supported by XRD-analysis of produced specimens. The experimental processing of reactor graphite was conducted by the use of self-sustaining reactions in C - Al - TiO2 mixtures. A search of modifier additives was performed to perfect end product properties. Test specimens were produced by mass ranging from 0.2 to 3 kg in the argon atmosphere. Various techniques were applied to characterize the produced specimens. The compressive strength of specimens of doped carbide-corundum matrices synthesized ranged from 7 to 18 MPa. The carry over of Cs-137 and Sr-90 during synthesis reaction was about 3 % wt. The teachability attained of Cs-137 and Sr-90 from specimens was around 10-5 g/(cm2.day). The carbon-14 is combined in the end product in chemically and thermic stable titanium carbide. The carry-over of the carbon combined in carbon monoxide from the reacting mixtures during exothermic process was less than 1% wt. This corresponds roughly to up 0.01% wt. of the carbon-14 inventory, which can be present in the irradiated reactor graphite.
Ojovan MI, Lee WE, 2003, Self sustaining vitrification for immobilisation of radioactive and toxic waste, Annual Meeting of the Society-of-Glass-Technology, Publisher: SOC GLASS TECHNOLOGY, Pages: 218-224, ISSN: 0017-1050
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Bergman GA, Karlina OK, Klimov VL, et al., 2003, Thermodynamic functions of zirconolite CaZrTi<inf>2</inf>O<inf>7</inf>, Atomnaya Energiya, Vol: 94, Pages: 234-236, ISSN: 0004-7163
The aim of the paper is to carry out the calculations of the thermodynamic functions (TF) of zirconolite CaZrTi2O7. The experimental values of TF obtained earlier and their extrapolated values are used in calculations. The values of the enthalpy, entropy, specific heat and reduced Gibbs energy are presented in the temperature range of 298-3000 K.
Karlina OK, Klimov VL, Pavlova GY, et al., 2003, Thermodynamic analysis and experimental investigation of phase equilibria in the thermochemical reprocessing of irradiated graphite in the C-Al-TiO <inf>2</inf> system, Atomic Energy, Vol: 94, Pages: 405-410, ISSN: 1063-4258
A thermodynamic analysis and laboratory investigations of the phase equilibria in the system C-Al-TiO 2 are performed. This system is of interest for immobilizing irradiated reactor graphite. The region of the relative content of the components of the system that allows for maximum inclusion of carbon in a carbide-corundum matrix in the final product is determined. It is found that in the process the carbon is partially volatilized in the form of CO. The amount of 14C escaping is estimated.
Bergman GA, Karlina OK, Klimov VL, et al., 2003, Thermodynamic functions of zirconolite CaZrTi <inf>2</inf>O <inf>7</inf>, Atomic Energy, Vol: 94, Pages: 193-195, ISSN: 1063-4258
The thermodynamic functions of zirconolite CaZrTi 2O 7 were investigated. A zirconolite sample with impurity content 8.24×10 -2 moles of ZrO 2 per 1 mole of CaZrTi 2O 7 was used. The data were approximated by a combination of Debye and Einstein functions and were used to calculate the specific heat and other thermodynamic functions up to 1500 K. The computed thermodynamic functions can be used in the thermodynamic analysis of the thermochemical processes in the immobilization of radionuclides with participation of zirconolite.
Hyatt NC, Lee WE, Hand RJ, et al., 2003, Vapor phase hydration of magnox waste glass, Warrendale PA, 26th symposium, scientific basis for nuclear waste management, Boston, MA, December 2002, Publisher: Materials Research Society, Pages: 183-188
Ojovan MI, Ojóvan NV, Startceva IV, et al., 2002, Modelling of the waste form behaviour in a wet near-surface repository site over extended time periods, Pages: 1647-1650
Waste form behavior in a wet near-surface repository site was discussed. A mathematical model was used to predict radionuclide release from bitumen and glass waste forms over extended time periods. Results showed that fmax = 2.3×10-3% of the initial radioactivity will release from the waste glass into the environment with a proposed period of 300 years.
Bergman GA, Navrotsky A, Ojovan MI, et al., 2002, Thermodynamic functions of zirconolite and their uses in computer simulation, Materials Research Society Symposium - Proceedings, Vol: 713, Pages: 365-371, ISSN: 0272-9172
The enthalpy of melting and the heat capacity of liquid zirconolite (CaZrTi2O7) are estimated as equal to 200 ± 20 kJ·mol-1 and 350 ± 50 J·mol-1·K-1, respectively. Thermodynamic functions of solid and liquid zirconolite are calculated based on these estimated data and the results of Navrotsky et al. On the basis of these thermodynamic functions, computational thermodynamic simulation is performed on the thermochemical synthesis of zirconolite-bearing materials. Demonstration indicates that synthesis of zirconolite-like matrix materials is possible using the self-sustaining exothermic reaction.
Ojovan MI, Ojovan NV, Golubeva ZI, et al., 2002, Aging of a bitumen waste form in wet repository conditions, Pages: 713-718, ISSN: 0272-9172
SIA 'Radon' has been performing field tests of bitumen waste forms holding low and intermediate level wastes (LILW) for about three decades. The waste forms were made at an industrial bituminization plant from the actual LILW wastes. This paper presents results from analyzing the samples of the bitumen waste material taken from the bulky bitumen block with waste salts loading 31 wt.% after storing the block in a shallow-ground repository for 12 years. Rich in natrium nitrate NPP-operational waste was incorporated in bitumen. Salts were separated from bitumen in some waste form samples. Non-homogeneous distribution of both salts and radionuclides was detected in vertical direction of the bitumen block. Specific radioactivity of the extracted bitumen was of the same order as the specific radioactivity of salts, in some cases even greater. Bitumen samples free of salts were further separated into three main bitumen fractions, asphaltenes, saturated and aromatic hydrocarbons, using methods of solvent extraction. Essentially all radioactivity of the bitumen was associated with the asphaltene fraction. Aging of the bitumen waste form led to increase in asphaltene fraction content (with minimum 4%) and hardening of the waste material. The study has revealed a significant transfer of the waste salts radioactivity to the asphaltene fraction of the bitumen matrix. Changes in the properties of the bitumen waste form will be taken into consideration in modeling the long-term behavior of the bitumen waste materials under repository conditions.
Ojovan MI, Karlina OK, Klimov VL, et al., 2001, Self-sustaining reactions for the processing technologies of chemically stable matrices incorporating carbon and zirconium wastes, Pages: 1035-1039
During operation of uranium-graphite reactors, waste graphite, containing fragments of nuclear fuel and fission products, as well as radioactive zirconium alloy components from fuel assemblies are produced. A large number of experiments should be carried out for the synthesis of appropriate matrix materials for radioactive nuclides that occur in these wastes. For the choice of processing technologies, an approach was used based on the thermodynamic simulation and application of self-sustaining reactions. A preliminary batch compaction and a hot pressing of the end product were not carried out. The end composite matrix product provides strong retention of the carbon-14 and other radionuclides. The processing technologies proposed are rather simple in implementation, can be realized without complex production equipment and energy supply.
Karlina OK, Varlackova GA, Ojovan MI, et al., 2001, Ash and soil conditioning using exothermic metallic compositions, Pages: 65-70, ISSN: 0272-9172
A thermochemical approach to conditioning ash residue that involves the use of exothermic metallic compositions (EMC) has been previously reported. EMC intermixed with ash residue at the appropriate ratios can sustain a glass forming reaction wave that produces monolith glass-like blocks. Herein, the thermochemical conditioning process is extended for conditioning contaminated clay soils. The results are reported for a study of the conditioning process and product materials to determine the optimal processing parameters and to characterize the product materials.
Ojovan MI, Ojovan NV, Startceva IV, et al., 2001, Simulation of the waste glass behavior in a loamy soil of the wet repository site, Pages: 837-842, ISSN: 0272-9172
A model developed for description of waste glass corrosion has been applied to assess the radionuclide release from real radioactive (intermediate level) vitrified material over extended storage periods. Field data generated during the long-term testing of the prototype waste glass packages were mathematically processed and the derived parameters used in model calculations. Regardless of the corrosive saturated conditions of the wet near-surface repository, the fairly high safety of trench disposal has been demonstrated for borosilicate glass containing real NPP-operational waste.
Karlina OK, Varlakova GA, Ozhovan MI, 2001, Conditioning of radioactive ash residue in the wave of solid-phase exothermal reactions, Atomnaya Energiya, Vol: 90, Pages: 38-43, ISSN: 0004-7163
Possibility is studied of using the heat of exothermal solid-phase reactions for conditioning the ash residue from incineration of solid radioactive waste. It is shown that the process of conditioning the ash residue with obtaining a final glass-like product may be realized without external heating sources. The conditioning is carried out in a special container-crucible intended for not only for process realization but and for the following storage or disposal of the final product.
Ojovan MI, Poluectov PP, 2001, Surface self-diffusion instability in electric fields, ISSN: 0272-9172
The plane form is the equilibrium one for surfaces of condensed matter. Deviations can be caused usually by crystal structure. Herein we will describe an effect of surface instability due to self-diffusion processes of atoms and molecules in the near surface electric field. Self-diffusion processes (as it was shown by Mullins) cause relaxation of any deviation (protuberance) from the plane form due to the increased concentration of surface atoms and its consequent smoothing. This process we studied for the case when there is an electric field near the surface. The near surface electric field can be due to either the location of material in an external (homogeneous or inhomogeneous) electrical field or self-charges on the surface. There is an increasing of electric field intensity near protuberances both in external and self-formed electrical fields: the higher is the curvature of surface the stronger is the intensity of the near surface electrical field. Consequently two competing processes occur during surface molecules mass transfer: both the self-diffusion smoothing of surface molecule concentration and drawing of molecules in the strong electric field regions. Depending on the initial shape of the protuberance either relaxation or instability occurs. There is a critical wavelength λ0=RskBT/2Uz, which shows that shorter wavelength deviations decrease their amplitudes and longer wavelength deviations grow in amplitude by time. Here Rs is characteristic of the material, T is temperature, and Uz is the interaction energy of surface molecules with the electric field. Since there are random variations of any surface from the plane form, being placed in an electric field these surfaces will be unstable depending on the intensity of electric field and properties of material.
Puzanov YV, Ojovan MI, Sobolev IA, 2001, Requirements for the radiation monitoring systems in highly populated areas, Pages: 589-591
A connection between radiation dose, arising radiation risk and parameters of radiation monitoring system is discussed in the paper. Quantitative values of parameters are offered. It is possible to use them for assessment of radioecological state of two areas with different density of population. It is shown that requirements for radiation monitoring system are more severe for area with the lower level of radioactive contamination and bigger population.
Barantseva GE, Kolotushkina SP, Ozhovan MI, et al., 2001, Some problems of ecological insurance of radioactive waste transportation (RWT), Atomnaya Energiya, Vol: 90, Pages: 43-49, ISSN: 0004-7163
A general scheme of approach to the conclusion of insurance treaties in the field of RWT is proposed. The algorithm of ecological insurance for radioactive waste management is developed. Recommendations are formulated for determining the accident probability during the RWT process. Proposals are developed on the conclusion of insurance treaties in dependence on the kind of transportable RW. The obtained developments are the basis for preparation of the documentation on the ecological insurance of RWT.
Sobolev IA, Ojovan MI, Karlina OK, 2001, Management of spent sealed radioactive sources at regional facilities "Radon" in Russia, Pages: 81-86
The management of spent sealed radioactive sources (SRS) in Russian Federation was discussed. This system involves 16 regional facilities known as 'Radon' which provide safe management of SRS. These facilities use the bore hole repositories technique to dispose the SRS. It was found that this operating system can solve basic problems and provide safe conditions for the environment.
Ojovan MI, Ojovan NV, Startceva IV, et al., 2001, Waste glass behavior in a loamy soil of a wet repository site, Journal of Nuclear Materials, Vol: 298, Pages: 174-179, ISSN: 0022-3115
Intermediate-level operational waste from a nuclear power plant was immobilized in borosilicate (BS) glass in a pilot vitrification plant. Glass blocks weighing in average 30 kg with a waste loading of 35 wt% and total α-, β-activity of about 3.75 × 106 Bq/kg were prepared and placed for testing in a near-surface repository and on an open site. Results of 12 years of exposure are presented. © 2001 Elsevier Science B.V. All rights reserved.
Karlina OK, Varlakova GA, Ozhovan MI, et al., 2001, Conditioning of radioactive ash residue in a wave of solid-phase exothermal reactions, Atomic Energy, Vol: 90, Pages: 43-48, ISSN: 1063-4258
The possibility is studied of using the heat of exothermal reactions in the solid phase for conditioning the ash residue produced when solid radioactive wastes are burned. It is shown that solidifcation of the ash wastes can be performed without input of energy from external heat sources with the final glassy product containing 50-60 mass % ash residue and meeting the requirements for solidified radioactive wastes. The conditioning is performed in a special crucible-container, intended not only for performing the process but also for subsequent storage or burial of the final product. 5 figures, 2 tables, 10 references.
Ojovan MI, Klimov VI, Petrov GA, et al., 2001, Thermochemical approach on treating spent ion exchange resins, Materials Research Society Symposium - Proceedings, Vol: 663, Pages: 19-26, ISSN: 0272-9172
A thermochemical approach was suggested for treating and conditioning specific streams of radioactive wastes for example spent ion exchange resins (IER), mixed, organic or chlorine-containing radioactive waste (PVC, etc.). Conventional thermal treatment of such waste encounters serious problems concerning Complete destruction of organic molecules and possible emissions of radionuclides, heavy metals and chemically hazardous species. The thermochemical treatment uses powdered metal fuels (PMF) that are specifically formulated for the waste composition and react chemically with the waste components. The composition of the PMF is designed in such a way as to minimize the release of hazardous components and radionuclides in the off gas and to confine the contaminants in the ash residue. The thermochemical procedures allow decomposition of organic matter and capturing hazardous radionuclides and chemical species simultaneously. Previous preliminary work demonstrated the feasibility of applying the thermochemical approach to treatment of spent IER using PMF. Herein, the results are presented of theoretical and experimental studies to define the optimal PMF composition as well as the PMF/waste ratio.
Korobkov VI, Golubtsov IV, Shunipova TI, et al., 2001, Autoradiographic method for the determination of radioactivity of the soil from place of the radioactive wastes burial, Vestnik Moskovskogo Universiteta Seriya 2 Khimiya, Vol: 42, Pages: 135-136, ISSN: 0579-9384
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Barantseva GE, Kolotushkina SP, Ozhovan MI, et al., 2001, Certain questions concerning the ecological insurance for shipping radioactive wastes, Atomic Energy, Vol: 90, Pages: 49-54, ISSN: 1063-4258
A general scheme is proposed for making insurance agreements and an algorithm for ecological insurance for handling radioactive wastes is developed. It is recommended that the IAEA information be used for determining the probability of an accident during shipment of radioactive wastes and that the total yearly shipment mileage be taken into account. Suggestions are made for making insurance agreements depending on the type of wastes shipped. These developments are the basis for preparing methodological documentation on ecological insurance for handling radioactive wastes. 1 figure, 3 tables, 10 references.
Ojovan MI, Karlina OK, Klimov VL, et al., 2000, Graphite processing with carbon retention in a waste form, Pages: 565-570, ISSN: 0272-9172
Conversion of waste graphite into a stable waste form acceptable for long term storage and disposal was considered both theoretically and experimentally. A self-sustaining transformation process of graphite composited with suitable precursors was studied. The powdered precursors that were used were used were: Al+SiO2 (1), Al+TiO2 (2) and Ti+SiO2 (3). Numeric thermodynamic simulation was performed. Equilibrium temperatures and chemical compositions of reaction products were determined for a wide range of component ratios in the source mixtures. The highest temperatures (up to 2300 K) were observed for precursor type (2). Precursor type (3) demonstrated a minimal rise of temperature of up to 1900 K. Regions of compositions with complete binding of all chemical elements as well as production of stable final products were found to be rather narrow. About 10-13 wt.% of carbon can be processed in composition with given precursors. The gas phase reaction products were studied to minimize carry over of radionuclides. Carbon monoxide was shown to be the main component of the gas phase. The self-sustaining synthesis process was conducted in ceramic crucibles at ambient pressure in an air atmosphere. Batch masses ranged between 0.1-1 kg. Best results were obtained for processing of graphite composited with Al and TiO2. XRD analysis has identified titanium carbide and corundum in the waste form produced. These experiments confirmed that carbon can be converted completely into a stable waste form.
Manykin EA, Ojovan MI, Poluektov PP, 2000, Condensed states and their decay in a system of excited cesium atoms, Chemical Physics Reports, Vol: 18, Pages: 1353-1380, ISSN: 1074-1550
Condensed excited states in a system of excited centers - atoms, molecules, and admixtures in solids, are considered. Particular attention is paid to experimental and theoretical data on condensed excited states of cesium atoms. Experimental observations of these states are depicted at length and the results of calculations of equilibrium condensate parameters are presented. The decay of condensed excited states due both to intrinsic (radiative and radiationless) recombination and admixture recombination is described. The nature of an abnormally long lifetime of the condensed state excited to high levels is discussed.
Sobolev IA, Dmitriev SA, Ozhovan MI, et al., 2000, Investigation of the properties of uranium-containing composites produced in chemical reactor with using the powder-like metallized fuel, Fizika i Khimiya Obrabotki Materialov, Pages: 61-68, ISSN: 0015-3214
Immobilization of uranium, simulating actinide component of radioactive wastes (RAW) to a large-sized mineral blocks using the powder-like high-energy metallized fuel has been investigated. Chemical reactor on the base of steel container as well as RAW immobilization technology were developed. The composites obtained give a sure radionuclides immobilization. Interphase distribution of U and Hf in polymineral block is studied.
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