Using through diffusion method at room temperature, migration of simulants of RW elements (P, Se, Br, Mo, Cs, U) in compacted samples of clay materials of various mineral compositions was studied during porous diffusion from model solutions: underground water and leachates of phosphate waste forms having a total salt content of up to 500 mg/L. Based on the results of experiments, effective diffusion coefficients and sorption distribution coefficients of elements in barrier materials were determined. Numerical models are proposed to describe diffusion transfer of selenium, cesium, and uranium depending on porosity, mineral composition of materials, and concentration of elements in pore solution. Patterns of diffusion of elements from solutions of different salt composition were revealed.

A method of coupled processes was proposed to maintain concentrations in the model leachate of the radioactive waste phosphate matrix, which served as a source of elements in the study of through-diffusion of P, Se, Br, Mo, Cs, and U in the pore solution of compacted clay materials. The method consisted in adding an leachatable solid phase to the solution in the source chamber of the diffusion cell. The use of this method made it possible to stabilize the boundary conditions and expand the range of element concentrations in the source chamber of diffusion cells. The new as-obtained data on the effective diffusion coefficients of radioactive waste elements in clay rocks were used to refine the empirical models of diffusion transfer. It is shown that in different geochemical systems (model groundwater and phosphate glass leachate) for some elements (Br, Mo, Cs) it is possible to use unified models in the form of effective diffusion coefficients as a function of factors influencing this process: sample porosity, smectite content in the sample, and concentration of radionuclide (element) in pore solution, while for Se and U, diffusion models for various geochemical systems differ. The specificity of diffusion behavior of elements is associated with structural features and physicochemical properties of particles of these elements in aqueous solutions.

The article reflects the problem of graphite-containing sludge accumulated, as deposit, in technological systems during the operation of uranium-graphite reactors. Technical and methodological approaches to the performance of work on the survey of reservoirs are presented, which made it possible to remotely select highly active samples of graphite-containing sludge from closed reservoirs located in radiation-hazardous rooms, and to estimate the volume, layer thickness, and surface profile of deposits. Methods are described, including radiochemical ones, used in the course of research to determine the radionuclide composition, activity levels, amounts of solid and liquid phases, moisture content, and mass loss on ignition. To predict the behavior of radionuclides sorbed on deposits, the selective desorption method was used to determine the water-soluble, ion-exchange, low-mobility, strongly bound, and strongly fixed forms of radionuclides in samples of graphite-containing sludge. The results obtained were used to assess nuclear and radiation safety, as well as to develop technological methods for handling graphite-containing sludge during the decommissioning of reservoirs. Proven approaches to survey and laboratory analysis of sediments can be used in the inspection of other similar reservoirs containing high-level deposits.

Using the through diffusion method at room temperature, the migration of radionuclides (3H, 99Tc, 137Cs, 233U) in compacted samples of clay materials was studied through pore diffusion from a model leachates of phosphate matrix of radioactive waste with a total salt content of about 400 mg/L. The effective diffusion coefficients and sorption distribution coefficients of radionuclides for the studied clay barrier materials were determined based on the experimental results. The patterns of diffusion transfer of tritium, cesium and uranium were revealed depending on the structure, mineral composition of clay materials and the content of radionuclides in the pore solution. Numerical solutions are proposed for calculating the effective pore diffusion coefficients of these radionuclides. There are not yet enough data to analyze the factors influencing the diffusion of technetium.

This article focuses on modeling 90Sr migration in strong nitrate solutions in aquifers used for radioactive waste disposal. This type of radioactive waste disposal is typical only for the Russian Federation and is a unique object of study. The calculations are based on the laboratory study of strontium sorption in nitrate solutions on sandy, loamy and clayey rocks under biotic (with natural microbial communities obtained from Seversky repository) and abiotic conditions. To obtain a strontium sorption model, first, an ion exchange model in PHREEQC software is fitted to the experimental data both manually and automatically (using MOUSE software). Since real nitrate-ion concentrations at radioactive waste injection sites can reach values of hundreds of grams per liter, strontium Kd values are predicted for high ionic strength (for which no experimental study of strontium sorption efficiency has been carried out) with PHREEQC-model. The strontium transport models accounting for sorption and the nitrate reduction processes have been developed using two numerical software packages: the GeRa 3D hydrogeological simulation code and the PHREEQC reactive transport code. Reactive transport modeling under different conditions shows a high sensitivity to dispersion. A significant effect of sorption of nitrate ion on Sr sorption is shown and a relatively small contribution of microbial processes to strontium transport is noted for liquid radioactive waste injection sites.

The paper focuses on the 14 C, 36 C1, and 60 Co radionuclide distribution in graphite stacks and blocks, which play a crucial role in the decommission of uranium-graphite reactors. The shutdown uranium-graphite reactor ADE-5 provides an access to the graphite stack for the removal of graphite blocks. New interpretations are suggested for the radionuclide distribution in bulk graphite stacks based on testing results of graphite blocks removed from the uranium-graphite reactor ADE-5 in 2018.

The paper presents the results of studies of the spatial distribution of radionuclides14C,36Cl and60Co, which are key for decommissioning uranium-graphite reactors, in graphite stacks and individual graphite blocks. The studies were carried out on a shutdown PUGR ADE-5 of «PDC UGR» JSC, which provides access to the graphite stack with the ability to extract individual graphite blocks. Based on the results of a detailed experimental study of graphite blocks retrieved in 2018, the new interpretations of some features of the spatial distribution of radionuclides in UGR graphite was proposed.

A new approach to the study and description of leaching processes was proposed based on the fundamental difference in water exchange regimes during technological tests of confinement matrices of radioactive wastes (RW) according to standard leaching techniques and in the interaction of conditioned solid RW with groundwater under disposal conditions. According to this approach, in the leaching of matrices in a closed system, the main parameters characterizing the result of the process are the concentrations (specific activities) of saturation of the leachate with matrix elements and radionuclides. The proposed approach is illustrated by the example of calculating the leaching parameters of a model phosphate glass with imitators of radioactive waste, carried out using an appropriate kinetic model based on the data of experiments conducted in a static mode (in the absence of water exchange). The found values of the saturation concentrations of leaching solutions with matrix elements can be used to perform experiments and calculations on the migration of radionuclides in the materials of engineering barriers and host rocks.

The oxidation state of 99Tc adsorbed on the surface of AG-3 (I) and KAU (II) activated carbon from an aqueous potassium pertechnetate (KTcO4) solution has been studied by X-ray photoelectron spectroscopy at binding energies Eb in the range from zero to 1250 eV. The measured Eb(Tc 3d5/2) in sample I is 260.0, 258.2, and 258.3 eV, indicating the presence of three states: Tc7+, Tc5+, and Tc3+. The Tc 3d electron spectrum of the surface of sample II shows only two spin doublets, corresponding to two oxidation states of technetium, Tc3+ and Tc5+, in the ratio 23 : 77.

The release of long-lived radionuclides 14C and 36Cl , which are key products for assessment of disposal variants (near-surface, buried, or deep) of irradiated graphite, was analyzed. Test material was irradiated graphite from IUGR and RBMK-1000 reactors. Unlike other types of radioactive waste (RAW), graphite has the developed pore structure and, consequently, its effective surface area in contact with the leaching medium is increased in comparison with non-porous materials. With due to regard for the graphite pore structure, the mechanism describing the leaching process is proposed and experimentally substantiated. The parameters numerically characterizing graphite RAW in terms of leaching resistance, independent of the geometric shape of individual fragments (samples) are also introduced. The proposed parameters and an algorithm for their determination are recommended by the authors for use in predictive calculations of the release of radionuclides from graphite under conditions of long-term storage/disposal.

The paper deals with measurements of carbon-14 in irradiated nuclear graphite. It suggests approaches to the problem, impart experience in production and calibration of new C-14, reference standards got from irradiated elements (sleeve, block) of the uranium–graphite reactor (UGR) stacks. Reference standards in the form of irradiated graphite were created and certified by the C-14 concentration. The concentration of C-14 in different samples was determined by interlaboratory comparisons. Results of the work proved an opportunity of creating reference materials for the C-14 radionuclide based on irradiated graphite obtained from the graphite elements of the uranium–graphite reactors. C-14 reference standards with different concentrations of interfering Cl-36, Co-60, H-3, Cs-137, Sr-90 and other radionuclides can be used for future development of new methods of determination the C-14 concentration in irradiated graphite and spectrometric equipment calibration.

Work has been performed as part of the implementation of measures to reduce the hazard of irradiated graphite during the decommissioning of uranium-graphite nuclear reactors. Electrochemical decontamination is studied as a method of purifying the indicated wastes from radioactive contamination. The results of experiments on removing 60Co,134,137Cs, 154Eu, and 241Am from the surface of irradiated graphite under different conditions are presented. It is shown that electrochemical decontamination of irradiated nuclear graphite in an acidic medium makes it possible to reduce the activity of 60Co by a factor of 2–10 and Cs by a factor of 7–100 without gaseous products of reaction being formed.

Specific features of the porous structure of irradiated and virgin reactor graphite of Russian brands have been analyzed. The porosity parameters affecting the leaching of radionuclides from graphite radioactive waste (RAW) were determined: fraction of open porosity, distribution of pore throat cross-sections, surface of open pores. The volume of open pores accessible for leaching aqueous media in the irradiated graphite was estimated. A model describing the leaching of radionuclides from irradiated graphite accounting for the porosity contribution is proposed. The parameters that numerically characterize the resistance of graphite RAW against leaching and are independent of the geometric shapes of separate fragments (samples) were determined.

Neptunium poses challenges for radioactive wastes management due to its long half-life and ingrowth from americium decay, as well as its high mobility in water in its oxidized forms. Np(V)-bearing aqueous solutions were contacted with polished and crushed gneiss and dolerite from the Yeniseisky site, Russia, which is under consideration as a location for an underground repository, both in oxic and anoxic conditions at 20 and 90 °C. Neptunium sorption on crushed gneiss and dolerite in oxic conditions at 20 °C was quite low, but increased substantially in anoxic conditions. Autoradiography on samples treated at 20 °C revealed that sorbed Np was associated mostly with dark colored iron-bearing minerals, and solvent extraction indicated partial reduction of Np(V) to Np(IV). Sorption of Np from aqueous solutions contacted with dolerite and gneiss at 90 °C was in all cases higher than at the lower temperature. Autoradiography of polished rock samples treated in anoxic conditions at 90 °C revealed that Fe-bearing minerals had 3–4 orders of magnitude more Np sorbed on them than the quartz and feldspar grains. X-ray photoelectron spectroscopy indicated this sorbed Np was present as NpO2. This study supports the concept that the geologic conditions present at the planned repository site will retard Np migration.

Safe disposal of nuclear waste in a geologic repository will rely on natural geologic features and engineered barriers to greatly retard the movement of radionuclides from the repository. Clay minerals including bentonite are effective in retarding the migration of many radionuclides, but are ineffective for anionic radionuclides, of which pertechnetate is of particular concern owing to its relatively long half-life and the lack of natural isotopes that dilute it. Activated carbon is proposed as an additive material for reducing pertechnetate mobility in the nearfield. Activated carbon materials of different origins quantitatively sorb pertechnetate from aqueous solution under oxidizing conditions during the first day of contact, and sequential extraction showed that 73 % of this technetium is in the strongly bound fraction. X-ray photoelectron spectra (XPS) and extended X-ray absorption fine structure (EXAFS) spectra both demonstrated that no reduction of technetium occurred in the studied systems. The interaction of technetium with a composite material consisting of bentonite and activated carbon was studied at the first time. Effective technetium sorption was shown, with distribution coefficients (Kd) up to 740 cm3. g-1.

The effect of microorganisms on immobilization of Cs, Sr, U, and Tc on sandy rocks from upper aquifers was studied. Aquifer water samples taken from observation boreholes in the region of the suspended liquid radioactive waste (LRW) repository, B-2 basin of the Siberian Chemical Combine, were studied in a laboratory, and computer geochemical simulation was performed to estimate the diversity of radionuclide species taking into account the physicochemical conditions in upper aquifers. Analysis showed that the activity of microorganisms could affect the reducing conditions in this zone, favoring immobilization of radionuclides with variable oxidation state. Geochemical simulation showed that microbial processes can lead to the formation of U(IV) and Tc(IV) precipitates in a mixture with biogenic sulfide. Fouling of the surface of aquifer rock particles with biofilms reduced the sorption of Cs and Sr but improved the efficiency of the uranium immobilization.