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The United States and Russia both possess large quantities of nuclear waste, generated during the production of nuclear energy and nuclear weapons. To ensure that this radioactive material remains safely sequestered for tens of thousands of years or more, both countries plan to bury it in deep geologic repositories. However, US and Russian repository design strategies are highly distinct. For example, Russia plans to dilute waste in aluminophosphate glass, package waste in stainless steel containers, and bury waste in hard, crystalline granite gneiss rock. The US approach includes the use of borosilicate glass, multi-component superalloy containers, and porous volcanic tuff or highly-plastic bedded salt. The relative efficacies of these design choices remain uncertain. This represents a unique opportunity for applied, comparative study of various natural and engineered barriers to the release of radioactive materials. US-Russian collaboration and sharing of data on repository performance could provide a better technical basis for the long-term immobilization of nuclear waste. [ABSTRACT FROM AUTHOR]

Abstract The Middle-Jurassic Opalinus Clay is the foreseen host rock for radioactive waste disposal in central northern Switzerland. An extensive drilling campaign aiming to characterize the argillaceous formation resulted in a comprehensive drill core data set. The rheologically weak Opalinus Clay is only mildly deformed compared to the over- and underlying rock units but shows a variety of natural fractures. While these structures are hydraulically indistinguishable from macroscopically non-deformed Opalinus Clay today, their analysis allows for a better understanding of the deformation behaviour in the geological past. Here, we present an overview of the different fracture and fault types recorded in the Opalinus Clay and a detailed microstructural characterization of veins—natural dilational fractures healed by secondary calcite and celestite mineralizations. Macroscopic drill core analysis revealed five different natural fracture types that encompass tension gashes of various orientations with respect to bedding and small-scale faults with displacements typically not exceeding the drill core diameter. The occurrence of different fault types generally fits well with the local tectonic setting of the different drilling sites and with respect to the neighbouring regional fault zones. The microstructural investigations of the various vein types revealed their often polyphase character. Fibrous bedding-parallel veins of presumable early age were found to be overprinted by secondary slickenfibres. The polyphase nature of fibrous bedding parallel veins and slickenfibres is supported by differing elemental compositions, pointing towards repeated fracturing and mineralization events. Direct dating of vein calcites with U–Pb was unsuccessful. Nevertheless, age constraints can be inferred from structural orientations and fault slip kinematics. Accordingly, some of the veins already formed during sediment compaction in Mesozoic times, others possibly relate to Early Cenozoic foreland uplift. The youngest veins are most likely related to Late Cenozoic regional tectonic events, such as the Jura fold-and-thrust belt to the south and the Hegau-Lake Constance Graben to the northeast of the study area. During these latest tectonic events, previously formed veins acted as rheologically stiff discontinuities in the otherwise comparably weak Opalinus Clay along which deformation of the rock formation was re-localized.

Integrated underground spatial maps have been under construction in South Korea since the 2015 demonstration project as part of ground subsidence preventive measures aimed at augmenting underground safety management. However, the Integrated Underground spatial map constructed for each region at the time remained unchanged until now because of the absence of a renewal system. A clear statement on the obligatory submission of construction completion reports for underground information subject to changes and updates is found in Article 42, paragraph 2 of the Special Act on Underground Safety Management. However, the collection of reports is impeded by difficult access to the submission system and a lack of submission standards. Therefore, the Ministry of Land, Infrastructure, and Transport required by law the direct submission of reports through an online submission system for construction completion reports who underground space developers. In this study, a standard database for construction completion report submission was designed to create an automatic renewal system based on the submitted reports. By presenting the framework for three-dimensional underground facility model processing and renewal automation, a method of rapidly building a three- dimensional underground facility model at minimum cost is presented. [ABSTRACT FROM AUTHOR]

In this study, the chemical durability of Gd 2 Zr 2 O 7 transparent ceramics was investigated under different pH environments. The influence of acidic and alkaline environments on the leaching behavior of Gd 2 Zr 2 O 7 immobilization matrix were studied to simulate the disposal of nuclear waste forms in deep geological landfills. The measured results demonstrate that the normalized leaching rates in acidic environments are higher than those in alkaline and neutral environments. The macroscopic color of all samples fades after leaching, and the visible range in-line transmittance of the leached transparent ceramics decreases slightly at pH ≤ 9. The leaching behavior leads decrement of lattice volume and increment of surface roughness, but it exhibits no significant effect on the surface elemental distribution. Never the less, Gd 2 Zr 2 O 7 transparent ceramic exhibits excellent chemical durability in weak alkaline environment, and the leaching process leads to the discoloration of Gd 2 Zr 2 O 7 transparency. [ABSTRACT FROM AUTHOR]

The Gd 2 Zr 2 O 7 (GZO)-based ceramic waste form exhibits promising potential for radioactive waste immobilization. However, the isomorphic immobilization capacity and irradiation effects of GZO-based ceramic waste forms with varying waste loads remain inadequately understood. This study employed a facile combustion method to synthesize ceramic powders and utilized conventional sintering with a relatively short dwelling time to fabricate GZO-based ceramic waste forms. We investigated the isomorphic immobilization behavior and He ion irradiation effects of the synthesized waste forms. The outcomes demonstrate that GZO effectively immobilizes simulated waste composed of 11 types of oxides, achieving a waste load as high as 50 wt% with a single pyrochlore phase and homogeneous element distribution. Under He+ irradiation, aggravated structural degradations, including exacerbated lattice swelling, increased amorphization degree, and premature He bubble chain/microcrack formation, were observed in GZO-based waste forms with an increasing waste load. [ABSTRACT FROM AUTHOR]

The article presents the results of a radiation survey carried out on the liquid radioactive waste disposal system of decommissioned MR and RFT research reactors. The system is located in the buildings adjacent to the dismantled reactor building on the territory of the Kurchatov Institute national research center. The results of the visual and radiation survey carried out on the equipment and premises of the disposal system for liquid radioactive waste will inform the development of design documentation for its elimination. Following the removal of liquid radioactive waste from storage tanks, the equipment and building structures at this location should be dismantled. The carrying out of recommended works will complete the decommissioning of MR and RFT research reactors. Further, the site of the reactor buildings and their auxiliary systems will be rehabilitated to achieve a radiation situation that meets the sanitary and hygienic standards for the group B personnel. [ABSTRACT FROM AUTHOR]

Salt offers an optimal medium for the permanent isolation of heat‐producing radioactive waste due to its impermeability, high thermal conductivity, and ability to close fractures through creep. A thorough understanding of the thermal‐hydrological‐mechanical (THM) processes, encompassing brine migration, is fundamental for secure radioactive waste disposal within salt formations. At the Waste Isolation Pilot Plant (WIPP), we conducted joint in situ geophysical monitoring experiments during active heating to investigate brine migration near excavations. This experiment incorporated electrical resistivity tomography (ERT) alongside high‐resolution fiber‐optic‐based distributed temperature sensing within a controlled heating experiment. Additionally, discrete element model (DEM) based numerical simulations were conducted to simulate THM processes during heating, providing a more mechanistic understanding of the coupled processes leading to the observed changes in geophysical measurements. During heating, resistivity shifts near the heater were reasonably explained by temperature effects. However, in more distant, cooler regions, the resistivity decrease exceeded predictions based solely on temperature. DEM simulations highlighted brine migration, propelled by pore pressure gradients, as the likely primary factor contributing to the additional resistivity decline beyond temperature effects. The comparison between the predicted ERT responses and observations was much improved when considering the effects of brine migration based on the DEM simulations. These geophysical and simulation findings shed light on brine migration in response to salt heating, enhancing our understanding of the coupled THM processes in salt for safe radioactive waste disposal. Plain Language Summary: Brine migration within a salt nuclear waste repository induces corrosion and the dissolution of radionuclides, influencing their safety. Understanding the impact of waste‐generated heat on fluid dynamics and salt rock deformation is pivotal for the secure disposal of radioactive waste in salt formations. A controlled heating experiment at the Waste Isolation Pilot Plant, New Mexico, was monitored for geophysical changes, including formation resistivity and temperature. Numerical simulations modeled rock behaviors to elucidate the coupled Thermo‐Hydro‐Mechanical processes and their impacts on observed signals. While resistivity changes near the heater were adequately explained by temperature effects alone, more distant and cooler regions exhibited disproportionately larger resistivity decreases. Numerical simulations identified brine migration as the primary cause for this significant decrease, surpassing temperature effects. Incorporating brine migration effects in simulations substantially improved the prediction of resistivity changes. These findings highlight the role of joint geophysical monitoring and numerical simulations for mechanistic understanding of the coupled processes during heating, which is important for safe storage of nuclear waste in salt formations. Key Points: We conducted in situ geophysical monitoring during a controlled heating experiment in salt formationsWhile temperature effects explain resistivity changes near the heater well, more resistivity decreases were observed at the further regionsNumerical simulations indicate pore pressure driven brine migration is likely the primary reason for the resistivity changes beyond temperature effects [ABSTRACT FROM AUTHOR]

This paper reports a method for the fabrication of mineral-like SrMoO4 ceramics with a powellite structure, which is promising for the immobilization of the high-energy 90Sr radioisotope. The reported method is based on the solid-phase "in situ" interaction between SrO and MoO3 oxides initiated under spark plasma sintering (SPS) conditions. Dilatometry, XRD, SEM, and EDX methods were used to investigate the consolidation dynamics, phase formation, and structural changes in the reactive powder blend and sintered ceramics. The temperature conditions for SrMoO4 formation under SPS were determined, yielding ceramics with a relative density of 84.0–96.3%, Vickers microhardness of 157–295 HV, and compressive strength of 54–331 MPa. Ceramic samples demonstrate a low Sr leaching rate of 10−6 g/cm2·day, indicating a rather high hydrolytic stability and meeting the requirements of GOST R 50926-96 imposed on solid radioactive wastes. The results presented here show a wide range of prospects for the application of ceramic matrixes with the mineral-like composition studied here to radioactive waste processing and radioisotope manufacturing. [ABSTRACT FROM AUTHOR]

To study the thermal healing effect on stress-induced damage of Beishan granite, the host rock of China's high-level radioactive waste (HLW) geological disposal underground research laboratory, comprehensive experimental works were carried out. The deformation rate analysis test was carried out on granite after different stress levels and different temperature conditions (room temperature and mild temperature, 150 °C). Results showed that the deformation memory to the previous stress loading (stress damage) was fading after mild temperature treatment, and the Kaiser effect was also faded notably that AE activity was suppressed after mild temperature treatment, which implied stress-induced damage was recovered under mild temperature. In addition, stress-induced pores, especially large pores, were decreased markedly after mild temperature treatment characterized by the nuclear magnetic resonance measurement. Overall, the mild temperature possesses the ability to heal stress (σcc to σci, σci to σcd) induced damage. The permeability reduction, strength enhancement, pores decreasing, and the mechanism under mild temperature were jointed analyzed, which is positive for the long-term stability and tightness of the HLW geological repository. Highlights: The healing effect of temperature on stress-damaged Beishan granite was studied. The DME of granite after different stress loaded was fading after 150 °C treated. The AE activity was suppressed and the Kaiser effect disappeared after 150 °C treated. The mild temperature exhibits the ability of healing stress-induced damage. [ABSTRACT FROM AUTHOR]

A compacted bentonite buffer is one of the most important components in engineered barrier systems for the disposal of high-level radioactive waste produced by nuclear power generation. The compacted bentonite buffer contributes significantly to the protection of the disposal canister against the external impact and penetration of groundwater. Therefore, it must satisfy thermo–hydraulic–mechanical requirements. Although thermal–hydraulic properties have been investigated extensively, the mechanical properties of compacted bentonite buffers have not been comprehensively evaluated. Therefore, a series of uniaxial compression and point load tests were conducted in this study. The mechanical properties of Korean compacted bentonite specimens with varying water contents and dry densities, such as uniaxial compressive strength, elastic modulus, and point load index, were obtained. As the dry density increased, the uniaxial compressive strength, elastic modulus, and point load strength index increased, whereas Poisson's ratio decreased slightly. This indicates that the uniaxial compressive strength, elastic modulus, and point load strength index are proportional to the dry density. However, the mechanical properties and water content do not exhibit a significant correlation. [ABSTRACT FROM AUTHOR]

Site-specific hydrogeological parameters are a prerequisite to investigate the migration of radionuclides in groundwater due to leaching from low-level solid radioactive waste disposal facility (SRWDF) site. The leaching event is a low probability event due to all the engineered safety features provided in SRWDF, but the impact assessment, even due to this low probability event of leaching is a regulatory requirement. The present study focuses on SRWDF located in Tarapur, India. Hydrogeological parameters such as radionuclide specific distribution coefficient (Kd), hydraulic conductivity (HC) and porosity are estimated for a watershed area in and around the SRWDF. Values of HC and porosity for soil samples collected from the disposal site are found to be in the range 0.006–0.15 cm/s (average 0.059 cm/s) and 25.64–32.62% (average 29.19%), respectively. Average hydraulic gradient is found to be 0.007. Using the estimated values of hydraulic gradient, porosity and HC in Darcy's law, groundwater velocity is calculated as 0.13–3.22 m/day. The distribution coefficient of caesium, the most prevalent radionuclide in low-level solid radioactive waste, is estimated to be ranging between 116.73 and 393.54 ml/g. Physical and chemical characteristics of soil like pH, electrical conductivity, organic matter content, moisture content, CaCO3 content and effective cation exchange capacity (ECEC) are determined experimentally. Correlation among the estimated parameters is also studied. The Kd of caesium is found to be positively correlated with porosity, ECEC, clay and silt fraction of the soil and negatively correlated to sand fraction. The site-specific parameters estimated in this study would be useful for carrying out the radiological impact assessment (RIA) of the SRWDF at Tarapur through the terrestrial pathway. Research Highlights: The values of site-specific distribution coefficient (Kd) of the soil samples range from 116.73 to 393.54 ml/g. Hydrogeological parameters like hydraulic conductivity, porosity, groundwater velocity have been estimated, and the values range from 0.006 to 0.15 cm/s, 25.64 to 32.62% (average 29.19%), 0.13 to 3.22 m/day, respectively. Physico-chemical properties like pH, electrical conductivity, organic matter content, moisture content, CaCO3 content, and effective cation exchange capacity have been estimated. Correlation among estimated parameters is studied. Site-specific Kd is found to be positively correlated with porosity, effective cation exchange capacity, clay and silt fraction of the soil and negatively correlated to sand fraction. [ABSTRACT FROM AUTHOR]

[Display omitted] Understanding the behaviour of uranium waste, for disposal purposes, is crucial due to the correlation between pH values and the disposal of distinct types of waste, with low level waste typically associated with acidic pH values, and higher and intermediate level waste commonly related to alkaline pH values. We studied the adsorption of U(VI) on sandstone and volcanic rock surfaces at pH 5.5 and 11.5 in aqueous solutions with and without bicarbonate (2 mM HCO 3 –) using XAS and FTIR. In the sandstone system, U(VI) adsorbs as a bidentate complex to Si at pH 5.5 without bicarbonate and as uranyl carbonate species with bicarbonate. At pH 11.5 without bicarbonate, U(VI) adsorbs as monodentate complexes to Si and precipitates as uranophane. With bicarbonate at pH 11.5, U(VI) precipitated as a Na-clarkeite mineral or remained as a uranyl carbonate surface species. In the volcanic rock system, U(VI) adsorbed to Si as an outer sphere complex at pH 5.5, regardless of the presence of bicarbonate. At pH 11.5 without bicarbonate, U(VI) adsorbed as a monodentate complex to one Si atom and precipitated as a Na-clarkeite mineral. With bicarbonate at pH 11.5, U(VI) sorbed as a bidentate carbonate complex to one Si atom. These results provide insight into the behaviour of U(VI) in heterogeneous, real-world systems related to the disposal of radioactive waste. [ABSTRACT FROM AUTHOR]

Copyright of World Nuclear Geoscience is the property of World Nuclear Geoscience Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of Perspektiven der Wirtschaftspolitik is the property of De Gruyter and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Fluid flow problems in bi‐porous media have strong implications in a variety of industrial applications, such as nuclear waste disposal and composites manufacturing. Despite various numerical solutions proposed in the literature, the prediction of dual‐scale flow remains a challenging task due to the requirement of fine meshes to resolve the complex microstructures of bi‐porous media. This paper introduces a new numerical solver based on fast Fourier transform (FFT) for the Stokes–Brinkman problem to predict fluid flow in bi‐porous media with local anisotropy. The novel FFT solver is derived from a velocity‐form of the problem, in contrast to the stress‐form proposed by a recent work. The Anderson acceleration technique is applied for the first time to the Stokes–Brinkman problem, leading to a substantial (orders of magnitude) improvement of the convergence rate. A range of detailed numerical examples are provided to validate the method against the analytical and literature results. Parametric studies are also demonstrated to aid in the selection of model parameters to achieve optimum numerical performance. With a 3D unit cell of a woven textile fabric, we demonstrate that the proposed method is capable of handling high‐resolution simulations with strongly heterogeneous microstructures. Combined with a parallelized implementation over high‐performance computing systems, our method demonstrates a new state‐of‐the‐art in numerical solvers for the Stokes‐Brinkman problem, in terms of computation capacity. [ABSTRACT FROM AUTHOR]

The design of an underground nuclear waste disposal requires a full characterization of concrete under various thermo-hydro-mechanical-chemical conditions. This experimental work studied the characterization of coupled thermo-hydro-mechanical effects using concretes made with cement CEM I or CEM V/A (according to European norms). Uniaxial and triaxial compression under 5 MPa confining pressure tests were performed under three different temperatures (T = 20°C, 50°C, and 80°C). The two concretes were dried under relative humidity (RH) to obtain a partially saturated state of approximately 70%. The results showed that the effects of water saturation and confining pressure are more important than that of temperature. Drying in high RH at different target temperatures led to an increase in uniaxial and triaxial strength but drying at 105°C had a negative effect on the strength and Young's modulus. Oven drying caused microcracking in the concrete. The microcracks deeply influenced the thermal damage to the material, affecting its mechanical behavior. The triaxial strength clearly increased due to the presence of confining pressure. Moreover, an important influence of cement type was observed on the mechanical properties; the concrete based on CEM V/A had a greater porosity than CEM I, and a finer pore structure appeared due to the presence of mineral admixtures. [ABSTRACT FROM AUTHOR]

Many residents of the Russian Southern Urals were exposed to radioactive environmental pollution created by the operations of the Mayak Production Association in the mid- 20th century. There were two major releases: the discharge of about 1x1017 Bq of liquid waste into the Techa River between 1949 and 1959; and the atmospheric release of 7.4 * 1016 Bq as a result an explosion in the radioactive waste-storage facility in 1957. The releases into the Techa River resulted in the exposure of more than 30,000 people who lived in riverside villages between 1950 and 1961. The 1957 accident contaminated a larger area with the highest exposure levels in an area that is called the East Urals Radioactive Trace (EURT). Current epidemiologic studies of the exposed populations are based on dose estimates obtained using a Monte-Carlo dosimetry system (TRDS-2016MC) that provides multiple realizations of the annual doses for each cohort member. These dose realizations provide a central estimate of the individual dose and information on the uncertainty of these dose estimates. In addition, the correlation of individual annual doses over realizations provides important information on shared uncertainties that can be used to assess the impact of shared dose uncertainties on risk estimate uncertainty.This paper considers dose uncertainties in the TRDS-2016MC. Individual doses from external and internal radiation sources were reconstructed for 48,036 people based on environmental contamination patterns, residential histories, individual 90Sr body-burden measurements and dietary intakes. Dietary intake of 90Sr resulted in doses accumulated in active bone marrow (or simply, marrow) that were an order of magnitude greater than those in soft tissues. About 84% of the marrow dose and 50% of the stomach dose was associated with internal exposures. The lognormal distribution is well-fitted to the individual dose realizations, which, therefore, could be expressed and easily operated in terms of geometric mean (GM) and geometric standard deviation (GSD). Cohort average GM for marrow and stomach cumulative doses are 0.21 and 0.03 Gy, respectively. Cohort average dose uncertainties in terms of GSD are as follows: for marrow it is 2.93 (90%CI: 2.02–4.34); for stomach and the other non-calcified tissues it is 2.32 (90% CI: 1.78–2.9). [ABSTRACT FROM AUTHOR]

While one should be aware that its zero CO2 emission is actually achievable only when electric power is generated, nuclear power is one of the most viable and proven "carbon-free" energy sources to provide baseload electricity to the current energy-demanding society. Even after the power generation, the major part of spent nuclear fuels still consists of recyclable nuclear fuel materials such as U and Pu, promising circular economy of nuclear energy systems in principle. However, actual situations are not very simple due to the following issues: (1) resource security of nuclear fuel materials, (2) issues of depleted uranium, and (3) treatment and disposal of high-level radioactive wastes. In this Perspective, I discussed how chemistry can contribute to resolving these problems and what task academic research in fundamental chemistry should take on there. [ABSTRACT FROM AUTHOR]

Copyright of World Nuclear Geoscience is the property of World Nuclear Geoscience Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of World Nuclear Geoscience is the property of World Nuclear Geoscience Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of World Nuclear Geoscience is the property of World Nuclear Geoscience Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of World Nuclear Geoscience is the property of World Nuclear Geoscience Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of World Nuclear Geoscience is the property of World Nuclear Geoscience Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of World Nuclear Geoscience is the property of World Nuclear Geoscience Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of World Nuclear Geoscience is the property of World Nuclear Geoscience Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of World Nuclear Geoscience is the property of World Nuclear Geoscience Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of World Nuclear Geoscience is the property of World Nuclear Geoscience Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of World Nuclear Geoscience is the property of World Nuclear Geoscience Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of World Nuclear Geoscience is the property of World Nuclear Geoscience Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of World Nuclear Geoscience is the property of World Nuclear Geoscience Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of World Nuclear Geoscience is the property of World Nuclear Geoscience Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Copyright of World Nuclear Geoscience is the property of World Nuclear Geoscience Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

A model is proposed for the mixing chamber of a centrifugal extractor, which is used in existing technological plants for the processing of waste nuclear fuel and designed for the recovery of target components from a solution with the use of a solvent (extragent). The model is developed by the freely distributed OpenFoam software package, which provides an option of modifying the program code via the addition of new dependences to the equations to be solved. [ABSTRACT FROM AUTHOR]

The optimal conditions for the preconcentration of 90Sr from natural water samples with carbonate-containing zirconium hydroxide of T-3K grade were determined. A method was developed for the determination of 90Sr in natural waters with strontium concentration with T-3K sorbent of a low detection limit: 0.03 Bq/L from 1 L water samples, 0.02 Bq/L from 2 L sample. The method was tested during monitoring of 90Sr in natural reservoirs in the Sverdlovsk oblast and Chelyabinsk oblast and in water samples from monitoring and observation wells in a long-term RW storage facility locations. It was shown that the developed method can be used without constant monitoring of the yield of strontium in the concentrate for natural waters with a calcium concentration of ≤76 mg/L and a total hardness of ≤5.2 mM, but in view of the established elution yield of 91 ± 1 and 81 ± 2% from 1 and 2 L water samples, respectively, which simplifies and speeds up the analysis. With higher hardness of natural waters, the strontium elution yield as a function of calcium concentration or water hardness reported in the article can be used to determine the yield of strontium in the concentrate. [ABSTRACT FROM AUTHOR]

For the safe disposal of high-level radioactive waste, it is necessary to establish a numerical model that can simulate the phenomenon of buffer expansion due to groundwater inflow. The Barcelona basic model (BBM), one of many models that describe the swelling behavior of buffer, can represent the behavior of expansive soil but requires various hydro-mechanical input parameters. Conventional experiments to determine these parameters are time-consuming and complicated; therefore, this study proposes a method of determining the BBM parameters by comparing the results of swelling tests and numerical analysis. The relationships between these parameters were determined through an artificial neural network and multi-objective optimization was used to derive the Pareto optimal sets. Among various optimized solutions, single BBM parameters were derived from Pareto sets by considering additional conditions. The results of the numerical analysis using the identified parameters and the experimental results exhibited similar trends. [ABSTRACT FROM AUTHOR]

This paper presents a numerical analysis of the mechanical performance of a bentonite clay buffer for the containment of nuclear waste in the context of deep geological disposal. The design of the buffer is based on the Swiss concept where the waste canisters are emplaced on pedestals of compacted bentonite blocks and the remaining space between the tunnel and the canister is backfilled with grains of highly compacted bentonite. A complete analysis of the long‐term performance of the repository requires a good understanding of the mechanical evolution of the bentonite upon heating from the radioactive waste and hydration from the host rock. Despite its importance, the implications of the initial heterogeneous bentonite layout, characterised by blocks and grains, on the final dry density at the repository scale in the steady state have not been previously studied. The present study aims to shed light into these processes by means of finite element modelling using an advanced constitutive model for the bentonite behaviour that considers several thermo‐hydro‐mechanical couplings. The constitutive model is shown to be able to reproduce several laboratory tests involving saturation of block and pellets at different dry densities. The model predictions, extended up to 100,000 years, indicate that the bentonite blocks and grains tend to homogenise in terms of dry density as the buffer reaches full saturation. Due to the different swelling potential of the block pedestal and the granular backfill, the canister is subjected to movements, although these remain relatively small. The impact of initial segregation of the granular bentonite is also studied and it is seen to not to affect substantially the mechanical evolution of the buffer, although it might reduce canister displacements. [ABSTRACT FROM AUTHOR]