Safe Disposal Of Radioactive Waste Research Articles

NON-CARCINOGENIC VCI SOLUTION FOR LONG TERM METAL PROTECTION

Corrosion is a dangerous and extremely costly problem. Because of it, buildings and bridges can collapse, oil pipelines break, chemical plants leak, and bathrooms flood. Corroded electrical contacts can cause fires and other problems, corroded medical implants may lead to blood poisoning, and air pollution has caused corrosion damage to work around the world. Corrosion threatens the safe disposal of radioactive waste that must be stored in containers for tens of thousands of years. When reduction and oxidation take place on different kinds of metal in contact with one another, the process is called galvanic corrosion. In electrolytic corrosion, which occurs most commonly in electronic equipment, water or other moisture becomes trapped between two electrical contacts that have an electrical voltage applied between them. The result is an unintended electrolytic cell. VCI is the generic term for “volatile corrosion inhibitor” a revolutionary technology that simplifies corrosion protection and is ideal for keeping enclosed void spaces (e.g., packages, equipment internals, or structural metal cavities) rust-free. Keywords: Corrosion; VCI; Rust.

Harnessing Geothermal Energy Potential from High-Level Nuclear Waste Repositories

The disposal of high-level nuclear waste (HLW) has been one of the most challenging issues for nuclear energy utilization. In this study, we have explored the potential of extracting decay heat from HLW, taking advantage of recent advances in the technologies to utilize low-temperature geothermal resources for the co-generation of electricity and heat. Given that geothermal energy entails extracting heat from natural radioactivity within the Earth, we may consider that our approach is to augment it with an anthropogenic geothermal source. Our study—for the first time—introduces a conceptual model of a binary-cycle geothermal system powered by the heat produced by HLW. TOUGHREACT V3.32 software was used to model the heat transfer resulting from radioactive decay to the surrounding geological media. Our results demonstrate the feasibility of employing the organic Rankine cycle (ORC) to generate approximately 108 kWe per HLW canister 30 years after emplacement and a heat pump system to produce 81 kWth of high-potential heat per canister for HVAC purposes within the same timeframe. The proposed facility has the potential to produce carbon-free power while ensuring the safe disposal of radioactive waste and removing the bottleneck in the sustainable use of nuclear energy.

Plutonium mobility and reactivity in a heterogeneous clay rock barrier accented by synchrotron-based microscopic chemical imaging

The long-term safe disposal of radioactive waste corresponds to a challenging responsibility of present societies. Within deep geological waste disposal concepts, host rocks correspond to the ultimate safety barrier towards the environment. To assess the performance of such barriers over extended time scales, mechanistic information on the interaction between the radiotoxic, long-lived radionuclides like plutonium and the host rock is essential. Chemical imaging based on synchrotron microspectroscopic techniques was used to visualize undisturbed reactive transport patterns of Pu within pristine Opalinus Clay rock material. Pu+V is shown to be progressively reduced along its diffusion path to Pu+IV and Pu+III due to interaction with redox-active clay rock constituents. Experimental results and modeling emphasize the dominant role of electron-transfer reactions determining the mobility of Pu in reactive barrier systems. The effective migration velocity of Pu is controlled by the kinetic rates of the reduction to Pu+IV and Pu+III and the redox capacity of the involved electron donor pools. To advance our predictive capabilities further, an improved understanding of the nature and capacity of redox-active components of the reactive barrier material is fundamental. The findings represent an essential contribution to the evaluation of the long-term safety of potential nuclear waste repositories and have implications regarding the development of effective geological disposal strategies.

Influence of radiation on borosilicate glass leaching behaviors

Vitrification is widely recognized as a promising method for the geological disposal of high-level radioactive waste (HLW) worldwide. To ensure the safe disposal of radioactive waste, the borosilicate glass that vitrifies HLW must exhibit exceptional water resistance to prevent the possibility of groundwater corrosion and subsequent radioactive leaks. Radiation might change the water resistance of borosilicate glass. A series of zirconium-containing borosilicate glass with an irradiation dose of 0.3 dpa were utilized to examine the radiation effect on glass-water interaction. Scanning electron microscopy (SEM), Time-of-Flight Secondary ion mass spectrometry (ToF-SIMS), X-ray photoelectron spectroscopy (XPS), Inductively Coupled Plasma Optical Emission spectroscopy (ICP-OES) and Fourier transform infrared spectroscopy (FTIR) were used to investigate the leaching behavior of the non- and irradiated samples. The depth profile of the leached samples implied the interdiffusion dominated glass-water interaction. The results from FTIR and ICP-OES indicated that, after irradiation, the initial leaching rate increased by threefold. Additionally, the impact of different zirconium contents on the water resistance of borosilicate glass was also presented.

Unlocking the key role of bentonite fungal isolates in tellurium and selenium bioremediation and biorecovery: Implications in the safety of radioactive waste disposal

Research on eco-friendly bioremediation strategies for mitigating the environmental impact of toxic metals has gained attention in the last years. Among all promising solutions, bentonite clays, to be used as artificial barriers to isolate radioactive wastes within the deep geological repository (DGR) concept, have emerged as effective reservoir of microorganisms with remarkable bioremediation potential. The present study aims to investigate the impact of bentonite fungi in the speciation and mobility of selenium (Se) and tellurium (Te), as natural analogues 79Se and 132Te present in radioactive waste, to screen for those strains with bioremediation potential within the context of DGR. For this purpose, a multidisciplinary approach combining microbiology, biochemistry, and microscopy was performed. Notably, Aspergillus sp. 3A demonstrated a high tolerance to Te(IV) and Se(IV), as evidenced by minimal inhibitory concentrations of >16 and >32 mM, respectively, along with high tolerance indexes. The high metalloid tolerance of Aspergillus sp. 3A is mediated by its capability to reduce these mobile and toxic elements to their elemental less soluble forms [Te(0) and Se(0)], forming nanostructures of various morphologies. Advanced electron microscopy techniques revealed intracellular Te(0) manifesting as amorphous needle-like nanoparticles and extracellular Te(0) forming substantial microspheres and irregular accumulations, characterized by a trigonal crystalline phase. Similarly, Se(0) exhibited a diverse array of morphologies, including hexagonal, irregular, and needle-shaped structures, accompanied by a monoclinic crystalline phase. The formation of less mobile Te(0) and Se(0) nanostructures through novel and environmentally friendly processes by Aspergillus sp. 3A suggests it would be an excellent candidate for bioremediation in contaminated environments, such as the vicinity of deep geological repositories. It moreover holds immense potential for the recovery and synthesis of Te and Se nanostructures for use in numerous biotechnological and biomedical applications.

Robust nitrogen-rich covalent polymeric networks for ultra-efficient and selective palladium capture under harsh conditions

Selective capture of palladium (Pd) from high-level liquid waste (HLLW) is one of the frontier issues in materials science in order to alleviate the Pd resource scarcity and guarantee the safe disposal of radioactive wastes. However, this task remains a great challenge in HLLW treatment associated with low capacity, poor selectivity, and limited stability of materials in strongly acidic and radioactive environments. Here we conquer this challenge by designing covalent polymeric networks (CPNs) equipped with abundant bidentate (CPN-1) or tridentate (CPN-2) nitrogen-rich moieties as specific nano-traps for Pd(II). These CPNs are constructed through efficient and facile click reactions, among which CPN-1 bearing plentiful pyridinyl-triazolyl moieties sets a new record for Pd(II) uptake capacity (535 mg g−1), notably surpassing all the reported materials for Pd(II) separation from HNO3 solutions to date. Moreover, CPN-1 features additional attractive merits including good reusability, feasible dynamic column separation, outstanding Pd(II) selectivity over 17 coexisting cations, and excellent stability under highly acidic and radioactive conditions, making it highly promising for practical applications. This work provides a facile way to customize CPN adsorbents with superior performance and highlights the great potential of the prepared materials for metal capture applications.

Evaluation of the Solidification of Radioactive Wastes Using Blast Furnace Slag as a Solidifying Agent.

The decommissioning process of nuclear power facilities renders hundreds of thousands of tons of various types of waste. Of these different waste types, the amount of concrete waste (CW) varies greatly depending on the type of facility, operating history, and regulation standards. From the previous decommissioning projects, CW was estimated to comprise 60-80 wt.% of the total weight of radioactive wastes. This represents a significant technical challenge to any decommissioning project. Furthermore, the disposal costs for the generated concrete wastes are a substantial part of the total budget for any decommissioning project. Thus, the development of technologies effective for the reduction and recycling of CW has become an urgent agenda globally. Blast furnace slag (BFS) is an industrial byproduct containing a sufficient amount (higher than 30%) of CaO and it can be used as a substitute for ordinary Portland cement (OPC). However, there have been few studies on the application of BFS for the treatment of radioactive waste from decommissioning processes. This study was conducted to evaluate the performance of the solidification agent using ground granulated BFS (SABFS) to pack radioactive wastes, such as the coarse aggregates of CW (CACW), waste soil (WS), and metal waste (MW). The analytical results indicated that the CaO content of the ground granulated BFS was 36.8% and it was confirmed that calcium silicate hydrate (CSH) could be activated as the precursor of the hydration reactions. In addition, the optimum water-to-binder ratio was determined to be 0.25 and Ca(OH)2 and CaSO4 were found to be the most effective alkaline and sulfate activators for improving the compressive strength of the SABFS. The maximum packing capacities of the SABFS were determined to be 9 and 13 wt.% for WC and WM, respectively, when the content of CW was fixed at 50 wt.%. The results of the leaching tests using SABFS containing radioactive wastes contaminated with Co, Cs, and Sr indicated that their leachability indices met the acceptance level for disposal. Consequently, the SABFS can be used as a solidifying agent for the safe disposal of radioactive waste.

BARIK: an extended Hoek–Brown-based anisotropic constitutive model for fractured crystalline rock

Abstract. The disposal of heat-generating radioactive waste in deep geologic formations is a global concern. Numerical methods play a key role in understanding and assessing the disposal scenarios of radioactive waste in deep geological repositories. However, the complexities of the thermal, hydrological, mechanical, chemical, and biological processes associated with the disposal of radioactive waste in porous and fractured materials constitute significant challenges. One of the most challenging issues in this field is the complex material behavior of fractured crystalline rock. The presence of fractures makes the rock anisotropic, nonlinear, and dependent on loading paths. Additionally, the Biot coefficient cannot be considered constant throughout the critical and subcritical fracture development regions. These factors make the development of an accurate constitutive model for fractured crystalline hard rock a critical component of any deep geological disposal project. Furthermore, to demonstrate the integrity of the containment-providing rock zone in crystalline host rock, the qualitative integrity criteria must be quantified so that numerical simulation can be performed with concrete numerical values. Part of this assessment for a crystalline host rock is a dilatancy criterion, which is currently based on the Hoek–Brown constitutive model. BARIK is the German acronym for the research project on which this paper is based. This contribution provides an overview of the development and verification of the BARIK constitutive model, an extended Hoek–Brown model for fractured crystalline hard rock that takes into account up to three fracture systems. The model enables the consideration of the matrix and joint behavior of the rock separately, with each component having unique strength characteristics and failure criteria. These criteria are formulated such that suitable consideration of the strength-reducing properties of the respective fracture systems during barrier integrity verification is possible. The BARIK model has been implemented into two computer codes, FLAC3D and MFront for OpenGeoSys, allowing for the identification and evaluation of any inaccuracies that may arise from the use of different codes. The model enables isotropic–elastic, orthotropic–elastic, isotropic–elasto-plastic, and orthotropic–elasto-plastic calculations of the matrix, making it a valuable tool for the site selection process and for the construction and long-term safety of underground repositories. Furthermore, this poster presentation will show how the constitutive model was evaluated in relation to the dilatancy criterion and how the BARIK constitutive model's suitability for conducting an integrity assessment was validated. In conclusion, the development of BARIK is a significant step forward in the understanding and modeling of the complex material behavior of fractured crystalline hard rock. This contribution will provide insights into the development and verification of this model for the safe disposal of radioactive waste.

Joint Geophysical and Numerical Insights of the Coupled Thermal‐Hydro‐Mechanical Processes During Heating in Salt

AbstractSalt 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.

Impact of Fe(II) on 99Tc diffusion behavior in illite

A comprehensive understanding of the geochemical behavior of 99Tc is of great importance for safe disposal of radioactive waste and remediation of contaminated environmental sites. Illite is one of the most common constituents of clay rocks, and thus used in this work as a model system for studying the retention and transport of 99Tc in clay-rich systems. In this study, a through-diffusion technique was applied to investigate the diffusion behavior of Tc in compacted illite clay under oxic and anoxic conditions. Particular focus of this investigation was on the role of Fe(II) on the redox state and mobility of Tc in clay. As the diffusion cells contained stainless steel filters for confining the clay plug, 99Tc-diffusion in the filters was assessed first, followed by 99Tc diffusion in the illite column with or without Fe-loading. Two types of Fe(II) loadings in illite were considered, surface complexed Fe(II) on illite edge sites and Fe(II) added as pyrite grains. COMSOL Multiphysics was used to analyze the diffusion data. The measured filter porosity was about 0.2 and the effective diffusion coefficient, De, for Tc (as TcO4−) in the filter was 0.59 × 10−10 m2/s. Tc diffusion in illite under ambient conditions showed a typical anion diffusion behavior with De in the range 0.38-0.44 × 10−10 m2/s and the anion accessible porosity ε of approximately 0.2. In the presence of Fe(II), De for Tc migration in illite was one order of magnitude lower, showing that Fe(II) has a significant effect on the migration of 99Tc. Analysis of the Tc distribution in the system suggests that most Tc was retarded at the filters, especially the ones connected to the high concentration reservoir. The remaining Tc quantities were immobilized at sample boundaries next to the filters with higher concentrations observed in the domains close to the reservoirs. Almost no Tc was immobilized in the middle part of the sample, where the Fe(II) was preloaded. This observation contradicts the anticipation, because Tc was expected to be immobilized in the middle of the sample preloaded with Fe(II). A possible explanation is that a delocalized redox reaction may occur, which means that electrons from the oxidation of the preloaded Fe(II) in the central part of the cell are transferred to the filter via the walls of the steel diffusion cell. Tc reacts with the electrons in the filter at a relative slow rate, which results in most of the Tc retarded in the filter, while some small amounts of it may diffuse through the clay.

MoO3 Solubility and Chemical Durability of V2O5-Bearing Borosilicate Glass

In the vitrification of high-level radioactive liquid waste (HLW), the separation of sodium-molybdate melts is a problem because it reduces the chemical durability of the vitrified waste. A glass with both high MoO3 solubility and chemical durability is required for the safe disposal of radioactive waste. In this study, we investigate the effects of vanadium oxide on the phase separation of the molybdenum-rich phase and the water resistance of the resulting glass by phase equilibrium experiments and chemical durability test. Phase equilibrium experiments were performed on SiO2-B2O3-Al2O3-ZnO-CaO-Na2O-LiO2-MoO3 system glasses and on glasses with V2O5 added. The results showed that MoO3 solubility increased when V2O5 was added. The increase in MoO3 solubility in borosilicate melts may be associated with the viscosity-lowering effect of V2O5. Chemical durability tests were performed on borosilicate glass compositions obtained from phase equilibrium experiments. The normalized leaching rates of V2O5-bearing glasses were higher than those of other glasses. This is due to the higher network modifier/network former ratio of the glass tested. The normalized elemental mass loss of glass containing waste components increases with increasing leaching duration. This suggests that the waste component prevents the formation of a gel layer at the reaction front.

Effect on structure, stability, and H+ irradiation on Nd3+/Ru4+‐loaded iron phosphate glass for nuclear applications

AbstractNuclear energy generation technology is critically linked with the safe disposal of radioactive waste. In this context, iron phosphate glass (IPG) is gaining predominance as nuclear waste vitrification matrix that necessitates a thorough study on the effect of the loading of various nuclear fission waste materials in it. In this study, the effect of the loading of Nd3+ (which acts as a surrogate for radioactive curium (Cm)) and Ru4+ (which is a fission product of 235U) in IPG has been assessed. The optimum loading of Nd3+/Ru4+ leading to the formation of homogenous melt has been ascertained via powder X‐ray diffraction and scanning electron microscopy techniques. The modification in the Fe3+/Fe2+ ratio in IPG and the consequent change in its average coordination number with Nd3+/Ru4+ loading has been deduced from the Mössbauer studies. Local structure analysis has been done using X‐ray absorption spectroscopy at Nd/Ru/Fe K‐edge (as applicable) for all the single and co‐loaded IPG samples. All the co‐loaded samples show enhanced glass stability and glass forming ability compared to unloaded IPG which has been ascertained via detailed thermal studies. The variation in IPG network structure on the addition of Nd2O3 and RuO2 has been ascertained through spectroscopic techniques like Fourier transform infrared (FTIR) and Raman. The base glass and a few representative homogenous single and co‐loaded IPG samples have been irradiated with 4.5 MeV proton beam to simulate the hosting of radioactive elements and the radiation effect on glass structure has been ascertained using FTIR and Raman spectroscopies. The suitability of IPG as nuclear waste vitrification matrix for Nd3+ and Ru4+ is established through all the above analyses.

Numerical investigation of heating and cooling-induced damage and brine migration in geologic rock salt : Insights from coupled THM modeling of a controlled block scale experiment

Controlling brine flow is important for the safe disposal of radioactive waste in salt rocks. Thermal and mechanical processes can have a significant impact on brine flow, although this has never been thoroughly investigated. In this study, we conduct fully coupled THM modeling for analyzing brine flow in rock salt, considering non-isothermal two-phase flow through deformable porous media. To rigorously represent rock salt behavior, we incorporate suitable phenomenological models for creep and shear and tensile-induced dilatancy/damage, and their effect on the flow properties. To validate such a complex model, we use it to analyze an experiment on a meter-scale salt block subjected to multistage heating and cooling under controlled laboratory conditions. The experimental data and our model predictions of temperature and brine inflow show good agreement. Our modeling shows that it is important to consider the coupling between the heating- and cooling-induced damage and the flow properties for rigorously estimating brine inflow. Based on our modeling, a cooling-induced brine inflow spike is caused by a permeability increase due to tensile dilatancy. Thus, our modeling will be useful for a proper design of a cooling phase in the salt repositories to avoid or minimize damage and the resulting brine inflow.

Disposal of Radioactive Waste Generated at the End of 18F- Fluorodeoxyglucose Production for Positron Emission Tomography - Computed Tomography (PET-CT) Scan at Armed Forces Institute of Radiology and Imaging, Rawalpindi

Objective: To find out the optimal timeline for the safe disposal of radioactive wastes generated at the end of normal and anomalous productions of 18F-Fluorodeoxyglucose for positron emission tomography-computed tomography scans.
 Study Design: Quasi-experimental study.
 Place and Duration of Study: Armed Forces Institute of Radiology and Imaging, Rawalpindi Pakistan, from Aug to Dec 2020.
 Methodology: Wastes generated at the end of 18F-Fluorodeoxyglucose production were tested for residual radioactivity at 2, 12 and 24 hours after synthesis to ascertain the optimal timeline for discarding. The radiation dose at 5 cm from the surface of the container was also assessed.
 Results: Fifty productions were included in the study, of which 46(92%) were normal, and 4(8%) were anomalous productions. The mean activity at 2 and 24 hours after the end of synthesis was 4.102±0.831mCi and 0.0047±0.00116mCi, respectively, in normal productions and 45.125±2.332mCi and 0.005±0.00026mCi respectively in anomalous productions. The mean radiation dose at 5cm from the surface of the container at 2 and 24 hours after the end of synthesis was 8.2107±1.665mSv/h and 0.00000966±0.00000212mSv/h in normal productions and 90.32±4.66mSv/h and 0.00975±0.0005-mSv/h respectively in anomalous productions.
 Conclusion: The residual radioactivity in wastes was negligible 24 hours after the end of synthesis in both normal and anomalous productions. Radioactive wastes from 18F-Fluorodeoxyglucose production should, therefore, be conserved for at least 24 hours before their disposal to the environment.

Radioactive waste management: Societal challenges in the era of green nuclear energy

Abstract Recently, there have been international debates over the idea of labeling nuclear power plants as a green investment. To be considered sustainable, to get a green investment label and to contribute to green growth, nuclear energy should be based on clear plans regarding safe disposal of radioactive waste. Final disposal of radioactive waste is witness to many disputes, one of the most important being that of public acceptance. Society represents a critical and a decisive stage in the process of radioactive waste management. Management and disposal process of radioactive waste requires community confidence and acceptance. The socio-political context must be addressed continuously through stakeholder commitment and public concerns. The final stage of radioactive waste management is especially characterized by population involvement, stakeholder availability, and mutual dialog. The objective of this study is based on finding the relationship between the location of a radioactive waste disposal facility (RWDF) and the perception of the population regarding various aspects related to nuclear energy and radioactive waste. A series of questions in the form of a questionnaire were used for data collection. A total of 200 valid questionnaires were used for this analysis. The study was focused on the Romanian population following a homogenous distribution throughout the country. Therefore, to better understand the issue, the paper investigates the correlation between the location of a radioactive waste disposal facility and a variety of factors, such as: attitude, perceived benefits, perceived risks, overall knowledge and perceived costs. Spearman rank correlation was used for data analysis. Bringing together theoretical information about nuclear energy and radioactive waste and the empirical data collected at the national level, this research study showed that the acceptance of a radioactive waste disposal facility is much stronger if the population is in favor of nuclear energy. As well, perceived risks create major concerns throughout the population and influence the location of a radioactive waste disposal facility. Cost aspects are controversial and create disagreements in relation with radioactive waste management options. That being said, this study brings value to the nuclear field by emphasizing the important societal factors which influence the location of a radioactive waste disposal facility in the green energy era.

THEREDA – Thermodynamic Reference Database

Abstract. Part of the process to ensure the safety of radioactive waste disposal is the predictive modeling of the solubility of all relevant toxic components in a complex aqueous solution. To ensure the reliability of thermodynamic equilibrium modeling as well as to facilitate the comparison of such calculations done by different institutions, it is necessary to create a mutually accepted thermodynamic reference database. To meet this demand several institutions in Germany joined efforts and created THEREDA (Moog et al., 2015). THEREDA is a suite of programs at the base of which resides a relational databank. Special emphasis is put on thermodynamic data along with suitable Pitzer coefficients, which enable the calculation of solubilities in high-saline solutions. Registered users may either download single thermodynamic data or ready-to-use parameter files for the geochemical speciation codes PHREEQC, Geochemist's Workbench, CHEMAPP, or TOUGHREACT. Data can also be downloaded in a generic JSON format to enable the import into other codes. The database can be accessed via the world wide web: http://www.thereda.de (last access: 1 November 2021). Prior to release, the released part of the database is subjected to many tests. Results are compared to results from earlier releases and among the different codes. This is to ensure that by additions of new and modification of existing data no adverse side effects on calculations are caused. Furthermore, our website offers an increasing number of examples for applications, including graphical representation, which can be filtered by components of the calculated system.

Efficacy of a Graphene Oxide/Chitosan Sponge for Removal of Radioactive Iodine-131 from Aqueous Solutions.

Iodine-131 is increasingly used for diagnostic and therapeutic applications. The excretion of radioactive iodine is primarily through the urine. The safe disposal of radioactive waste is an important component of overall hospital waste management. This study investigated the feasibility of using graphene oxide/chitosan (GO/CS) sponges as an adsorbent for the removal of iodine-131 from aqueous solutions. The adsorption efficiency was investigated using iodine-131 radioisotopes to confirm the results in conjunction with stable isotopes. The results revealed that the synthetic structure consists of randomly connected GO sheets without overlapping layers. The equilibrium adsorption data fitted well with the Langmuir model. The separation factor (RL) value was in the range of 0–1, confirming the favorable uptake of the iodide on the GO/CS sponge. The maximum adsorption capacity of iodine-131 by GO/CS sponges was 0.263 MBq/mg. The highest removal efficiency was 92.6% at pH 7.2 ± 0.2. Due to its attractive characteristics, including its low cost, the ease of obtaining it, and its eco-friendly properties, the developed GO/CS sponge could be used as an alternative adsorbent for removing radioiodine from wastewater.

Investigation of the insulating properties of the Cherkasy deposit clays for the creation of underlying screens of radioactive waste at the ‘Vector’ site

The lack of scientifically substantiated requirements, comprehensively developed and approved in a prescribed manner, for the usage of clays as a barrier material poses risks for the safe disposal of radioactive waste in facilities at the ‘Vector’ site for the period of their operation and closure. The bentonite clay from Ukraine’s largest Cherkasy deposit of bentonite and palygorskite clays is considered the most durable as the main component of the insulating (underlying) screens of radioactive waste disposal facilities. The main properties and compositional features of the Cherkasy natural bentonite clay (Dashukovskaya site, layer II) and its variety such as alkaline earth bentonite (activated soda bentonite), which provide isolation of radioactive waste in disposal, are considered. It is shown that the Cherkasy field has good waterproofing and barrier properties, including a high sorption capacity with respect to 90Sr and 137Cs, which is one of the main characteristics that ensure the safe disposal of radioactive waste. The alkaline earth bentonite absorbs 90Sr and 137Cs more efficiently than natural bentonite does. However, 90Sr is sorbed in larger quantities than 137Cs on both types of bentonite. With increasing time of interaction with an aqueous solution, both types demonstrate a redistribution of the mobile (exchangeable) and immobile (non-replaceable) forms of radionuclides. The contribution of the stationary form that does not participate in migration processes also increases. A comprehensive analysis of the bentonite clays of the Cherkasy deposit was carried out, taking into account the significance of recoverable reserves and the potential for improving the technical and economic parameters of clays. Thus, the Cherkasy bentonite clays can be recommended as an additional anti-migration engineering barrier for ground/near-surface facilities for the disposal of radioactive waste. When choosing the type of bentonite clay for use as a barrier in a radioactive waste disposal facility, one could take into account the data published in the article, but the question of applying the bentonite clays of the Cherkasy deposit to ensure the safe disposal of radioactive waste remains to be  further studied.

Результаты исследований в существующих скважинах на участке недр «Енисейский», в т . ч . для определения основных систем трещин и анизотропии массива пород

The conditions in the crystalline geological massif, which determine the long-term safety of radioactive waste disposal, are reduced to two fundamental factors - the features of the host rocks and the fracture network inherent in the studied area of the massif [3]. Refinement of the parameters identified at the early stages of geological study of cracks is one of the highest priority tasks in the safety assessments of the radioactive waste disposal. These parameters should also be taken into account in the formation of requirements for the composition and structure of the system of engineering barriers preventing the release of radionuclides outside the disposal. The presented material contains a description of the work and the approaches applied in processing the results of borehole surveys of the prospecting and appraisal stages of geological exploration of the Yeniseisky area, performed using the specialized probe with video logging equipment.

Efficient cesium encapsulation from contaminated water by cellulosic biomass based activated wood charcoal

In this study, cost-effective cellulosic biomass based activated wood charcoal was developed from Japanese Sugi tree (Cryptomeria japonica) by concentrated nitric acid modification for adsorption of Cs from contaminated water. The physicochemical properties of specimens were investigated using N2 adsorption-desorption isotherms (BET method), FESEM, FTIR, and XPS spectra analysis. The experimental results revealed that the surface area of the raw wood charcoal was significantly decreased after boiling nitric acid modification. However, several oxygen-containing acidic function groups (-COOH, –CO) were introduced on the surface. The adsorption study confirmed that the equilibrium contact time was 1 h, the optimum adsorption pH was neutral to alkaline and the suitable adsorbent dose was 1:100 (solid: liquid). The maximum Cs was removed when the concentration of Na and K were lower (5.0 mM) with Cs in solution. The Cs adsorption processes well approved by the Langmuir isotherm and pseudo-second-order kinetic models and the maximum adsorption capacity was 35.46 mgg−1. The Cs adsorption mechanism was clearly described and it was assumed that the adsorption was strongly followed by chemisorptions mechanism based on the adsorbent surface properties, kinetic model and Langmuir isotherm model. Most importantly, about 98% of volume reduction was obtained by burning (500 °C) the Cs adsorbed charcoal, which ensured safe storage and disposal of radioactive waste. Therefore, this study can offer a guideline to produce a functional adsorbent for effective Cs removal and safe radioactive waste disposal.