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Radionuclide transport in smectite clay barrier systems used for nuclear waste disposal is controlled by diffusion, with adsorption significantly retarding transport rates. While a relatively minor component of spent nuclear fuel, 79Se is a major driver of the safety case for spent fuel disposal due to its long half-life (3.3 × 105 yr) and its low adsorption to clay (KD < 10 L/kg), thus a thorough understanding of Se diffusion through clay is critical for understanding the long-term safety of spent fuel disposal systems. Through-diffusion experiments with tritiated water (HTO, conservative tracer) and Se(VI) were conducted with a well-characterized, purified montmorillonite source clay (SWy-2) under a constant ionic strength (0.1 M) and three different electrolyte compositions: Na+, Ca2+, and a Na + -Ca2+ mixture at pH 6.5 in order to probe the effects of electrolyte composition and interlayer cation composition on clay microstructure, Se(VI) aqueous speciation, and ultimately diffusion. The results were modeled using a reactive transport modeling approach to determine values of porosity (ε), De (effective diffusion coefficient), and KD (distribution coefficient for adsorption). HTO diffusive flux was higher in Ca-montmorillonite (De = 1.68 × 10−10 m2 s−1) compared to Na-montmorillonite (De = 7.83 × 10−11 m2 s−1). This increase in flux is likely due to a greater degree of clay layer stacking in the presence of Ca2+ compared to Na+, which leads to larger inter-particle pores. Overall, the Se(VI) flux was much lower than the HTO flux due to anion exclusion, with Se(VI) flux following the order Ca (De = 1.03 × 10−11 m2 s−1) > Na–Ca (De = 2.12 × 10−12 m2 s−1) > Na (De = 1.28 × 10−12 m2 s−1). These differences in Se(VI) flux are due to a combination of factors, including (1) larger accessible porosity in Ca-montmorillonite due to clay layer stacking and smaller electrostatic effects compared to Na-montmorillonite, (2) larger accessible porosity for neutral-charge CaSeO4 species which makes up 32% of aqueous Se(VI) in the pure Ca system, and (3) possibly higher Se(VI) adsorption for Ca-montmorillonite. Through a combination of experimental and modeling work, this study highlights the compounding effects that electrolyte and counterion compositions can have on radionuclide transport through clay. Diffusion models that neglect these effects are not transferable from laboratory experimental conditions to in situ repository conditions.

Rock salt is considered a suitable medium for the permanent disposal of heat-generating radioactive waste due to its isolation properties. However, excavation damage and heating induce complex and heterogeneous thermal-hydrological-mechanical (THM) processes across different zones. Quantifying this heterogeneity is crucial for accurate long-term performance assessment models, but traditional methods lack the necessary resolution. This study employs 4D electrical resistivity tomography (ERT) monitoring during controlled heating experiments in a salt formation to unravel the spatiotemporal dynamics of THM processes. Advanced time-lapse inversion and clustering analysis quantify subsurface properties and map the heterogeneity of THM dynamics. The ERT results can estimate subsurface properties and delineate the damaged and intact zones, enabling appropriate parameterization and representation of processes for long-term modeling. This approach may be used in further improving the predictive models and ensuring the safe long-term disposal of radioactive waste in rock salt.

Ocean disposal of industrial waste from technical DDT [mainly 1,1-(2,2,2-trichloroethane-1,1-diyl)bis(4-chlorobenzene), or 4,4-DDT] manufacture occurred historically in the Southern California Bight. However, the paucity of historical records highlights uncertainties as to the mode, location, and timing of disposal or ongoing ecological effects of these wastes. This study combines sampling, chemical analysis, and numerical modeling of deep San Pedro Basin sediments revealing substantial DDT contamination that extends at least 25 km from the mainland. These findings narrate bulk DDT waste disposal to the offshore that peaked in the 1950s, prior to the onset of formal regulations; was agnostic to later-designated disposal sites; and has experienced sluggish transformation. Our findings further indicate an attenuating secondary source for the DDT daughter product, 1-chloro-4-[2,2-dichloro-1-(4-chlorophenyl)ethenyl]benzene (4,4-DDE), which still deposits into deep San Pedro Basin sediments. While demonstrating the severity of DDT contamination to the region, these findings further define the burial potential of DDT wastes and inform the past, present, and future contamination potential that is needed to understand and predict ecological consequences. This work also points firmly to bulk, not containerized, disposal of DDT waste and to potential alternative contents of collocated waste.

Understanding mass transport in the alteration layers of glass surfaces is a crucial component of the safety assessment of nuclear waste glass. In this work, we model such an alteration layer as a silicate gel with water through a molecular dynamics (MD) simulation with a reactive force field. Gels with various water contents (WCs) ranging from 5.1 to 30.7wt%$30.7 \,{\rm wt}\%$ are produced via high‐temperature annealing with water and silica. It is found that an increase in the water content destroys the polymerized structure of the silicate network and promotes the formation of silanol groups. The pore size and water connectivity formed by the silicate networks are investigated for the modeled gels. Gel with a WC of 5.7wt%$5.7 \,{\rm wt}\%$ is composed of isolated water in the pores; in contrast, pores filled with interconnected water are formed in gel with a WC of 30.7wt%$30.7 \,{\rm wt}\%$. The water diffusivity in the modeled gel is calculated using the mean‐squared displacement at various temperatures. An attempt is made to formulate a linear relationship between the water diffusivity and porosity derived from the MD simulation. The porosity is calculated using a probe atom with a radius, which was optimized from a linear relationship between the water diffusivity and porosity. This approach successfully explains the water diffusivity in terms of the porosity. Random walk (RW) simulations for the structures derived from the MD simulations are performed to determine the geometrical effects of the pores. The diffusivity obtained from RW simulation is compared with the results of the MD simulations, which include chemical interactions such as the formation and breakage of hydrogen bonds. This comparison highlights how geometrical effects and chemical interactions contribute to water diffusivity depending on the WC. [ABSTRACT FROM AUTHOR]

We investigated the radiolytic hydrogen production of geopolymers that are a candidate material for solidifying radioactive waste. The effects of water content and solid properties on the hydrogen production were evaluated by 60Co γ-ray irradiation using geopolymers with and without boric acid, which controls the setting time of the geopolymer. We also constructed a preliminary model of the geopolymer pore structure and calculated the dose to water on the pore surface using Monte Carlo simulation. Hydrogen production was influenced by water content, specific surface area, and electron density of elements in the solid, and increased as these values increased. Simulation results suggested that the amount of hydrogen was affected by the dose of Compton electrons generated by γ-ray irradiation on the geopolymer solid to water on the pore surface. The apparent hydrogen production yield of the geopolymer containing boric acid was 1.1 × 10−8 mol/J, less than 1/4 of that of water. When this sample was heated over 800°C, it decreased to 1.0 × 10−9 mol/J. In contrast, the apparent hydrogen production yield based on the energy absorbed by water in the geopolymer was larger than that of water and increased with a decrease in water content. Furthermore, it decreased below 0.2 wt.% water content. [ABSTRACT FROM AUTHOR]

By means of vibrational spectroscopy, the study examines the structure of matrix materials synthesized through the quenching of sodium-rubidium aluminoborosilicate melts with the addition of zirconium. Zirconium was found to have a significant effect on the ratio of the main structural units and the distribution of modifier cations among them in the glass structure. The obtained results were used to explain changes in the glass-melt transition temperature and synthesized material density; they can be useful in adjusting the composition and synthesis parameters of matrix materials for the immobilization of high-level radioactive waste containing significant amounts of zirconium. [ABSTRACT FROM AUTHOR]

The oil ash radioactive waste was generated from electrical power plants and it was presented great concern due to environmental impact. In this study oil ash was characterized from technical and radiological aspects. Oil ash average activity concentrations for 226Ra, 232Th, and 40K were (1718 ± 85.9, 278.1 ± 13.9, and 136 ± 6.7) Bq/kg respectively, that were higher than the worldwide average. The average value of AEDtot was (10.5 ± 0.5) mSv/y, which was higher than the dose limit of public. All the radiological parameters were higher than worldwide. The samples examined by XRF, and XRD. That contained economic elements as iron, vanadium, nickel. [ABSTRACT FROM AUTHOR]

The 17 Sustainable Development Goals (SDGs) of the United Nations, established in 2015, aim to achieve global sustainability by 2030 through the improvement of environmental, social, and economic parameters. However, unlike earlier concepts such as the Agenda 21 of 1992, the SDGs overlook radioactive waste management and related challenges of radiation itself. First, we investigate the historic consideration and unexplained disappearance of radioactive waste in earlier sustainability concepts. Then, we propose amending seven SDGs to address this gap. For SDGs 6 (Clean Water and Sanitation), 14 (Life Below Water), and 15 (Life on Land), new or revised indicators should monitor the release of hazardous materials. SDGs 9 (Industry, Innovation and Infrastructure), 16 (Peace, Justice and Strong Institutions), and 17 (Partnerships for the Goals) require additional targets and indicators to integrate international cooperation and social implications of nuclear facilities' operation. Redefining "hazardous waste" in SDG 12 (Responsible Consumption and Production) and "environmentally sound technologies" in SDG 17 is necessary to encompass radioactive waste. Implementing these changes demands statistical efforts, but the existing monitoring infrastructure, particularly in Europe and North America, can facilitate this. As 2030 approaches, it is crucial to reintroduce radioactive waste management into sustainability agendas, whether within the SDGs themselves or in a subsequent framework. [ABSTRACT FROM AUTHOR]

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During vitrification of radioactive wastes, excessive foaming reduces processing rates within melters by hindering heat transfer from the molten glass pool to the reacting melter feed. Formulations of low‐activity waste (LAW) melter feeds, for vitrification at the Waste Treatment and Immobilization Plant at the Hanford Site, conventionally include the addition of sucrose to mitigate excessive foaming by hastening the denitration process. However, incomplete combustion of sucrose produces organics such as acetonitrile (C2H3N) that may exceed bounding limits of downstream effluent treatment facilities. Using boron nitride (BN) as an alternate reductant to sucrose, in a representative LAW melter feed reduced C2H3N production by 90% by preventing the low‐temperature sucrose–nitrate reactions. Furthermore, foaming was suppressed due to the higher decomposition temperature of BN than H3BO3 meaning a delayed reaction of a large fraction of boron with the transient glass‐forming melt until above the foam onset temperature, thus reducing the quantity and viscosity of the connected melt and trapping less gas in the foam layer. [ABSTRACT FROM AUTHOR]

The activation product chlorine-36 (36Cl) is an important radionuclide within the context of the disposal of nuclear wastes, due to its long half-life and environmental mobility. Its behaviour in a range of potential cementitious encapsulants and backfill materials was studied by evaluating its uptake by pure cement hydration phases and hardened cement pastes (HCP). Limited uptake of chloride was observed on calcium silicate hydrates (C-S-H) by electrostatic sorption and by calcium monosulphoferroaluminate hydrate (AFm) phases, due to anion exchange/solid solution formation. Diffusion of 36Cl through cured monolithic HCP samples, representative of cementitious materials considered for use in deep geological repositories across Europe, revealed a markedly diverse migration behaviour. Two of the matrices, a ground granulated blast furnace slag/ordinary Portland cement blend (GGBS–OPC) and an ordinary Portland cement (CEM I) effectively retarded 36Cl migration, retaining the radionuclide in narrow, reactive zones. The migration behaviour of 36Cl within the cementitious matrices is not strictly correlated to the measured sorption distribution ratios (Rd-values), suggesting that physical factors related to the microstructure can also have a distinct effect on diffusion behaviour. The findings have implications when selecting cementitious grouts and/or backfill materials for 36Cl-bearing radioactive wastes. [ABSTRACT FROM AUTHOR]

For the peaceful use of nuclear energy, materials, and technology, various methods have been developed for safeguards purposes. Among them, the In-Situ Object Counting System has demonstrated its usefulness and reliability in various nuclear material verification activities, due to its capability of operating without standard materials. In the present study, the performance of In-Situ Object Counting System for uranium quantification was evaluated using an experimental drum at KEPCO NF as well as low-level radioactive waste drums at the Korea Radioactive Waste Agency. The results showed that In-Situ Object Counting System could quantify uranium in radioactive waste drums, with a relative error ranging from −1 to −67 % compared with destructive assay results. This study also examined, using the Monte Carlo N-particle transport 6.2 code, the factors that affect uncertainty in uranium quantification: geometrical conditions, spectrum quality, heterogeneity, and the self-shielding effect. These findings could prove useful in efforts to optimize the In-Situ Object Counting System as a tool for nuclear material verification in the safeguards context.

The high-throughput, plasma-based, mass separation techniques of nuclear waste are still an active research topic due to the theoretical and practical complexity of processing such waste. The nuclear waste can be separated according to its masses by plasma centrifugation technique. Because of this, the present research has been performed by setting up $ \vec{E} \times \vec{B} $ E→×B→ rotation in cold cylindrical plasma surrounded by an ideally conducting homogenous shell inside a plasma centrifuge device. The results obtained are useful in analyzing the magneto-rotational instability (MRI) and trajectories of the plasma particles which are beneficial for improving the throughput and separation factor of the device. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

The optimal conditions for the synthesis of a pelletized sorbent based on mixed Mn(III, IV) oxide by reacting aqueous solutions of MnSO4 and KMnO4 in an alkaline medium were determined: Mn2+/MnO4– molar ratio 1.70–1.80; pH of the reaction mixture 11.0–12.5; calcination temperature 220°C. For the sorbent fabricated under optimal conditions, the values ​​of the distribution coefficient (Kd) of 90Sr in 0.01 M CaCl2 solution, static exchange capacity for calcium, hydromechanical strength of granules, and Kd of 90Sr as a function of the concentration of sodium and calcium ions were established. The sorbent afforded higher sorption characteristics with respect to strontium compared to known sorbents. A technology for producing pilot batches of the sorbent named MDM, was developed. Examples of the use of the MDM sorbent for the purification of various types of liquid radioactive waste from strontium and radium radionuclides are reported. [ABSTRACT FROM AUTHOR]

Microstructural evolution of borosilicate glass–sulfate (BaSO4 or Na2SO4) mixture was observed in situ using a high‐temperature confocal scanning laser microscope. The phase compositions and sulfate retention of the mixture at different temperatures were also examined. The results indicate that for the glass–BaSO4 sample (abbreviated as BS), the borosilicate glass begins to melt at about 900°C, and the triangular BaSO4 crystals grow obviously between 900°C and 1100°C. At temperatures above 1100°C, the BaSO4 crystals gradually melt and decompose. For the glass–Na2SO4 sample (abbreviated as NS), strip‐shaped BaSO4 crystals begin to form at about 820°C after the borosilicate glass melting, and then gradually melt with further increasing temperature. When the temperature increases to 1000°C, the upper layer of the NS sample possesses Na2SO4, LiNaSO4, and BaSO4 phases, while the lower layer possesses an amorphous phase. Moreover, the SO3 retention in the BS sample remains almost unchanged with temperature increases from 900°C to 1100°C, and then decreases with further increasing temperature. The SO3 retention retained in the NS sample decreases rapidly at temperatures above 900°C. These findings can provide a guidance for vitrifying sulfur‐rich high‐level liquid waste. [ABSTRACT FROM AUTHOR]

High-level waste (HLW) separation into fractions simplifies isolation in matrices and disposal. One of these fractions consists of rare earths (REE) and minor actinides (MA = Am, Cm). Among the rare earth elements there are the isotopes 144Ce, 147Pm, 151Sm, and 154,155Eu with half-lives of up to 93 years; decay of the nuclides will cause heating of the matrices with HLW. It has been shown that preliminary storage of the REE–MA matrix for ten years or more will significantly reduce the content of REE radioisotopes and their contribution to temperature increase in the geological repository. [ABSTRACT FROM AUTHOR]

In this study, we developed a high efficiency sample pretreatment technique using microwave-assisted acid digestion (MAD) under a mixture of HNO3-HCl-HF-H2O2 conditions, capable of leaching various radioactive nuclides from contaminated soil within 2 h. Our validation results using SRM 2709a soil standard material confirmed that when the MAD process was performed and then repeated, the chemical recovery rates of all 9 elements were consistently above 90%, demonstrating an outstanding leaching efficiency compared to conventional acid leaching (AL) methods. The activity of radionuclides obtained through AL were found to be underestimated compared to those obtained through MAD. We anticipate that the application of this technique for precise activity of radionuclides will contribute to cost savings in decommissioning waste management. [ABSTRACT FROM AUTHOR]

Zirconolite ceramics have been proven to be a promising matrix for high-level waste immobilization, especially for minor actinides. In this study, a series of CaZr 1-2x REE 2x Ti 2-x Mo x O 7 (REE = Nd, Sm, Gd, Ho, Yb) samples were prepared to investigate the formation of zirconolite REE as surrogates of minor actinides and Mo6+ as charge compensator. Synchrotron and in-house powder X-ray diffraction (XRD), X-ray absorption near edge spectroscopy (XANES), and scanning electron microscopy-energy dispersive X-ray spectroscopy (SEM-EDX) were used to study their solid solubility, phase evolution, and structural details. XRD results show that the formation of zirconolite-2M can be up to x = 0.15 in the CaZr 1-2x Nd 2x Ti 2-x Mo x O 7 samples when co-existence of 2 M and 4 M was observed at x = 0.20 and 0.25. The "yellow phase" appears when x exceeds 0.4. The phase evolutions of CaZr 1-2x Sm 2x Ti 2-x Mo x O 7 , CaZr 1-2x Gd 2x Ti 2-x Mo x O 7 , CaZr 1-2x Ho 2x Ti 2-x Mo x O 7 and CaZr 1-2x Yb 2x Ti 2-x Mo x O 7 solid solutions are similar to each other: (1) zirconolite-2M in x = 0.05 and 0.10; (2) co-existence of 2 M and 4 M and then formation of 4 M phase from x = 0.15 to 0.25; (3) formation of "yellow phase" and pyrochlore when x ≥ 0.30. Interestingly, perovskite exits in all the ceramic samples, and its lattice parameters slightly change when increasing the dopants. Rietveld refinement results reveal that Sm/Ho mainly incorporates into 8f (Wyckoff position) Zr-sites while Mo replaces the 8f (Wyckoff position) Ti-sites. XANES results further confirm the occupation of Mo in TiO5 coordination environments. Furthermore, there is a considerable amount of Sm/Ho incorporated into 8f Ca-sites, demonstrating a different substitution mechanism from the preliminary design. [ABSTRACT FROM AUTHOR]

Radiation and radioactive substances result in the production of radioactive wastes which require safe management and disposal to avoid risks to human health and the environment. To ensure permanent safe disposal, the performance of a deep geological repository for radioactive waste is assessed against internationally agreed risk‐based standards. Assessing postclosure safety of the future system's evolution includes screening of features, events, and processes (FEPs) relevant to the situation, their subsequent development into scenarios, and finally the development and execution of safety assessment (SA) models. Global FEP catalogs describe important natural and man‐made repository system features and identify events and processes that may affect these features into the future. By combining FEPs, many of which are uncertain, different possible future system evolution scenarios are derived. Repository licensing should consider both the reference or "base" evolution as well as alternative futures that may lead to radiation release, pollution, or exposures. Scenarios are used to derive and consider both base and alternative evolutions, often through production of scenario‐specific SA models and the recombination of their results into an assessment of the risk of harm. While the FEP‐based scenario development process outlined here has evolved somewhat since its development in the 1980s, the fundamental ideas remain unchanged. A spectrum of common approaches is given here (e.g., bottom–up vs. top–down scenario development, probabilistic vs. bounding handling of uncertainty), related to how individual numerical models for possible futures are converted into a determination as to whether the system is safe (i.e., how aleatoric uncertainty and scenarios are integrated through bounding or Monte Carlo approaches). [ABSTRACT FROM AUTHOR]

The object of research is the spent open-pit mines themselves where the proposed system could be applied. The primary reason for this studying is the following circumstance: up to the present time period, in all countries of the world, no methods of HLW disposal in a storage facility has been identified that is absolutely safe for any length of time, taking into account the impact of catastrophic natural emergencies and man-made emergencies. The research was conducted to address the problem of safely managing and storing HLW, leveraging the unique characteristics of spent open-pit mines, such as their large volume and geological stability, to prevent environmental contamination and ensure long-term safety. In the article has been justified a novel approach to the burial of sarcophagus containers with solid HLW in exhausted mining pits and studied the usabilities of the basalt sarcofagous container. Robust materials and advanced robotic systems proposed in the article aims to address the challenges associated with long-term radioactive waste disposal effectively. The robotic systems transfer the basalt container with HLW, ensuring personnel safety by minimizing human presence near radioactive materials. In the article have been established the distribution of temperature into the multi-layered composite structure of the basalt sarcophagous with HLW from 300 °C into the inner space to 50 °C onto on the its outer suffer where the thickness of each layers (from inner to outer radius) was respectively: for lead matrix: from r1=0.1 m to r2=0.2 m; for clay layer: from r2=0.2 m to r3=0.3 m; for basalt block: from r3=0.3 m to r4=0.4 m. The findings on temperature distribution are crucial as they directly affect the performance and longevity of the basalt containment system [ABSTRACT FROM AUTHOR]

A novel method has been developed for the simultaneous determination of difficult-to-measure (DTM) nuclides 93Zr, 237Np, 232Th, 230Th and 228Th using a DGA resin® (containing N,N,N′,N′-tetra-n-octyl-diglycolamide) column for simultaneous separation of Zr and actinides followed by purification of Zr, Np and Th on a TEVA resin® (containing Aliquat 336) column. Samples are destroyed by fusion with sodium hydroxide or by acid destruction using HNO3, HCl and HF. Pre-concentration procedure tailored to the extraction chromatography is based on co-precipitation of Zr and actinides and removal of alumina, silica, iron and calcium. The concentrations of 93Zr, 237Np and Th nuclides are determined by ICP-MS and α spectrometry, respectively. High chemical recoveries (≥ 73%) and separation factors were obtained for Zr, Np and Th both in evaporation concentrates of an NPP and various samples of mineral origin such as soil and concrete. The procedure can be easily combined with the simultaneous separation of other actinides (Pu, Am, U) using the same DGA column that has been described in our previous papers (Cassette et al. in Appl Radiat Isot 68:122–130, 2010 and Groska et al. in J Radioanal Nucl Chem 309(3):1145–1158, 2016). [ABSTRACT FROM AUTHOR]

90Sr is one of the most monitored fission products. New efficient methods for strontium separation are therefore sought. Over the past 20 years, crown ethers solutions in ionic liquids (ILs) have been confirmed to efficiently extract strontium from acidic solutions. Despite all the work done in this field, the extraction mechanism is still not completely clear. Depending on experimental conditions, the extraction is believed to proceed either in the form of an ion-association complex [Sr·CE·(NO3)2] or via exchange between [Sr·CE]2+ and two IL cations. This work aims to study the influence of several extraction parameters and compare them with the presumed mechanisms. [ABSTRACT FROM AUTHOR]

This study -- which employs the methodology of archive-based historiography -- focuses on the awkward elaboration of an Italian «roadmap» to deal with nuclear waste and, notably, the crucial role of a national repository. Drawing on a number of unpublished archival sources, but crucially on uncatalogued original documents of ENEA -- then the Italian nuclear authority -- the study investigates the specific contribution of experts and scientists, and their relationship with political decision-makers. Preliminary conclusions: in the Seventies, the basic assumption was that nuclear waste would be reprocessed in a single national facility, but against the ever-decreasing perspective for nuclear power in Italy, an open-cycle approach was adopted. By 1984 ENEA had prepared definite proposals on the possible repository sites, considering the issue of low- and medium-activity waste urgent. However, it received no acknowledgment of any kind by the Ministry of Industry in spite of repeated urgings. After Chernobyl the issue was buried, and eventually resurrected only in mid-Nineties. After 2001 political elections, the «roadmap» devised with great difficulty was turned upside down by favouring a top-down approach to the siting issue which would eventually backfire. [ABSTRACT FROM AUTHOR]

Background: Nuclear medicine is an integral and developing field in diagnosing and treating diseases. Monitoring individuals’ protection and radiation contamination in the workplace is vital for preserving working environments.Objective: This study aimed to monitor the nuclear medicine department’s personnel, environment, and wastes to determine the level of occupational radiation and environmental pollution in Bushehr’s nuclear medicine department.Material and Methods: In this cross-sectional study, the initial activity of each radioisotope, radiopharmaceutical, and radioactive waste was measured using a “well counter” daily for three months. Three irradiators’ absorbed doses were measured using a direct reading dosimeter. The contamination was determined using an indirect wipe test method on various surfaces. A Geiger Müller dosimeter was employed to examine personnel’s hands, clothing, and footwear.Results: The highest activity was observed in technetium waste (1118.31 mCi). Every irradiator received a lower absorption dose than the International Commission on Radiological Protection (ICRP) standard threshold. The majority of contamination was associated with the exercise test room (0.04 Bq/cm2) and its work surface (0.013 Bq/cm2), which were both below the threshold (0.5 Bq/cm2). Staff monitoring indicated that two nurses (10 and 11 individuals) had the highest contamination rate (23.7%). Conclusion: Daily assessment of the type, activity, and method of radiopharmaceutical administration to the patient is advantageous for waste management. Surface contamination monitoring can significantly contribute to the estimation of the level of radiation pollution in the environment.

This is an article in the field of mining environment. Most of the uranium mines discovered and mined in my country are located in Hunan, Jiangxi, and Guangdong. With the delineation of the ecological red line and the increasing attention of the people to ecological and environmental issues, environmental restoration of uranium mines becomes more and more important. This article briefly introduces the mining methods of uranium mines in southern China, systematically analyzes the impact of the production process of uranium mines on the environment, focuses on the principles and applications of physical, chemical, and biological remediation measures, summarizes the corresponding advantages and disadvantages, and prospects the future development of mine restoration. It was pointed out that the next step should be focused on environmental testing before the decommissioning of uranium mines and related facilities, and comprehensive environmental data collection and recording of the entire production process of the mine. Close attention should be paid to the treatment and disposal of wastewater from uranium mines, increase research investment, learn from foreign advanced technology and methods, and use comprehensive repair technology to repair the polluted environment more efficiently, safely and quickly. The restoration of uranium mines would be combined with rural governance, develop tourism, and increase the development motivation and income levels of local residents.

The current effort in environmental remediation is aimed at removing iodine-131 radionuclide from radioactive liquid waste produced by an Egyptian nuclear power plant using some synthesised alginate derivatives. Two different copolymers, namely sodium alginate poly (acrylic acid) (P1) and sodium alginate poly (acrylic acid-methacrylic acid) (P2), are prepared using gamma radiation. The ability of these polymers to remove 131I radionuclide as sorbents has been investigated. The synthesised polymers exhibit excellent adsorption performance for 131I ions, and the adsorption equilibrium requires only 30 min, which reveals that the sorption process is kinetically faster than most of the other materials reported previously. The removal percents for 131I radionuclide at a pH of 3.0 at room temperature on P1 and P2 are 77.7% and 84.2%, respectively. The sorption capacities of the two polymers demonstrate that P2 > P1, with capacities of 67.9 and 58.5 mg/g, respectively. Four linear kinetic models are investigated: pseudo-first order, pseudo-second order, Elovich, and Weber–Morris models. Regarding their calculated parameters, these models indicate that the adsorption process of I-ions on both P1 and P2 is controlled by chemisorption. Four equilibrium isotherm models (Redlich-Peterson, Langmuir, Freundlich, and Harkin-Jura) are investigated, revealing that the adsorption process is a monolayer and multilayer process on a heterogeneous surface. [ABSTRACT FROM AUTHOR]

This study is part of numerical simulations performed on an in-situ heating test conducted by the French National Radioactive Waste Management Agency (Andra) at the Meuse/Haute-Marne Underground Research Laboratory (URL) to study the thermo-hydromechanical behavior of the host Callovo-Oxfordian COx claystone in quasi real conditions, through the international research project DECOVALEX. We present a numerical study of damage and cracking process in saturated claystone subjected to thermo-hydromechanical coupling by considering material heterogeneity distribution. For this purpose, a macroscopic elastic model is first determined by using two steps of homogenization by taking into account the effects of porosity and mineral inclusions. This model is implemented into a finite element code devoted to solving thermo-hydromechanical coupling problems. The nucleation and propagation of cracks are described by using an extended phase-field method, considering the effects of temperature and fluid pressure on the evolution of phase-field. The proposed model is applied to the numerical analysis of cracking process due to excavation and heating around a group of boreholes (CRQ). The numerical results of the 3D simulation are compared with in-situ measurements of temperature and pore pressure distribution. The excavation damage zone and heating fracture is reproduced and analysed according to the structure of the heating position and the heterogeneity of the rock. Highlights: A phase-field model is developed with thermo-hydromechanical processes; Both tensile and shear cracks are taken into account; Crack nucleation is emphasized by spatial heterogeneity of material properties; Cracking processes due to thermal-hydraulic interaction are analyzed in three-dimensional conditions. [ABSTRACT FROM AUTHOR]

Cutting-edge research introduces a chemical reactor for liquid radioactive waste treatment, utilizing a four-stage radiochemical separation. Biomass conversion yielded activated carbons with impressive adsorption capacities up to 1 × 104 Bq/g and proved effective for Cs removal. Standout coagulants, AlCl3·6H2O and FeSO4·7H2O, achieved an 80% decontamination factor for six radionuclides. Sulphonated DVB ion-exchange resins showed high efficiency, achieving 100% Cs decontamination at pH 4 ± 2. Integrated processes successfully decontaminated twelve radionuclides up to 99 ± 1% with substantial demineralization of waste, up to 95%, was achieved. Current reactor system is invaluable for routine and emergency scenarios for treating liquid radioactive waste with varying radionuclide compositions. [ABSTRACT FROM AUTHOR]

Am Fall der Entsorgung radioaktiver Abfälle analysiert der Beitrag, wie Gesellschaften Wissen über ihre Tiefengründe produzieren, um langfristig stabile Entsorgungsszenarien zu entwickeln. Hierzu wird untersucht, wie geologische Formationen in ein solches Entsorgungsvorhaben eingebunden bzw. gar als handlungstragend adressiert werden. Im Zentrum der Untersuchung stehen soziogeologische Relationen im Feld der naturwissenschaftlichen Endlagerforschung. Aus einer soziomateriell orientierten ethnografischen Perspektive wird gezeigt, wie Wirtsgesteinen stabilisierende Fähigkeiten aber auch Anfälligkeiten mit Blick auf Endlagervorhaben zugeschrieben werden. Zudem wird das Experimentieren mit ‚endlagerrelevanten' Materialien in Laboratorien fokussiert. Ziel des Beitrags ist die Formulierung einer Soziologie der Entsorgung im Gefüge von Natur, Technik und Gesellschaft. Based on the case of radioactive waste management, the article analyzes how societies produce knowledge about their undergrounds in order to develop long-term stable disposal scenarios. For this purpose, it is investigated how geological formations are integrated into such disposal projects or even addressed as action-bearing. The focus is on sociogeological relations in the field of scientific repository research. From a sociomaterially oriented ethnographic perspective, the paper discusses how host rocks are attributed stabilizing abilities but also vulnerabilities with regard to repository projects. In addition, experimentation with 'repository-relevant' materials in laboratories will be scrutinised. The aim of the contribution is to formulate a sociology of disposal in an assembly of nature, technology and society. [ABSTRACT FROM AUTHOR]

Energy transition is one of the basic actions taken to counteract and prevent climate change. The basic assumption of energy-related changes is its sustainable use according to the closed-loop model, as well as moving away from fossil fuels, in particular from coal, the combustion of which contributes to excessive harmful carbon dioxide emissions. One of the most popular solutions towards green energy is nuclear energy. Its use allows for a significant reduction in greenhouse gas emissions harmful to the environment and climate, but it also involves the generation of radioactive waste that requires appropriate processing. This paper presents the results of the flotation removal of barium(II) ions from a dilute aqueous solution using ionized acyclic polyethers. The basic factors determining the efficiency and kinetics of the process were defined. It has been shown that as the acidity of the attached polyether molecules increases: the flotation rate constant 1 (0.1667 min−1) < 3 (0.2468 min−1) < 2 (0.3616 min−1) and the separation degree Ba2+: 1 (86.8%) < 3 (99.3%) < 2 (99.4%). The presented results of ion flotation tests may facilitate the collective or selective separation of radioactive isotopes, i.e., Cs-137, Sr-90, Ba-133 and Co-60, from radioactive wastewater in the future. The results of the experimental work described in the article can also be used to develop individual processes for separating mixtures of radioactive isotopes (radioactive wastewater) into individual components (isotopes) and subjecting them to subsequent transformation processes. The obtained results allow us to claim that the tested organic compounds can be used in the future in the selective treatment of hazardous wastewater, which will translate into a reduction in unit costs of industrial processes. The selective recovery of individual pollutants is the basis for the next step in waste management, i.e., designing a cheap method of waste disposal, which also directly affects the economics of the process and its use in industrial conditions. [ABSTRACT FROM AUTHOR]

This article presents the first study on the use of alginate and moringa as an effective adsorbent for uranium treatment under various conditions. A new method of preparing adsorbents was used by adding Moringa oleifera seeds (MOS) to alginate through simple grafting. The resulting beads have a heterogeneous, sponge-like structure that is reversible. The adsorbents were tested for their ability to remove uranium ions from water and showed excellent efficiency with an adsorption capacity of 443.60 mg/g for alginate-PEI-MOS and 341.38 mg/g for alginate-Glu-MOS. Additionally, these adsorbents showed an interesting rate of swelling in water and uranium solution. The results of isotherm, and kinetic models indicate that the adsorption is monolayer, followed by multilayer adsorption after a certain concentration range with mixed adsorption (physical and chemical). Therefore, MOS could be a useful tool for adsorbing uranium from water and real wastewater. [ABSTRACT FROM AUTHOR]

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Background: Nuclear medicine is an integral and developing field in diagnosing and treating diseases. Monitoring individuals' protection and radiation contamination in the workplace is vital for preserving working environments. Objective: This study aimed to monitor the nuclear medicine department's personnel, environment, and wastes to determine the level of occupational radiation and environmental pollution in Bushehr's nuclear medicine department. Material and Methods: In this cross-sectional study, the initial activity of each radioisotope, radiopharmaceutical, and radioactive waste was measured using a "well counter" daily for three months. Three irradiators' absorbed doses were measured using a direct reading dosimeter. The contamination was determined using an indirect wipe test method on various surfaces. A Geiger Müller dosimeter was employed to examine personnel's hands, clothing, and footwear. Results: The highest activity was observed in technetium waste (1118.31 mCi). Every irradiator received a lower absorption dose than the International Commission on Radiological Protection (ICRP) standard threshold. The majority of contamination was associated with the exercise test room (0.04 Bq/cm²) and its work surface (0.013 Bq/cm²), which were both below the threshold (0.5 Bq/cm²). Staff monitoring indicated that two nurses (10 and 11 individuals) had the highest contamination rate (23.7%). Conclusion: Daily assessment of the type, activity, and method of radiopharmaceutical administration to the patient is advantageous for waste management. Surface contamination monitoring can significantly contribute to the estimation of the level of radiation pollution in the environment. [ABSTRACT FROM AUTHOR]

The incorporation of crown ether into metal‐organic frameworks (MOFs) is garnered significant attention because these macrocyclic units can fine‐tune the inherent properties of the frameworks. However, the synthesis of flexible crown ethers with precise structures as the fundamental building blocks of crystalline MOFs remains a challenging endeavor, with only a limited number of transition metal examples existing to date. Herein, 18‐crown‐6 and 24‐crown‐8 struts are successfully incorporated into the skeleton of zirconium‐based MOFs to obtain two new and stable crown ether‐based MOFs, denoted as ZJU‐X100 and ZJU‐X102. These newly developed MOFs displayed high porosity and remarkable stability when exposed to various solvents, boiling water, pH values, and even concentrated HCl conditions. Thanks to their highly ordered porous structure and high‐density embedding of specific binding sites within tubular channels, these two MOFs exhibited extremely fast sorption kinetics and demonstrated outstanding performance in the uptake of strontium and cesium ions, respectively. Furthermore, the structures of Sr‐adsorbed ZJU‐X100 and Cs‐adsorbed ZJU‐X102 are solved and confirmed the precise location of Sr2+/Cs+ in the cavity of 18‐crown‐6/24‐crown‐8. This makes modular mosaic of different crown ethers into the skeleton of stable zirconium‐based MOFs possible and promote such materials have broad applications in sorption, sensing, and catalysis. [ABSTRACT FROM AUTHOR]

Materials based on Portland cement and concrete are widely used for radioactive waste solidification and construction of engineered safety barriers at radioactive waste repositories. As compared to other candidate matrices, cement also exhibits the potential for chemical immobilization by reacting with many radionuclides and binding them. The water permeation through the bulk of concrete leads to its degradation, to a change in the composition of its pore water, and to transformations of solid mineral phases constituting the concrete. The long-term degradation of concrete can be estimated using geochemical modeling. A procedure for forecasting the degradation of materials based on Portland cement concrete is described. The model takes into account the stages of cement hydration and chemical degradation. The geochemical changes in the course of concrete degradation were calculated using numerical modeling with the PhreeqC code. Schemes of the stepwise concrete degradation are considered for estimating the radionuclide transport. Relevant radionuclide distribution coefficients are applied to each stage. The duration of concrete degradation stages largely depends on the filtering water composition. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

A purification method is developed to determine 90Sr in radioactive waste. A separation based on Sr-resin® with a pretreatment using TRU-resin® provides satisfactory Sr recovery yields before the 90Sr measurement by liquid scintillation counting (LSC) regardless the analyzed samples with low or high activity level. The selectivity of the procedure is checked by measuring the 90Y ingrowth after different days of separation without waiting for secular equilibrium. In order to obtain a REACH compliant method without scintillation cocktails, a plastic scintillation resin selective for Sr is implemented on the basis of the developed protocol. The optimized method is applied successfully to representative nuclear waste including samples with high Pu content (effluents, concretes and sludges). [ABSTRACT FROM AUTHOR]

The types and contents of minerals contained in natural zeolites vary depending on the production conditions. As a result, the adsorption properties of natural zeolites also vary depending on the production area. To utilize natural zeolites as adsorbents in various fields, it is crucial to identify the adsorption properties based on the type and content of minerals they contain. Therefore, this study aimed to analyze the adsorption characteristics of Cs and Sr using six types of natural zeolites from Mongolia (M-1, M-2, M-3, M-4, M-5, and M-6) to determine their potential as radioactive waste treatment agents. The experiments conducted showed that the adsorption equilibrium for Cs and Sr was reached within 4 and 8 h., respectively. Among the six samples, M-2 and M-6 exhibited relatively high adsorption rates, with Cs adsorption rates of 80% and 84%, and Sr adsorption rates of approximately 18% and 21%. Furthermore, the single-component isothermal adsorption experiments demonstrated that all six samples followed the Langmuir model. The maximum adsorptions of Cs was highest in M-2 (154.7 mg/g), M-4 (148.8 mg/g), and M-6 (150.2 mg/g) respectively. Additionally, the maximum adsorptions of Sr for M-2 and M-6 were 29.66 and 33.67 mg/g, respectively. It was observed that the natural zeolite with a higher content of 8-ring in the framework and higher CEC exhibited greater selectivity for Cs compared to the other six natural zeolites. Therefore, the maximum adsorption of Cs was attributed to the high total content of 8-rings in the natural zeolite framework, as well as the cation exchange capacity (CEC). [ABSTRACT FROM AUTHOR]

Abstract Rock salt is considered a suitable medium for the permanent disposal of heat‐generating radioactive waste due to its isolation properties. However, excavation damage and heating induce complex and heterogeneous thermal‐hydrological‐mechanical (THM) processes across different zones. Quantifying this heterogeneity is crucial for accurate long‐term performance assessment models, but traditional methods lack the necessary resolution. This study employs 4D electrical resistivity tomography (ERT) monitoring during controlled heating experiments in a salt formation to unravel the spatiotemporal dynamics of THM processes. Advanced time‐lapse inversion and clustering analysis quantify subsurface properties and map the heterogeneity of THM dynamics. The ERT results can estimate subsurface properties and delineate the damaged and intact zones, enabling appropriate parameterization and representation of processes for long‐term modeling. This approach may be used in further improving the predictive models and ensuring the safe long‐term disposal of radioactive waste in rock salt.

Performance assessment in deep geological nuclear waste repository systems necessitates an extended knowledge of the pore water chemical conditions prevailing in host-rock formations. In the last two decades, important progress has been made in the experimental characterization and thermodynamic modeling of pore water speciation, but the influence of experimental artifacts and uncertainties of thermodynamic input parameters are seldom evaluated. In this respect, we conducted an uncertainty propagation study in a reference geochemical model describing the pore water chemistry of the Callovian-Oxfordian clay formation. Nineteen model input parameters were perturbed, including those associated to experimental characterization (leached anions, exchanged cations, cation exchange selectivity coefficients) and those associated to generic thermodynamic databases (solubilities). A set of 13 quantities of interest were studied by the use of polynomial chaos expansions built non-intrusively with a least-squares forward stepwise regression approach. Training and validation sets of simulations were carried out using the geochemical speciation code PHREEQC. The statistical results explored the marginal distribution of each quantity of interest, their bivariate correlations as well as their global sensitivity indices. The influence of the assumed distributions for input parameters uncertainties was evaluated by considering two parametric domain sizes.

This report summarizes the findings of a workshop held at the safeND Research Symposium and hosted by the German Federal Office for the Safety of Radioactive Waste Management (BASE) in Berlin in September 2023. The workshop aimed to channel perspectives from various fields of expertise to discuss key sustainability concepts in terms of radioactive waste management. Therefore, the report highlights that current sustainability concepts, such as the United Nations' Sustainable Development Goals (SDG) as well as the concept of Planetary Boundaries, neglect challenges arising from the production and storage of human-made radioactive materials. The workshop consisted of three group tasks. The first attempted at identifying the interrelations between "sustainability" and radioactive waste management. The second was to map the global nature of the challenges. The third took first steps to determine a human-made radioactive material as a potential planetary sub-boundary for "novel entities". All three groups identified valuable knowledge gaps that should be addressed by future research and concluded that radioactive waste management is underrepresented in these sustainability concepts. [ABSTRACT FROM AUTHOR]

Hundreds of thousands of tons of waste are generated from decommissioned nuclear- power facilities, and it has become a critical global issue to secure technology for reducing and recycling this waste. Concrete waste (CW) is estimated to comprise 60–80% of the total waste, and concrete-waste powder (CWP) includes enough inorganic substances used as effective materials for waste treatment. Accordingly, it can be used to produce recycled cement (RC). This study aimed to evaluate the performance of a solidification agent manufactured using recycled cement (SRC) for the safe packing of radioactive wastes, such as coarse aggregates of CW, waste soil, and metal wastes originating from decommissioned nuclear facilities. The experimental results indicated that the most relevant incineration temperature of CWP for RC was 700 °C. The optimum water-to-binder ratio was determined to be 0.4, and the most relevant substitution ratio of ground granulated blast furnace slag for CWP was determined to be 15%. In addition, calcium silicate hydrate is the most effective hydration product for improving the compressive strength of SRC. The maximum packing capacities of the SRC for coarse aggregates, waste soil, and metal waste, which were simulated as radioactive wastes, were determined to be 30, 5, and 7 wt%, respectively. The results of leaching tests using SRC containing radioactive wastes contaminated with Co, Cs, and Sr indicated that their leachability indices met the acceptance level for disposal. Consequently, the RC composed of CWP can be used as a solidifying agent to safely dispose of radioactive wastes, such as coarse aggregates, waste soil, and metal waste. [ABSTRACT FROM AUTHOR]