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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]

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]

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.

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]

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]

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.

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]

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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]

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.

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]

Sr2+ ions in an aqueous solution were removed using Na-A zeolites synthesized from kaolin, a natural mineral. Na-A zeolites with high crystallinity were synthesized using NaOH/kaolin mass ratios of 0.6 (ZK06) and 0.9 (ZK09). The adsorption reached equilibrium within 120 min. The adsorption data obtained from experiments for Sr2+ using ZK06 and ZK09 were appropriately analyzed with pseudo-second-order kinetic and Langmuir isotherm models. Comparing the maximum adsorption capacities (qm) of ZK06 and ZK09 for Sr2+, the highest values were obtained at 1.90 and 2.42 mmol/g, respectively. Consequently, the Na-A zeolites synthesized from kaolin can be evaluated as adsorbents with high adsorption capacities for the removal of Sr2+, proportional to the degree of their crystallinity. [ABSTRACT FROM AUTHOR]

Biosorption is an impurity-free application developed from the use of nuclear technology for peaceful purposes in everyday life and can be used to treat wastewater streams contaminated with various radionuclides. In this study, a laboratory decontamination experimental approach was developed to apply commercial chitosan as a biosorbent applied for removing radiocesium (Cs-137) and/or radiocobalt (Co-60) from spiked aqueous media. The factors assumed to affect the biosorption of both radionuclides included contact time, pH, and initial radioactivity content. In addition, the biosorbent dose and temperature of the process were studied. Both the biosorption capacity and the biosorption efficiency of the treatment process were calculated. According to FT-IR analysis, it can be assumed that the chitosan amine group (-NH2) is almost accountable for the biosorption of both radionuclides from waste solution simulates. Based on the data obtained, commercial chitosan can be considered an economical and efficient biosorbent for handling low- and medium-level radioactive wastewater streams containing cesium and/or cobalt radionuclides. The acquired data showed that 144 h is an adequate time to remove more than 94% of radiocobalt and about 93% of radiocesium, from a separate solution for each, at pH ~6.5 and using 0.5 g of commercial chitosan. [ABSTRACT FROM AUTHOR]

In this interview, Cécile Massart looks back over her career. She discusses her beginnings. She started with printmaking, moved on to computers and made her consider the use of other media. The discovery of radioactivity issues constituted a "turning point" for her work. She emphasises upon her use of photography in order to question what the wastes we hide say about us as society and as a legacy? [ABSTRACT FROM AUTHOR]

The use of the fillers wollastonite and volcanic ash for the formulation of magnesium phosphate cements prepared at magnesium-to-phosphate molar ratios of 2, 3 and 4 has been investigated, with the objective of evaluating these formulations for the encapsulation of aluminium radioactive waste. The workability, mechanical strength, dimensional stability, pH, chemical composition and mineralogical properties of cement pastes and mortars were examined. All cement pastes presented fast setting, and the workability was only good at 3 and 4 M. The cement mortars presented high compressive strength and dimensional stability. K-struvite was confirmed as the sole reaction product of the reaction for all formulations. The pH of the cement pastes, measured in suspensions, achieved values in the range of 7.8 to 9.5 after the first days of setting, exceeding pH 8.5 for the 2 and 3 M formulations. pH values below 8.5 are theoretically preferred to avoid potential aluminium corrosion. Both fillers presented adequate characteristics (good workability, chemical compatibility) to be used in the formulation of magnesium phosphate cements. The increasing magnesium-to-phosphate molar ratio prevented unwanted efflorescence and increased the mechanical stability of the cement. [ABSTRACT FROM AUTHOR]

The potential for microbial activity to occur within the engineered barrier system (EBS) of a geological disposal facility (GDF) for radioactive waste is acknowledged by waste management organizations as it could affect many aspects of the safety functions of a GDF. Microorganisms within an EBS will be exposed to changing temperature, pH, radiation, salinity, saturation, and availability of nutrient and energy sources, which can limit microbial survival and activity. Some of the limiting conditions are incorporated into GDF designs for safety reasons, including the high pH of cementitious repositories, the limited pore space of bentonite-based repositories, or the high salinity of GDFs in evaporitic geologies. Other environmental conditions such as elevated radiation, temperature, and desiccation, arise as a result of the presence of high heat generating waste (HHGW). Here, we present a comprehensive review of how environmental conditions in the EBS may limit microbial activity, covering HHGW and lower heat generating waste (LHGW) in a range of geological environments. We present data from the literature on the currently recognized limits to life for each of the environmental conditions described above, and nutrient availability to establish the potential for life in these environments. Using examples where each variable has been modelled for a particular GDF, we outline the times and locations when that variable can be expected to limit microbial activity. Finally, we show how this information for multiple variables can be used to improve our understanding of the potential for microbial activity to occur within the EBS of a GDF and, more broadly, to understand microbial life in changing environments exposed to multiple extreme conditions. [ABSTRACT FROM AUTHOR]

Here it is attempted to give a brief introduction to the work of Ruggero San- tilli via a slightly more detailed discussion of two areas in which he has made major contributions. The motivation for his work is also discussed through- out and an attempt is made also to highlight some of the problems which have inspired him over the years. The two examples taken are, firstly, a fas- cination with the early work of Rutherford in viewing a neutron, which was undiscovered at the time, as a proton and an electron. This model was dis- carded because it wasn’t allowed by conventional quantum mechanics. The second was a deep interest in the paradox introduced into physics by the ap- pearance of the now famous article by Einstein, Podolsky and Rosen in 1935. Both these contributed to Santilli’s realization that, for further progress to be made in physics, a new approach had to be adopted which would involve the use of as yet unused, or unknown, mathematical methods. Over the years, a huge body of new mathematics has emerged through the efforts of Santilli, helped to some extent by various other workers, but the main body has been due to him. This relatively short foreword has as its aim to promote this lifetime’s work of Santilli and hopefully encourage others to be motivated to study it with truly open minds and, where necessary, carry out actual ex- periments to test independently some of the predictions made – especially the one relating to a possible method for the safe, rapid disposal of nuclear waste.

Abstract 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.

The relevance of the conducted research is determined by the need to control and forecast the costs of enterprises to maintain shutdown nuclear energy facilities in a save condition in order to solve budget planning problems. In the near future, an increase in operating expenses is expected due to an increase in the number of such facilities. To solve the problem econometric forecasting methods are used.Purpose of the research. This study is aimed to develop a methodology for estimating the costs of maintaining nuclear and radiation hazardous facilities in a safe condition and the safe storage of radioactive waste, considering the specifics of accounting at nuclear industry enterprises and forecasting such costs, taking into account plans for shutting down and decommissioning of facilities and removing accumulated radioactive waste from storage facilities in the future.Materials and methods. In the research there was developed the methodology of isolation from the overall structure costs, associated with maintaining shutdown facilities and storage facilities for radioactive waste of nuclear industry enterprises. A cumulative estimation of such costs has been carried out for facilities whose operation for the intended purpose has currently been discontinued. Based on obtained retrospective data, a correlation-regression model with structural changes was developed to predict the costs of maintaining nuclear energy facilities in a safe condition which are expected to be shut down in the future. The developed model allows considering the specifics of facilities, their level of radiation hazard and overall characteristics, the average level of wages at enterprises and the cost of resources.Results. Based on the plans for the shutdown and decommissioning of nuclear energy facilities and the disposal of radioactive waste from storage facilities, the costs of maintaining such facilities in a safe condition with a planning horizon up to 2050 have been forecasted. The results of calculations show that the costs for nuclear energy facilities (except for nuclear power plant units) will increase to 10 billion rubles per year by 2035 (in 2022 prices) and will significantly decrease after 2035 - up to 7.5 billion rubles per year. The reduction will be due to the transfer to a nuclear and radiation safe state of large facilities of radiochemical plants, as well as the conservation of sites for the placement of industrial uranium-graphite reactors. The costs of maintaining the storage facilities of accumulated radioactive waste in a safe condition will be reduced annually (almost linearly) as they are transferred to the disposal of the removed radioactive waste and the conservation of special radioactive waste facilities. By 2035, costs will be reduced from the current level of 5 billion rubles to 4.3 billion rubles per year.Conclusion. The results of the study confirm the high importance of the implementation of the Federal Target Program “Ensuring Nuclear and Radiation Safety for 2016-2020 and for the period up to 2035”, under which it is planned to decommission about 75 stopped nuclear and radiation hazardous facilities and transfer about 155 thousand cubic meters of radioactive waste to disposal.

The Republic of Croatia has announced a plan to build a radioactive waste repository in Mount Trgovska Gora located along the border area with Bosnia and Herzegovina (BiH), which has upset residents living in the border municipalities in BiH. This paper seeks to examine local residents’ opinion regarding the construction of a radioactive waste repository. Our study was conducted in Novi Grad, a place located in the immediate vicinity of Trgovska Gora. The aim of the study was to examine local residents’ perceptions and the implications of the planned project for human security. This study is based on a qualitative research approach using the focus group interview as a technique. Eleven respondents participated in the focus group interview. The group consisted of representatives of local communities, residents, and representatives of non-governmental organizations. Our findings indicate that local residents strongly disagree with the announced project and numerous negative implications for human security.

The distribution of tritium in the soil-vegetation complex was studied on a series of combined samples taken in the area of the radioactive waste storage facility (RWSF). It is shown that bioaccumulation of tritium by vegetation is determined mainly by the position of each specific object (testing point) in relation to the predominant direction of atmospheric transport of the nuclide. In locally selected samples of herbaceous and arboreal vegetation, the content of the nuclide changes according with a high degree of correlation, and its distribution in fractions of freely migrating (transpirational) tritiated water (TW) and in the form of organically bound tritium (OBT) also changes according with a high degree of correlation. In the zones of intense and persistent atmospheric migration tritium forms radiogeochemical anomalies in the soil and vegetation. Under the conditions of stationary atmospheric migration flows, pseudoequilibrium ratios of nuclide concentration in the components of the soil-vegetation system are established, while under the conditions of the functioning of short-term soil flows, the concentration equilibrium of the nuclide in the components of the soil-vegetation system is not observed. Higher concentrations of tritium are observed in more tightly bound forms in both soil and vegetation, which obviously reflects a large dynamic dependence of the mobile form on weather and climate conditions. The ratios of activity concentration of tritium in different forms of presence determined for different types of vegetation as concentration ratio (CR) indicate that tritium accumulates more effectively in arboreal plants - bushes, trees than in herbaceous ones (grass), it is equal to 1.10 and 1.20, respectively. Freely migrating (transpirational) tritiated water possibly due to greater mobility and greater dependence on changes in the activity concentration of tritium in the atmospheric plume and moisture saturation of the air, has a lower activity concentration of tritium than the organically fixed form (OBT), respectively 1.14 and 1.29, which does not have such temporal dependence. During the emission of tritium from RWSF there was a significant accumulation of the heavy isotope of hydrogen in an organically bound form, that is more than half of the tritium absorbed by vegetation is retained during the existence of organic matter and is temporally excluded from the biological cycle in the case of arboreal vegetation. In conditions of the forest landscape of the sanitary protection zone (SPZ) of the RWSF, root absorption of tritium by grass from the humus layer is from 20 to 31% of its total amount accumulated in the soil. The degree of absorption of tritium from the soil is almost the same in herbaceous and arboreal vegetation. This indicates that during the long-term existence of the atmospheric-radio-geochemical anomaly caused by the constant emission of tritium from the storage of RW in organic matter, an equilibrium balance of the content of transpirational HTO and OBT has been formed. In turn, the establishment of such a balance in plant matter can be used as an indicator of a prolonged regime of tritium emission from RWSF. Herbaceous and arboreal vegetation make it possible to determine the spatial position of atmogeomigratory paths of tritium distribution. Plant indicators indirectly reflect the impact of this migration flow on the direct absorption of tritium by organic matter from the air, or its absorption from the root layer of the soil, where the heavy isotope of hydrogen enters with atmospheric precipitation.

Radioactive waste poses significant risks to human health and the environment due to the emission of gamma rays, beta particles, and alpha particles from natural or artificial radioactive elements. This paper reviews the essential sources of radioactive waste, their potential risks to humans and the environment, their classification depending on their chemical, physical, and radiological features, and the treatment techniques utilized to prevent them from leaking into the environment. In addition, the current study focuses on using membranes exclusively and membrane technologies in the remediation of radioactive liquid wastes. The use of pressure-driven membrane technologies, such as reverse osmosis, nanofiltration, ultrafiltration, and microfiltration, is particularly emphasized. Additionally, the use of nanomaterials like titanium dioxide, zinc oxide, silica, alumina, silver oxide, zeolite, copper ferrocyanides, and carbon nanotubes embedded with membranes to improve their effectiveness and enhance their applicability decreases the risks associated with radioactive contamination via modifying membrane properties including flow rate, the result of rejecting radioactive particles dissolved or suspended in contaminated water, hydrophobic and hydrophilic membrane's ability and among other properties. Moreover, incorporating nanoparticles into reinforced membranes enhances the mechanics of nanocomposite membrane surfaces via processes like adsorption and ion exchange with various radioactive ions. Besides increasing nanocomposite membranes' performance in reducing the hazardous radionuclides effects, it modifies membrane properties, such as enhancing antibacterial capabilities, antifouling, mechanical stability, and thermal stability. The article review also discusses hybrid pressure-driven membrane processes in disposing of radionuclides generated in nuclear power plants, hospitals, research facilities, and decontamination projects to protect human health and the environment.

We developed a miniaturized multi-dimensional radiation sensor system consisting of an inorganic scintillator array and plastic optical fibers. This system can be applied to remotely obtain the radioactivity distribution and identify the radionuclides in radioactive waste by utilizing a scanning method. Variation in scintillation light was measured in two-dimensional regions of interest and then converted into radioactivity distribution images. Outliers present in the images were removed by using a digital filter to make the hot spot location more accurate and cubic interpolation was applied to make the images smoother and clearer. Next, gamma-ray spectroscopy was performed to identify the radionuclides, and three-dimensional volume scanning was also performed to effectively find the hot spot using the proposed array sensor.

Abstract Biomedical waste management is an integral part of health care as being the mainstay of hospital cleanliness, hygiene, and maintenance activities. Medical care is crucial for human health but mismanagement of biomedical waste harms the flora and fauna of ecosystem on earth, and badly influences the human health. Due to their stability, biomedical wastes are not eliminated by the solid or sewage treatment plants, and get bioaccumulated in environment, and eventually get to population through various means. They are harmful due to their potential carcinogenicity, genotoxicity, mutagenicity, and other toxicities. Current disposal techniques employed in disposing biomedical wastes are sewage/drains, incineration, and landfills. These practices are normally costly and might transform pollutants from one toxic form to another. Whereas bioremediation is an inexpensive method of employing naturally occurring microorganisms or plants to detoxify man-made pollutants to harmless products that make soil fertile as well in the process. Researchers have also evolved genetically engineered microbes to remediate environmental pollutants including radionuclides. Phytoremediation is also a type of sustainable methods, which is reasonable and competent in handling heavy metals and radioactive waste generated from hospitals. In this article, we have summarized rational applications of bioremediation techniques of using effective microorganisms and plants in enhanced removal of several recalcitrant pollutants including biomedical waste. Our review highlights the challenges and future perspectives to bioremediate non-biodegradable and potentially toxic hospital wastes.

The safety and robustness of Deep Geological Repositories (DGRs) are of paramount importance for the long-term management of spent nuclear fuel from electricity generation. The introduction of a multi-barrier system, which includes the host rock formation and an engineered barrier system (including the bentonite buffer), has been a widely used approach to ensure the safety of DGRs. The assessment of the long-term safety of DGRs involves the mathematical modeling of the coupled thermal–hydraulic–mechanical–chemical (THMC) processes that occur in the near-field of the DGRs and their impact on the behaviour and engineering properties of the bentonite buffer. This paper presents a review of the THMC-coupled processes that arise in the bentonite buffer as well as a mathematical model governing such coupled processes. The model is verified against existing analytical solutions and validated against measured data of a thermal diffusion experiment in a sand bentonite column. Also, scoping analyses were performed to assess the influence of coupled THM processes on solute transport in clayrocks. The results of the numerical model closely matched those of the analytical solutions and experimental data demonstrating the capability of the provided mathematical model as well as the numerical approach in enhancing our comprehension of DGR behaviour. This enhanced comprehension will be valuable for safety prediction and assessment in the context of DGRs. The work presented in this paper is part of the Canadian Nuclear Safety Commission's (CNSC) regulatory research to gain independent knowledge on the safety of the geological disposal of radioactive waste. [ABSTRACT FROM AUTHOR]

Conditioning of radioactive waste generated from the operation of medical institutions, nuclear cycle facilities, and nuclear facilities is important for the safety of the environment. One of the most hazardous radionuclides is radioactive cesium. There is a need for more effective solutions to contain radionuclides, especially cesium (Cs+). Geopolymers are promising inorganic materials that can provide a large active surface area with adjustable porosity and binding capacity. The existence of nanosized zeolite-like structures in aluminosilicate gels was shown earlier. These structures are candidates for immobilizing radioactive cesium (Cs+). However, the mechanisms of their interactions with the aluminosilicate framework related to radionuclide immobilization have not been well studied. In this work, the influence of alkaline cations (Na+ or K+) and the aluminosilicate framework structure on the binding capacity and mechanism of interaction of geopolymers with Cs+ is explored in the example of a sodalite framework. The local structure of the water molecules and alkaline ions in the equilibrium state and its behavior when the Si/Al ratio was changed were studied by DFT. [ABSTRACT FROM AUTHOR]

A novel solid-phase synthetic approach was developed to produce a mineral-like composite ceramic based on strontium titanate (SrTiO3) and yttrium titanate (Y2Ti2O7) matrices for immobilizing radionuclides such as 90Sr and its daughter product 90Y, as well as lanthanides and actinides, via reactive spark plasma sintering technology (SPS-RS). Using XRD, SEM, and EDS analyses, the sintering kinetics of the initial mixed oxide reactants of composition YxSr1–1.5xTiO3 (x = 0.2, 0.4, 0.6 and 1) and structure-phase changes in the ceramics under SPS-RS conditions were investigated as a function of Y3+ content. In addition, a detailed study of phase transformation kinetics over time as a function of the heating temperature of the initial components (SrCO3, TiO2, and Y2O3) was conducted via in situ synchrotron XRD heating experiments. The composite ceramic achieved relatively high physicomechanical properties, including relative density between 4.92–4.64 g/cm3, Vickers microhardness of 500–800 HV, and compressive strength ranging from 95.5–272.4 MPa. An evaluation of hydrolytic stability and leaching rates of Sr2+ and Y3+ from the matrices was performed, demonstrating rates did not exceed 10−5–10−6 g·cm−2·day−1 in compliance with GOST R 50926-96 and ANSI/ANS 16.1 standards. The leaching mechanism of these components was studied, including the calculation of solution penetration depth in the ceramic bulk and ion diffusion coefficients in the solution. These findings show great promise for radioactive waste conditioning technologies and the manufacturing of radioisotope products. [ABSTRACT FROM AUTHOR]

Highly radioactive materials classified as high-level nuclear waste (HLW) of atomic power engineering should be disposed of deeply underground in special geological disposal facilities (GDFs), which can be of either shaft or borehole type. The advantages of borehole-type GDFs result from smaller volumes of mining operations, a simpler construction technology, shorter construction time and cost. This allows us to consider them as an alternative to shaft-type GDFs. The parts of the boreholes in which waste containers should be placed can be both vertical and horizontal. Computer simulation of the migration of radionuclides from a group of parallel horizontal boreholes into the biosphere made it possible to conclude that horizontal GDF boreholes have significant advantages over vertical ones. We determined a forecast of 241Am migration by a method of mathematical modelling of 241Am release from vitrified HLW disposed of in several horizontal drillholes. The maximum concentrations of americium in the near-surface groundwater above the repository are calculated depending on the number of boreholes, the depth of their location and the distance between them, the permeability of rocks and the time of waste storage prior to disposal. Influence of different conditions on the safety of a GDF of borehole type is estimated. Calculations show that the heat generated by HLW causes a weaker groundwater convection near horizontal boreholes compared to vertical boreholes of the same capacity. In addition to that, at an equal thickness of the rock layer separating the HLW from the surface, the geothermal temperature of the host rocks in the near field of a horizontal borehole will be lower than the average geothermal temperature near a vertical borehole. As a result, the rate of radionuclides leaching from the waste forms by groundwaters will also be lower in the case of horizontal boreholes. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

A 3D numerical simulation of radionuclide migration from a potential repository of solid radioactive waste is carried out. The federal site is located in the Krasnoyarsk region (Siberia, Russia), at a distance of a few kilometers from the eastern bank of the river Yenisei. The models take into account the actual topography of the area and the heterogeneity of the hydraulic conductivity of the enclosing rocks. The migration of long-lived radionuclides is considered, e.g. 241Am (half-life of 432.6 years). The simulated time intervals range from the waste disposal closure to the fault emergence. The model settings consider different distances between the fault and the repository. The highly permeable near-vertical faults run either from South to North (meridional) or from West to East (latitudinal). The simulation results show that the meridional faults do not exert a substantial influence on contaminant plume movement. By contrast, the latitudinal faults can represent a significant ecological hazard depending on the distance between the repository and the fault. While this study is not an assessment of the safety of the potential repository site, the results point out that further investigations of the stress state of the enclosing rocks are needed to better constrain potential faults and flow directions. [ABSTRACT FROM AUTHOR]

Nuclear power generation, which began in the Soviet Union in June 1954, has now become a main source of electrical energy in 33 countries around the world. However, radioactive waste and its safety have always been at the center of controversy, and even after 70 years, there is no official regulation on the disposal of radioactive waste under international law. In this review, the policies and current status of radioactive waste disposal in major countries around the world, including Korea, are investigated. Based on this, the direction for radioactive waste management in Korea was presented. Currently, radioactive waste is disposed of in two ways. Finland, Sweden, and France are conducting deep permanent disposal according to the recommendation of the International Atomic Energy Agency, and Japan, the United Kingdom, and India are promoting reprocessing of spent nuclear fuel to reduce waste and the volume of the repository. Korea has been temporarily storing radioactive waste at the site of the power plant since the completion of Kori Unit 1 in 1978, but it is expected that the capacity of the temporary storage facility will reach its limit in 2031. Accordingly, the securing of a permanent disposal site and the development of reprocessing technology are being studied together, but no tangible results have been achieved so far. Korea's radioactive waste disposal has problems such as a small land area, institutional difficulties in developing reprocessing technology, frequent policy changes, ignorance and indifference, and lack of smooth communication. Although public concern about nuclear power generation has increased due to the Fukushima disaster, not many people are interested in spent nuclear fuel. Within the framework of reprocessing and permanent disposal, Korea should refrain from frequent policy changes that can confuse public opinion and research and development, and come up with a long-term realistic policy.

The requirements of NTD (Neglected Tropical Diseases) and technological regulations for the operation of NPP (Nuclear Power Plant) power units (NP-001-97 (OPB-88/97), NP-082-07) define the requirements for monitoring the specific activity of iodine-131 (the amount of iodine) in the NPP primary circuit coolants. The advantages of laboratory control include accuracy of measurement and the radionuclide composition of the primary coolant, measured using high-precision laboratory equipment. Instrumental spectra were obtained for the detection units BDKG-205m with various options for the placement of waste in a container, their composition, mass of waste, average density, and various activity levels of waste. The basic idea behind gamma-ray spectroscopy is to detect and analyze the energy of incident gamma rays. Gamma rays of varying energy and intensity are emitted from radioactive sources. The gamma-ray energy spectrum is produced when gamma rays are detected and examined using a spectroscopy instrument. The initial stage in gamma-ray spectroscopy is to detect gamma rays using a suitable detector. The detector captures and measures the energy of incoming gamma rays. Scintillation detectors, semiconductor detectors, and gas-filled detectors are among the detectors used in gamma-ray spectroscopy. The incoming gamma-ray energy is converted into electrical signals that can be processed and studied by these detectors. The spectroscopic system measures and records the energy of gamma rays when they are detected. The derived energy spectrum depicts the intensity distribution of gamma rays as a function of energy. The spectrum is a visual representation of the different energy levels found in gamma-ray emission.

In this study, ethylene propylene diene monomer (EPDM)-lead monoxide (PbO)-magnetite (Fe3O4) composite was prepared and changes in its properties with different gamma irradiation doses (50, 100, 200, 300, 500, and 1000 kGy) were investigated. Characterizations of irradiated EPDM-metal oxide composites were performed by X-ray Diffractometer (XRD), Field Emission Scanning Electron Microscope (FESEM) and Universal Testing Machine (UTM). It was observed that the synthesized sample shows an amorphous nature beyond the 500 kGy irradiation dose. The FESEM morphographs of irradiated samples show an increase in fracture size with an increase in irradiation dose. The mechanical tests have shown that prepared samples retain their mechanical strength up to ∼100 kGy. In addition, the effect of thermal and electrical properties on irradiation dose has also been studied. The synthesized samples show good thermal stability up to ∼500 kGy. The AC conductivity of the synthesized EPDM-metal oxide composites has been observed to increase slightly with an increase in irradiation dose. It was observed that synthesized EPDM-metal oxide composites are relatively good radiation resistant candidates and can be used for radioactive waste disposal management applications.