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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.

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.

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.

The release of radionuclide elements from nuclear, medical, research, and defense fields has become a serious environmental problem. Hence, the investigation of the adsorption kinetic and isotherm of metal ions (Co2+, Sr2+, and Cs+) which are commonly found in radioactive waste has been considered essential for radioactive contaminant removal process from aqueous environment. In this paper, the adsorption kinetic of three metal ions on Fe3O4 nanoparticle (Fe3O4 NPs) have been described by the pseudo-first-order, pseudo-second-order, Elovich, and Intra-particle diffusion models, while the adsorption isotherm have been described by the Langmuir, Freundlich, Dubinin-Radushkevich model, and Tempkin isotherm models. Comparing the correction coefficients (R2) of four adsorption kinetic models and four isothermal adsorption models, the adsorption kinetic of three ions fitted well to the pseudo-second-order model with R2 > 0.9913. Meanwhile, the adsorption isotherm of three metal ions indicate that the Langmuir model was best fit the isothermal adsorptions with R2 > 0.9799. The maximum adsorption capacities of Co2+, Sr2+, and Cs+ were found to be 30.488, 34.843, and 33.223 mg/g, respectively. Based on the data of the present investigation, the Fe3O4 NPs being eco-friendly and low-cost adsorbent might be a suitable alternative for the elimination of radioactive waste from environmental and industrial wastes.

Radioactive waste should be solidified before being disposed of in the repository to eliminate liquidity or dispersibility. Cement is a widely used solidifying media for radioactive waste, and cement solidified waste should satisfy the minimum compressive strength of the waste acceptance criteria of a radioactive repository. Although the compressive strength of waste should be measured by the test method provided by the waste acceptance criteria, the method differs depending on the operating repository of different countries. Considering the measured compressive strength changes depending on test conditions, the effect of test conditions should be analyzed to avoid overestimation or underestimation of the compressive strength during disposal. We selected test conditions such as the height-to-diameter ratio, loading rate, and porosity as the main factors affecting the compressive strength of cement solidified radioactive waste. Owing to the large variance in measured compressive strength, the effects of the test conditions were analyzed via statistical analyses using parametric and nonparametric methods. The results showed that the test condition of the lower loading rate, with a height-to-diameter ratio of two, reflected the actual cement content well, while the porosity showed no correlation. The compressive strength assessment method that reflects the large variance of strengths was suggested.

Rice bran (RB) is distinguished by its chemical stability, insoluble in water, and local availability, in addition to the presence of ionizable functional groups which able to bind with the elements. Hence, this material was preferred to be utilized in this study as an operative biosorbent of altered hazardous ions such as Pb2+, Sr2+, and Eu3+ from simulated radioactive waste to attain a safe environment free from the risk of such waste. The kinetic studies illustrated that the sorption process obeys the pseudo-second-order model. From the isotherm studies, Langmuir is the most applicable isotherm model to the experimental data with monolayer sorption capacities of 1.26, 0.76, and 0.33 mmol/g for Sr2+, Eu3+, and Pb2+, respectively. The efficiency of RB was evaluated by applying it to a simulated radioactive waste containing multicomponent to study the interfering ions. When the sorption capacities of RB and other sorbents were compared, it became clear that RB had a significantly higher sorption capacity than most sorbents. The findings exposed that RB is a material with the potential for the treatment of radioactive waste and can be successfully applied in this regard.

In this study, an NF3 plasma etching reaction with a cobalt oxide (Co3O4) films grown on the surface of inorganic compounds using granite was investigated. Experimental results showed that the etching rate can be up to 1.604 μm/min at 380 °C under 150 W of RF power. EDS and XPS analysis showed that main reaction product is CoF2, which is generated by fluorination in NF3 plasma. The etching rate of cobalt oxide films grown on inorganic compounds in this study was affected by surface roughness and etch selectivity. This study demonstrates that the plasma surface decontamination can effectively and efficiently remove contaminated nuclides such as cobalt attached to aggregate in concrete generated when decommissioning of nuclear power plants.

The review discusses various alternative approaches for spent nuclear fuel (SNF) reprocessing in aqueous carbonate media. The main stages, schemes, and methods of the most well-known and well-described processes for reprocessing SNF and some high-level radioactive waste using carbonate systems developed by research groups in Japan, the United States of America, the Republic of Korea, and the Russian Federation described and compared. The main advantages of such methods are outlined compared to the SNF reprocessing in nitric acid media. The levels of development and proximity of the designed processes to the industrial implementation are shown. The main principle achievements, prospects, and routes for the refinement of such methods for the technology of SNF reprocessing and handling of high-level radioactive waste formulated.

Ion exchange resins (IERs) are widely used to remove radioactive contaminants from various commercial nuclear power plant systems. After the completion of their useful life cycle, IERs are removed from the nuclear facility and are known as spent resins. In the past few decades, the development of Fenton/Fenton-like treatments for the management of spent resins has received considerable attention because of their potential to completely break down the IER structure into harmless compounds (e.g., carbon dioxide, water, and inorganic salts). In addition, after Fenton-like treatments, the resulting solutions containing radionuclides can be easily immobilized to stable waste. In this review, we critically discuss the key developments in Fenton/Fenton-like dissolution, degradation, and mineralization of spent resins. We describe the important reaction parameters (initial pH, resin dosage, catalyst type and dosage, hydrogen peroxide dosage, flow rates of the catalyst and oxidant, reaction temperature, treatment time, and other specific parameters) for various Fenton/Fenton-like treatments of spent resins. Moreover, this review focuses heavily on the major reaction intermediates generated in Fenton-like treatments. In the final section of this review (conclusions and perspectives), we discuss the major challenges and suggest future research directions need to be addressed to improve the efficiency of Fenton-like treatments of spent resins.

Intermediate-level radioactive waste (ILW) is not subject to legal approval for cave disposal in Korea. To solve this problem, a spent resin treatment device that separates 14C-containing resin from zeolite/activated carbon and desorbs 14C through a microwave device has been developed. In this study, we evaluated the radiological safety of the operators performing repair work in the event of a failure in such a device treating 1 ton of spent resin mixture per day. Based on the safety evaluation results, it is possible to formulate a design plan that can ensure the safety of workers while developing a commercialized device. When each component of the resin treatment device can be repaired from the outside, the maximum and minimum allowable repair times are calculated as 263.2 h and 27.7 h for the 14C-detached resin storage tank and zeolite/activated carbon storage tank, respectively. For at least 6 h per quarter, the worker's annual dose limit remains within 50 mSv/year; further, over 5 years, it remained within 100 mSv. At least 6 h of repair time per quarter is considered, under conservative conditions, to verify the radiological safety of the worker during repair work within that time.

Uplifting fractured granitic rocks occur in substantial areas of countries such as Japan. Some of these areas might be considered when siting a deep geological repository for radioactive wastes. A repository site would be selected in such an area only if it is possible to make a safety case, accounting for the changing conditions during uplift. The safety case must include robust arguments that chemical processes in the rocks around the repository will contribute sufficiently to minimise radiological doses to biosphere receptors. Numerical modelling is an important aspect of making these arguments. To provide confidence in the safety arguments, numerical models need to be sufficiently realistic, but also parameterised conservatively (pessimistically). However, model development is challenging because uplift involves many complex couplings between groundwater flow, chemical reactions between water and rock, and changing rock properties. The couplings would affect radionuclide mobilisation and retardation, by influencing diffusive radionuclide fluxes between groundwater flowing in fractures and effectively immobile porewater in the rock matrix (rock matrix diffusion, RMD) and radionuclide partitioning between water and solid phases, via: (i) mineral precipitation/dissolution; (ii) mineral alteration; and (iii) sorption/desorption. It is difficult to represent all this complexity in numerical models while showing that they are parameterised conservatively. Here we present a modelling approach, illustrated by simulation cases for some exemplar radioelements, to identify realistically conservative process conceptualisations and model parameterisations., (Copyright © 2024. Published by Elsevier B.V.)

A safety case for an underground radioactive waste repository must show that groundwater will not in future transport radionuclides from the repository to the near-surface environment (the biosphere) in harmful quantities. Safety cases are developed step-wise throughout a programme to site and develop a repository. At early stages, before a site is selected, safety cases are generic and based on simplified safety assessment models of the disposal system that have conservative parameter values. Later, when site-specific conditions are known, more realistic models are needed for the long-term geo-environmental evolution and their impacts on radionuclide migration/retention. Uplift is one such environmental change, which may be particularly important in countries near active tectonic plate boundaries, such as Japan. Here we review the state of knowledge about how the properties of fractured granitic rocks evolve during uplift, based on studies in Japan. Hence, we present conceptual models and a generic scenario for mass transport and retardation processes in uplifting granitic rocks as a basis for realistic numerical models to underpin safety assessment., (Copyright © 2024. Published by Elsevier B.V.)

In order to assess sites for a deep geological repository for storing high-level nuclear waste safely in Germany, various numerical models and tools will be in use. For their interaction within one workflow, their reproducibility, and reliability version-controlled open-source solutions and careful documentation of model setups, results and verifications are of special value. However, spatially fully resolved models including all relevant physical and chemical processes are neither computationally feasible for large domains nor is the data typically available to parameterize such models. Thus, simplified models are crucial for the pre-assessment of possible sites to narrow down the list of suitable candidates for which detailed site investigations and fully resolved models will be done at a later stage. Still, the accuracy of these simplified models is of importance as the pre-assessment of suitable sites will be based on them. In this study, we compare the modelling capabilities of TransPyREnd, a one-dimensional transport code based on finite differences, specifically developed for the fast estimation of radionuclide transport by the German federal company for radioactive waste disposal (BGE), with OpenGeoSys, which is a modelling platform based on finite elements in up to three spatial dimensions. Both codes are used in the site selection procedure for the German nuclear waste repository. The comparison of the model results of TransPyREnd and OpenGeoSys is augmented by comparisons with an analytical solution for a homogeneous material. For the purpose of numerical benchmarking, we consider a geological profile located in southern Germany as an example where the hypothetical repository is located in a clay-stone formation. TransPyREnd and OpenGeoSys yield overall similar results. However, both codes use different discretizations which impact is the highest for strongly sorbing compounds, while the difference gets negligible for less sorbing and more diffusive compounds as higher diffusion tends to blur the initial conditions. Overall, the OpenGeoSys model is more exact whereas the TransPyREnd model has considerable faster run times. We found in our example, that significant substance amounts are only leaving the host rock formation, if apparent diffusion is high, for which case both codes give similar results, while relative differences are considerable for strongly sorbing compounds. However, in the latter case no significant substance amount of radionuclides leaves the host-rock formation, thus deeming the differences in the model results minor for the overall safety assessments of sites., Competing Interests: Declaration of competing interest The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper., (Copyright © 2024 The Authors. Published by Elsevier B.V. All rights reserved.)

Iron-phosphate glasses are a wide group of materials with a wide range of applications. Among others, they are promising materials in toxic waste vitrification because of their high chemical durability and relatively low processing temperature and time. They are a novel group of glasses that are considered in the vitrification of radioactive waste, especially those that cannot be treated using conventional borosilicate ones. Since strontium isotopes are one of the main fission products present in the waste, the influence of Sr on the structural properties of the glasses is an important factor. Strontium-containing iron-phosphate glasses were subjected to structural studies using FT-IR and Raman spectroscopies. The obtained spectra were described, and appropriate band assignments were done. Based on the research conducted, the structural features of the phosphate network and their changes were determined. The results obtained showed that strontium in relatively low content up to 20 mol% acts as the glass network charge compensator and can stabilize the network. Above this threshold, SrO can be treated as a pure modifier, leading to gradual depolymerization. Thus, this point may be treated as the maximum waste loading for effective strontium immobilization., Competing Interests: Declaration of competing interest The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper., (Copyright © 2024 Elsevier B.V. All rights reserved.)

Deep geological repositories (DGRs) stand out as one of the optimal options for managing high-level radioactive waste (HLW) such as uranium (U) in the near future. Here, we provide novel insights into microbial behavior in the DGR bentonite barrier, addressing potential worst-case scenarios such as waste leakage (e.g., U) and groundwater infiltration of electron rich donors in the bentonite. After a three-year anaerobic incubation, Illumina sequencing results revealed a bacterial diversity dominated by anaerobic and spore-forming microorganisms mainly from the phylum Firmicutes. Highly U tolerant and viable bacterial isolates from the genera Peribacillus, Bacillus, and some SRB such as Desulfovibrio and Desulfosporosinus, were enriched from U-amended bentonite. The results obtained by XPS and XRD showed that U was present as U(VI) and as U(IV) species. Regarding U(VI), we have identified biogenic U(VI) phosphates, U(UO 2 )·(PO 4 ) 2, located in the inner part of the bacterial cell membranes in addition to U(VI)-adsorbed to clays such as montmorillonite. Biogenic U(IV) species as uraninite may be produced as result of bacterial enzymatic U(VI) reduction. These findings suggest that under electron donor-rich water-saturation conditions, bentonite microbial community can control U speciation, immobilizing it, and thus enhancing future DGR safety if container rupture and waste leakage occurs., Competing Interests: Declaration of Competing Interest The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper., (Copyright © 2024 The Authors. Published by Elsevier B.V. All rights reserved.)

This paper quantitatively discusses the role of subjective assessment in deciding popular acceptance for a geological disposal facility (GDF) for high-level radioactive waste. The Hirose survey that we used is an online poll answered by 1,930 Japanese adults over the age of 20; it was conducted twice, both before and after the Fukushima Daiichi Nuclear Power Plant accident, and enabled us to conduct a unique analysis. Using this balanced panel data, we focus on neutral respondents who answered Neither–Nor regarding their willingness to accept a GDF in their municipality. The analysis reached the following conclusions. First, neutral respondents were less likely to oppose GDFs, if it led to mitigating the stigma of GDF acceptance directed toward descendants and ancestors and instilling pride in the social benefits of GDF acceptance. Second, regarding the variability of GDF assessment between neutral and non-neutral respondents against external shock, neutrals tended to be more sensitive than non-neutrals. Furthermore, neutrals formed a large bracket wherein the median voter who was a decisive player in collective decision-making was likely to exist. Thus, we should reconsider the importance of neutral respondents for public decision-making about site selection of GDF.

The efficiency of decontamination of model spent ion exchange resins, contaminated with magnetite and hematite, with mineral acid solutions, and using electro-decontamination, was evaluated. It has been shown that effective hematite dissolution occurs in concentrated mineral acid solutions. However, the use of direct current increases the decontamination efficiency of spent ion exchange resins contaminated with hematite. It is determined that with increasing voltage and acid concentration, the dissolution efficiency of hematite deposits increases and can exceed 99%. It has been shown that hematite dissolution is accompanied by secondary adsorption of radionuclides due to ion exchange, which can be removed with sodium nitrate solutions.

Rare-earth (RE) industries generate a massive amount of radioactive residue containing high thorium concentrations. Due to the fact that thorium is considered a non-economic element, large volume of these RE processed residues are commonly disposed of without treatment. It is essential to study an appropriate treatment that could reduce the volume of waste for final disposition. To this end, this research investigates the applicability of carbon-based adsorbent in separating thorium from aqueous phase sulphate is obtained from the cracking and leaching process of solid rare-earth by-product residue. Adsorption of thorium from the aqueous phase sulphate by carbon-based electrodes was investigated through electrosorption experiments conducted at a duration of 180 minutes with a positive potential variable range of +0.2V to +0.6V (vs. Ag/AgCl). Through this research, the specific capacity obtained was equivalent to 1.0 to 5.14 mg-Th/g-Carbon. Furthermore, electrosorption of thorium ions from aqueous phase sulphate is found to be most favorable at a higher positive potential of +0.6V (vs. Ag/AgCl). This study's findings elucidate the removal of thorium from the rare-earth residue by carbon-based electrodes and simultaneously its potential to reduce disposal waste of untreated residue.

Both the design and safety assessment of radioactive waste disposal facilities demand an accurate evaluation of the hydraulic conductivity of the bentonite materials, especially compacted bentonite. For permeability tests of bentonite materials, the lengthy time necessary for specimen saturation and measurement may present a bottleneck. The permeability behavior of bentonite, such as the effects of the water quality and the exchangeable cations, has not been systematized sufficiently. For the present study, a hydraulic conductivity test system with 2-mm-thick specimens was developed. Its applicability was evaluated in terms of test accuracy. Six specimens of compacted Japanese sodium bentonite, with dry densities of 1.34–1.79 Mg/m3, were subjected to falling head hydraulic conductivity tests. The results showed that the hydraulic gradient set for this study did not affect the hydraulic conductivity, indicating that the macroscopic hydraulic behavior was consistent with Darcy's law. Furthermore, it was possible to reduce the test period considerably, by about one-tenth, compared to that using 10-mm-thick specimens. The obtained hydraulic conductivity was found to be similar to that in earlier studies. Furthermore, the values showed less variation particularly in terms of the consolidation test results. The results demonstrated that 2-mm-thick specimens are useful for hydraulic conductivity measurements of compacted bentonite.

Background: This work focused on cost-effective and collective adsorption of radionuclides followed by their subsequent selective desorption, which is necessary for separation. The potential of humic acid, a low-cost adsorbent, for the removal and recovery of Cs(I) and Sr(II) from simulated reactor waste, was investigated Basic Procedures: Experimental parameters like solution pH, adsorbent dose, contact time, and temperature were optimized. FT-IR of humic acid before and after adsorption was compared. Selective desorption of both the radionuclides was achieved with desorbing agents – citric acid and HCl Main findings: The process was governed by rapid kinetics where 91±2% of Cs(I) and 83±2% of Sr(II) was adsorbed within 10 min. Data modelling revealed that the process follows pseudo-second-order kinetics, and is spontaneous, feasible, and exothermic, with positive entropy indicating a high affinity of humic acid for the metallic ions. Maximum desorption of 97±2% was obtained for Sr(II), and 82±2% for Cs(I) Principal Conclusion: Thus, the adsorption of investigated radionuclides by humic acid and their desorption was time-efficient, economical and environmentally a benign alternative for the removal and recovery of Cs(I) and Sr(II).

Nowadays, evaluation of amounts and distributions of radioactive waste is an important preparatory step in the process of nuclear reactor decommissioning. For tentative estimation of radioactive waste, a cell-based rigorous 2 step (R2S) method usually is used; however, a poor resolution caused by the averaged flux and spectrum in a cell is still a great challenge because of leading to underestimated or overestimated results. To overcome the poor resolution, several systems were introduced. Neither system, however, provides any function for evaluation of radioactive waste amount and distribution. Thus, it is additionally required to classify radioactive waste based on the results of activation calculation. In this study, the advanced R2S (AR2S) system was developed. To verify the performance of the system, its results for a verification problem were compared with those of the cell-based R2S method. The results showed good agreement, which is to say, within 2.0% relative error. Also, several characteristics of fine/coarse mesh were analyzed. To demonstrate the performance of the AR2S system, the radioactive waste from the Japan Power Demonstration Reactor (JPDR) was estimated, and the result indicated a high-resolution distribution. Therefore, it is expected that the AR2S system will prove useful for precise evaluation of radioactive waste.

Indonesia has policies and strategies for the management of radioactive waste and spent nuclear fuel that arises from the use of nuclear research and development facilities, including three research reactors, and the use of radioisotopes in medicine and industries. The Indonesian government has provided extensive facilities such as an independent regulatory organization (BAPETEN) and a centralized radioactive waste management organization (CRWT-BATAN). Further, the presence of regulations and several international conventions guarantee the protection of the public from all risks due to handling radioactive waste and spent nuclear fuel. However, the sustainability of radioactive waste management in the future faces various challenges, such as disposal issues related to not only to site selection but also financing of radioactive waste management. Likewise, the problem of transportation persists; as an archipelago country, Indonesia still struggles to manage the infrastructure required for the transport of radioactive materials. The waste from the production of the radioisotopes, especially from the production of 99Mo, requires special attention because BATAN has never handled it. Indonesia should also resolve the management of NORM from various activities. In Indonesia, the definition of radioactive waste does not include NORM. Therefore, the management of this waste needs revision and improvement on the regulations, infrastructure, and technology.

We examined the (p,x), (α,x), and (n,x)reaction cross sections, for the destruction of the long-lived exotic 99Tc nuclide, which is in spent fuels of nuclear power plants and radioactive waste released from hospitals. The cross sections were calculated based on different models of nuclear level density and radiative gamma strength function. We found that there is a large uncertainty in the calculated cross sections, which is up to one order of magnitude, due to the difference in the theoretical models. The (p,n), (α,n), and (n,γ)reactions have large cross sections, which are about hundreds of millibarn. Hence, these reactions are suggested for the 99Tc transmutation. To examine the reliability of the models, the calculated cross sections were compared to evaluated data in the ENDF database. The results indicate that although all the models are consistent with the ENDF data, the Constant-temperature Fermi Gas and Kopecky–Uhl Generalized Lorentzian models are the most reliable models for the investigated reactions of 99Tc. The present study is useful for better transmutation of the radioactive waste and understanding of the dependence of the cross section on the nuclear level density and gamma strength function.

Microbial U(VI) reduction plays a major role in new bioremediation strategies for radionuclide-contaminated environments and can potentially affect the safe disposal of high-level radioactive waste in a deep geological repository. Desulfitobacterium sp. G1-2, isolated from a bentonite sample, was used to investigate its potential to reduce U(VI) in different background electrolytes: bicarbonate buffer, where a uranyl(VI)‑carbonate complex predominates, and synthetic Opalinus Clay pore water, where a uranyl(VI)-lactate complex occurs, as confirmed by time-resolved laser-induced fluorescence spectroscopic measurements. While Desulfitobacterium sp. G1-2 rapidly removed almost all U from the supernatants in bicarbonate buffer, only a low amount of U was removed in Opalinus Clay pore water. UV/Vis measurements suggest a speciation-dependent reduction by the microorganism. Scanning transmission electron microscopy coupled with energy-dispersive X-ray spectroscopy revealed the formation of two different U-containing nanoparticles inside the cells. In a subsequent step, artificial multispecies bio-aggregates were formed using derivatized polyelectrolytes with cells of Desulfitobacterium sp. G1-2 and Cobetia marina DSM 50416 to assess their potential for U(VI) reduction under aerobic and anaerobic conditions. These findings provide new perspectives on microbial U(VI) reduction and contribute to the development of a safety concept for high-level radioactive waste repositories, as well as to new bioremediation strategies., Competing Interests: Declaration of competing interest The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper., (Copyright © 2024 The Authors. Published by Elsevier B.V. All rights reserved.)

An extraction chromatography resin, prepared by the impregnation of bis-octyloxy-calix[4]arene-mono-crown-6 (BOCMC)onto an acrylic ester based polymeric support material, gave excellent uptake data for the removal of radio-cesium (Cs-137) from nitric acid feed solutions. The weight distribution coefficient (K d ) value of >300 obtained during the present study at 3 M HNO 3 was the highest reported so far while using a calix-crown-6 based extraction chromatographic resin material. Analogous resin reported previously has yielded a K d value <100 at comparable feed conditions. The sorbed metal ions could be efficiently desorbed with de-ionized water. Kinetic modeling of the uptake data indicated that both the film and the intra-particle diffusion mechanism are simultaneously operating in the sorption of Cs + ion onto the BOCMC resin. The metal ion sorption data were fitted to the sorption isotherm models and did not conform to the chemisorptions of physisorption models and indicated a pi-pi interaction between the benzene rings of the calix-crown-6 ligand and the Cs + ion. The reusability of the resins was quite satisfactory after 5 cycles and the radiation stability of the resin material was very good upto an absorbed dose of 500 kGy. The results of column studies were quite encouraging with 15 mL (9 bed volumes) as the breakthrough volume while the elution was complete in about 12 bed volumes of de-ionized water., Competing Interests: Declaration of competing interest The authors declare no conflict of interest., (Copyright © 2024 Elsevier B.V. All rights reserved.)

Multiple bench-scale filtration campaigns of Hanford tank waste supernatant on a backpulseable dead-end filtration skid have provided greater insight into the solids that cause fouling and reduce filter performance. The solids collected during each campaign were concentrated from the backpulse solutions and examined using automated particle analysis (APA) methods with scanning electron microscopy and X-ray energy dispersive spectroscopy to categorize particle types and their morphological characteristics. We show that with APA, thousands of particles can be analyzed to provide accurate insight into the phases that may be impacting filter performance., Competing Interests: Declaration of Competing Interest All authors declare no competing financial interests., (Copyright © 2024 Elsevier B.V. All rights reserved.)

Objectives: The purpose of this study is to estimate the activity release from near surface disposal repository for a hypothetical open pool light-water research reactor (OP-LWR) in the Riyadh region, Saudi Arabia. Methods: mathematical model was used to model six cement barriers near surface disposal repository. Water infiltration into the repository and into the radioactive waste is assumed to be the main cause of activity release into the environment. The hydraulic conductivity, specific mass, the porosity of the cement, and the local geosphere were considered, as well as the expected radioactive waste inventory produced by this reactor. Results: The results show that there is a direct proportionality between distribution coefficient, retardation factor, and the mean time to collapse (MTTC). Also, the results show that nickel has the highest mean time to collapse among all the other radionuclides in the proposed model. The highest activity release in the biosphere was found to be 1.1×104 Bq for Tc-99 after 1300 years. Conclusions: The impact of the radiological behavior of each selected radionuclide enabled finding the effect on the repository performance.

Radioactive iodine is considered one of the most dangerous radioactive elements in nuclear waste. Therefore, effective capture of radioactive iodine is essential for developing and using nuclear energy to solve the energy crisis. Some materials that have been developed for removing radioactive iodine still suffer from complex synthesis, low removal capacity, and non-reusability. Herein, covalent organic framework (COF)/chitosan (CS) aerogels were prepared using vacuum freeze-drying, and the COF nanoparticles were tightly attached on the green biomass material CS networks. Due to the synergistic effect of both COF and CS, the composite aerogel shows a three-dimensional porous and stable structure in the recycle usage. The COF/CS aerogel exhibits excellent iodine adsorption capacity of 2211.58 mg g -1 and 5.62 g g -1 for static iodine solution and iodine vapor, respectively, better than some common adsorbents. Furthermore, COF/CS aerogel demonstrated good recyclability performance with 87 % of the initial adsorption capacity after 5 cycles. In addition, the interaction between iodine and imine groups, amino groups, and benzene rings of aerogel are the possible adsorption mechanisms. COF/CS aerogel has excellent adsorption properties, good chemical stability, and reusable performance, which is a potential and efficient adsorbent for industrial radioactive iodine adsorption from nuclear waste., Competing Interests: Declaration of competing interest The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper., (Copyright © 2024 Elsevier B.V. All rights reserved.)

The sorption removal of radionuclides Sr2+ using a freestanding functional membrane is an interesting and significant research area in the remediation of radioactive wastes. Herein, a novel self-assembled membrane consisting of metal–organic framework (MOF) nanobelts and graphene oxides (GOs) are synthesized through a simple and facile filtration method. The membrane possesses a unique interwove morphology as evidenced from SEM images. Batch experiments suggest that the GO/Ni-MOF composite membrane could remove Sr2+ ions from aqueous solutions and the Sr2+ adsorption capacity and efficiency of the GO/Ni-MOF composite membrane is relevant to the MOF content in the composite. Thus, the dominant interaction mechanism was interface or surface complexation, electrostatic interaction as well as ion substitution. The maximum effective sorption of Sr2+ over GO/Ni-MOF membrane is 32.99% with 2 mg composite membrane containing a high content of Ni-MOF at 299 K in 100 mg/L Sr2+ aqueous solution. The FT-IR and XPS results suggest that the synergistic effect between GO and Ni-MOF is determinant in the sorption Sr2+ process. The GO/Ni-MOF composite membrane is demonstrated to have the advantages of efficient removal of Sr2+, low cost and simple synthesis route, which is promising in the elimination of radionuclide contamination.

In the ANSICHT project that was jointly carried out by DBE TECHNOLOGY GmbH, BGR, and GRS gGmbH, two generic geological site models were used to develop a first draft of a methodology to demonstrate the safety of a high-level waste (HLW) repository in argillaceous formations in Germany, taking into account the regulatory requirements. The main results of the project are characterised by the developed repository concepts adapted to the geological conditions. The specific quantifications of the integrity criteria and their exemplary application with calculational proofs were used to demonstrate the integrity of the host rocks. The development of site-specific FEP (features, events, and processes) cataloges provided a complete system description for evaluation of the repository evolution. The developed work flow of the demonstration concept illustrated the complete sequence of the safety proof in a transparent way. It shows that various steps have to be performed, possibly iteratively, to provide a successful safety proof. The results form a useful tool in the pending search for a HLW repository site, especially when providing a basis for comparing safety analyses of different sites in Germany.

Repositories for deep geological disposal of radioactive waste rely on multi-barrier systems to isolate waste from the biosphere. A multi-barrier system typically comprises the natural geological barrier provided by the repository host rock – in our case the Opalinus Clay – and an engineered barrier system (EBS). The Swiss repository concept for spent fuel and vitrified high-level waste (HLW) consists of waste canisters, which are emplaced horizontally in the middle of an emplacement gallery and are separated from the gallery wall by granular backfill material (GBM). We describe here a selection of five in-situ experiments where characteristic hydro-mechanical (HM) and thermo-hydro-mechanical (THM) processes have been observed. The first example is a coupled HM and mine-by test where the evolution of the excavation damaged zone (EDZ) was monitored around a gallery in the Opalinus Clay (ED-B experiment). Measurements of pore-water pressures and convergences due to stress redistribution during excavation highlighted the HM behaviour. The same measurements were subsequently carried out in a heater test (HE-D) where we were able to characterise the Opalinus Clay in terms of its THM behaviour. These yielded detailed data to better understand the THM behaviours of the granular backfill and the natural host rock. For a presentation of the Swiss concept for HLW storage, we designed three demonstration experiments that were subsequently implemented in the Mont Terri rock laboratory: (1) the engineered barrier (EB) experiment, (2) the in-situ heater test on key-THM processes and parameters (HE-E) experiment, and (3) the full-scale emplacement (FE) experiment. The first demonstration experiment has been dismantled, but the last two ones are on-going.

The mission of nuclear safety authorities in national radioactive waste disposal programmes is to ensure that people and the environment are protected against the hazards of ionising radiations emitted by the waste. It implies the establishment of safety requirements and the oversight of the activities of the waste management organisation in charge of implementing the programme. In Belgium, the safety requirements for geological disposal rest on the following principles: defence-in-depth, demonstrability and the radiation protection principles elaborated by the International Commission on Radiological Protection (ICRP). Applying these principles requires notably an appropriate identification and characterisation of the processes upon which the safety functions fulfilled by the disposal system rely and of the processes that may affect the system performance. Therefore, research and development (R&D) on safety-relevant thermo-hydro-mechanical-chemical (THMC) issues is important to build confidence in the safety assessment. This paper points out the key THMC processes that might influence radionuclide transport in a disposal system and its surrounding environment, considering the dynamic nature of these processes. Their nature and significance are expected to change according to prevailing internal and external conditions, which evolve from the repository construction phase to the whole heating–cooling cycle of decaying waste after closure. As these processes have a potential impact on safety, it is essential to identify and to understand them properly when developing a disposal concept to ensure compliance with relevant safety requirements. In particular, the investigation of THMC processes is needed to manage uncertainties. This includes the identification and characterisation of uncertainties as well as for the understanding of their safety-relevance. R&D may also be necessary to reduce uncertainties of which the magnitude does not allow demonstrating the safety of the disposal system.

To perform the permanent disposal of metal hydroxide waste from electro-kinetic decontamination, it is necessary to secure the technology for its solidification. The integrity tests on the fabricated solidification should also meet the criteria of the Korea Radioactive Waste Agency. We carried out the solidification of metal hydroxide waste using cement solidification. The integrity tests such as the compressive strength, immersion, leach, and irradiation tests on the fabricated cement solidifications were performed. It was also confirmed that these requirements of the criteria of Korea Radioactive Waste Agency on these cement solidifications were met. The microstructures of all the cement solidifications were analyzed and discussed.

A novel biosorbent, immobilized Saccharomyces cerevisiae in magnetic chitosan microspheres was prepared, characterized, and used for the removal of Sr2+ from aqueous solution. The structure and morphology of immobilized S. cerevisiae before and after Sr2+adsorption were observed using scanning electron microscopy with energy dispersive X-ray spectroscopy. The experimental results showed that the Langmuir and Freundlich isotherm models could be used to describe the Sr2+ adsorption onto immobilized S. cerevisiae microspheres. The maximal adsorption capacity (qm) was calculated to be 81.96 mg/g by the Langmuir model. Immobilized S. cerevisiae was an effective adsorbent for the Sr2+ removal from aqueous solution.

Research on eco-friendly bioremediation strategies for mitigating the environmental impact of toxic metals has gained attention in the last years. Among all promising solutions, bentonite clays, to be used as artificial barriers to isolate radioactive wastes within the deep geological repository (DGR) concept, have emerged as effective reservoir of microorganisms with remarkable bioremediation potential. The present study aims to investigate the impact of bentonite fungi in the speciation and mobility of selenium (Se) and tellurium (Te), as natural analogues 79 Se and 132 Te present in radioactive waste, to screen for those strains with bioremediation potential within the context of DGR. For this purpose, a multidisciplinary approach combining microbiology, biochemistry, and microscopy was performed. Notably, Aspergillus sp. 3A demonstrated a high tolerance to Te(IV) and Se(IV), as evidenced by minimal inhibitory concentrations of >16 and >32 mM, respectively, along with high tolerance indexes. The high metalloid tolerance of Aspergillus sp. 3A is mediated by its capability to reduce these mobile and toxic elements to their elemental less soluble forms [Te(0) and Se(0)], forming nanostructures of various morphologies. Advanced electron microscopy techniques revealed intracellular Te(0) manifesting as amorphous needle-like nanoparticles and extracellular Te(0) forming substantial microspheres and irregular accumulations, characterized by a trigonal crystalline phase. Similarly, Se(0) exhibited a diverse array of morphologies, including hexagonal, irregular, and needle-shaped structures, accompanied by a monoclinic crystalline phase. The formation of less mobile Te(0) and Se(0) nanostructures through novel and environmentally friendly processes by Aspergillus sp. 3A suggests it would be an excellent candidate for bioremediation in contaminated environments, such as the vicinity of deep geological repositories. It moreover holds immense potential for the recovery and synthesis of Te and Se nanostructures for use in numerous biotechnological and biomedical applications., Competing Interests: Declaration of competing interest The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper., (Copyright © 2023 The Author(s). Published by Elsevier B.V. All rights reserved.)

The presence of phosphorus in borosilicate glass (at 0.1 - 1.3 mol% P 2 O 5 ) and in iron-phosphate glass (at 53 mol% P 2 O 5 ) stimulated the growth and metabolic activity of anaerobic bacteria in model systems. Dissolution of these phosphorus containing glasses was either inhibited or accelerated by microbial metabolic activity, depending on the solution chemistry and the glass composition. The breakdown of organic carbon to volatile fatty acids increased glass dissolution. The interaction of microbially reduced Fe(II) with phosphorus-containing glass under anoxic conditions decreased dissolution rates, whereas the interaction of Fe(III) with phosphorus-containing glass under oxic conditions increased glass dissolution. Phosphorus addition to borosilicate glasses did not significantly affect the microbial species present, however, the diversity of the microbial community was enhanced on the surface of the iron phosphate glass. Results demonstrate the potential for microbes to influence the geochemistry of radioactive waste disposal environments with implication for wasteform durability., Competing Interests: Declaration of Competing Interest The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper., (Copyright © 2023 The Authors. Published by Elsevier B.V. All rights reserved.)

Microorganisms show a high affinity for trivalent actinides and lanthanides, which play an important role in the safe disposal of high-level radioactive waste as well as in the mining of various rare earth elements. The interaction of the lanthanide Eu(III) with the sulfate-reducing microorganism Desulfosporosinus hippei DSM 8344 T , a representative of the genus Desulfosporosinus that naturally occurs in clay rock and bentonite, was investigated. Eu(III) is often used as a non-radioactive analogue for the trivalent actinides Pu(III), Am(III), and Cm(III), which contribute to a major part of the radiotoxicity of the nuclear waste. D. hippei DSM 8344 T showed a weak interaction with Eu(III), most likely due to a complexation with lactate in artificial Opalinus Clay pore water. Hence, a low removal of the lanthanide from the supernatant was observed. Scanning transmission electron microscopy coupled with energy-dispersive X-ray spectroscopy revealed a bioprecipitation of Eu(III) with phosphates potentially excreted from the cells. This demonstrates that the ongoing interaction mechanisms are more complex than a simple biosorption process. The bioprecipitation was also verified by luminescence spectroscopy, which showed that the formation of the Eu(III) phosphate compounds starts almost immediately after the addition of the cells. Moreover, chemical microscopy provided information on the local distribution of the different Eu(III) species in the formed cell aggregates. These results provide first insights into the interaction mechanisms of Eu(III) with sulfate-reducing bacteria and contribute to a comprehensive safety concept for a high-level radioactive waste repository, as well as to a better understanding of the fate of heavy metals (especially rare earth elements) in the environment., Competing Interests: Declaration of Competing Interest The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper., (Copyright © 2023 The Authors. Published by Elsevier Inc. All rights reserved.)