Microbes in insect gut significantly influence host physiology. While Lepidoptera is a diverse insect order, the relationship between microbial symbiosis and host development remains elusive, especially concerning role of gut-colonizing bacteria in metamorphosis. We investigated the gut microbial diversity in Galleria mellonella throughout its life cycle using 16S rRNA amplicon sequencing. Our findings revealed a predominance of Enterococcus spp. in larvae and Enterobacter spp. in pupae. Remarkably, removing Enterococcus spp. hastened the larval-to-pupal transition. Transcriptome analysis showed an upregulation of immune response genes in pupae and hormone genes in larvae. Notably, the production of antimicrobial peptides in the host gut varied with developmental stages. Some of these peptides suppressed the growth of Enterococcus innesii, a dominant gut bacterium in G. mellonella larvae. This research underscores the pivotal role of gut microbiota shifts in metamorphosis, driven by the secretion of antimicrobial peptides in the G. mellonella digestive system.

Since 1997, the Republic of Korea (ROK) has been developing pyro-processing (Pyro) technology to reduce the disposal burden of high-level radioactive waste by recycling spent nuclear fuel (SNF). Compared to plutonium and uranium extraction process, Korean Pyro technology has relatively excellent proliferation resistance that cannot separate pure plutonium owing to its intrinsic characteristics. Regarding Pyro technology development of ROK, the Bush administration considered that Pyro is not reprocessing under the Global Nuclear Energy Partnership, whereas the Obama administration considered that Pyro is subject to reprocessing. However, the Bush and Obama administrations did not allow ROK to conduct full Pyro activities using SNF, even though ROK had faithfully complied with international nonproliferation obligations. This is because the US nuclear nonproliferation policy to prevent the spread of sensitive technologies, such as enrichment and reprocessing, has a strong effect on ROK, unlike Japan, on a bilateral level beyond the NPT regime for non-proliferation of nuclear weapons.

In this study, a milled carbon nanofiber-reinforced composite paint was prepared to enhance the anti-corrosive properties of concrete structures. Shorter-length (40 μm) milled carbon fibers (MCFs) showed an increased viscosity relative to longer MCFs (120 μm) owing to their 2 weeks (the decrease was especially strong in the acid solution). A carbon nanotube (CNT)- reinforced composite paint showed similar results in uniform distribution in the epoxy resin. The latter showed a decrease in viscosity owing to agglomerative movement in the epoxy resin. The surface hardness and tensile strength of the composite paint linearly increased as the carbon nanofiber loading was increased by up to 7.2 wt% in the epoxy resin, and slowly decreased after soaking in a sulfuric acid or sodium hydroxide solution for to those of the MCFs, whereas CNTs dispersed in isopropyl alcohol (IPA) in advance and mixed with resin showed lower hardness values than those without dispersion in IPA at the same loading. The mechanical properties such as the Shore D hardness and tensile strength of the MCF-reinforced composite paint increased significantly, resulting in a slower surface degradation of the composite paint concrete in a sulfuric acid and sodium hydroxide solution.

The bioreduction process from soluble U(VI) to insoluble U(IV) has been extensively studied in the field of radionuclides migration. Since acetic acid (AcOH) is widely used as an electron donor for bioreduction of U(VI), it is necessary to understand the effect of U(VI)-AcOH complexes that exist in different species depending on pH on this process. Changes in samples before and after bioreduction can be compared using time-resolved laser luminescence spectroscopy (TRLLS), which measures the characteristic luminescence spectra of different U(VI) species. Although luminescence properties of U(VI)-AcOH species were reported, experiments were conducted under conditions below pH 4.5. In this study, spectrophotometry and TRLLS for U(VI)-AcOH species (10−100 μM U(VI) and 20 mM AcOH) were performed in pH ranges extending to neutral and alkaline pH regions similar to groundwater conditions as well as acidic pH region. Two different complexes (UO2(AcO)+, UO2(AcO)2 with U(VI) and acetate ratios of 1:1, 1:2) were observed in the acidic pH region. The 1:1 complex, which appears as the pH increases, has no luminescence properties, but its presence can be confirmed because it serves to reduce the luminescence intensity of UO2 2+. In contrast, the 1:2 complex exhibits distinct luminescence properties that distinguish it from UO2 2+. The 1:3 complex (UO2(AcO)3 -) expected to appear with increasing pH was not observed. Two different complexes ((UO2)3(OH)5 +, (UO2)3(OH)7 - with U(VI) and OH ratios of 3:5, 3:7) were observed as the major species in the neutral pH region, but their luminescence lifetimes are remarkably short compared those in the absence of AcOH. Solid U(VI) particles were observed in the alkaline pH region, and they also had completely different luminescence properties from the aforementioned U(VI)-AcOH and U(VI)-hydrolysis species. Based on these results, the effect of pH in the presence of AcOH on the bioreduction process from U(VI) to U(IV) will be discussed.

Bacterial metabolisms influence the behavior of uranium (U) in deep geological repository (DGR) system because bacteria are ubiquitous in the natural environment. Nevertheless, most studies for the U(VI) bioreduction have focused on a few model bacterium, such as Shewanella putrefaciens, Desulfovibrio desulfuricans, and Geobacter sulfurreducens. In this study, the potential of aqueous U(VI) ((U(VI)aq) reduction by indigenous bacteria was examined under anaerobic conditions with addition of 20 mM sodium acetate for 24 weeks. Three different indigenous bacterial communities obtained from granitic groundwater at depths of 44–60 m (S1), 92–116 m (S2), and 234–244 m (S3) were applied for U(VI)aq reduction experiments. The S2 groundwater contained the highest U concentration of 885.4 μg/L among three groundwater samples, where U mainly existed in the form of Ca2UO2(CO3)3(aq). The S2 groundwater amended 20 mM of sodium acetate was used for the U(VI)aq bioreduction experiment. Variations in the U(VI)aq concentration and redox potential were monitored for 24 weeks to compare U(VI)aq removal efficiency in response to indigenous bacteria. The U(VI)aq removal efficiencies varied among three indigenous bacteria: 57.8% (S3), 43.1% (S2), and 37.7% (S1). The presence of the thermodynamically stable uranyl carbonate complex resulted in the incomplete U(VI)aq removal. Significant shifts in indigenous bacterial communities were observed through highthroughput 16S rRNA gene sequencing analysis. Two SRB species, Thermodesulfovibrio yellowstonii and Desulfatirhabdium butyrativorans, were dominant in the S3 sample after the anaerobic reaction, which enhanced the bioreduction of U(VI)aq. The precipitates produced by bacterial activity were determined to be U(IV)-silicate nanoparticles by a transmission electron microscope (TEM)-energy dispersive spectroscope (EDS) analysis. These results demonstrated that considerable U immobilization is possible by stimulating the activity of indigenous bacteria in the DGR environment.

The effects of an individual effective dose from radioactive contamination that will remain during site reuse after the decommissioning of nuclear facilities is generally assessed using the RESRAD code. The calculated results should meet the site reuse criteria presented by regulators, 0.25 mSv/yr in the United States and 0.1 mSv/yr in Korea. After completion of decommissioning, the dose is not subject to measurement, resulting in Derived Concentration Guideline Level (DCGL) remaining at the site that is practically consistent with the dose criteria. In order to assess dose using the RESRAD code, various requirements will need to be considered and determined, where the selection of input parameters is one of the important factors in the dose assessment. In addition, appropriate selection of site-specific parameters is important to reflect the site characteristics of each decommissioned Nuclear Power Plant (NPP). Therefore, this study intends to analyze the impact of site-specific parameters by referring to the cases of overseas decommissioned NPPs. In order to evaluate doses using RESRAD code, a site reuse scenario must first be selected. In general, in the case of unrestricted reuse, the resident farmer scenario can be applied, so the resident farmer scenario was also selected in this study. In addition, once a resident farmer scenario is selected, input parameters are selected according to the scenario, and the input parameter inputs a single value or distribution according to the deterministic or probabilistic evaluation method. Therefore, since this study is to evaluate the effect on site-specific parameters, a single value was applied as a deterministic evaluation method. For the 10 site-specific parameters considered in overseas cases, the difference was set twice using the F9 function key in the RESRAD code and the results were analyzed. In this study, we used prior research data targeting domestic nuclear facility for sensitivity analysis. Related parameters include the category of contamination layer, soil, water transport, ingestion, and occupancy. The parameters that appeared as the greatest influence among the 10 parameters were different in radionuclide on the contaminated zone. We showed the changes according to the difference in input parameters was presented using the graph provided by the RESRAD code. As a result, in the evaluation for Co-60 in this study, no significant change was observed. However, in case of H-3, several parameters values were changed, indicating that the effect on dose will be different depending on the site characteristics of the nuclear facilities.

The Derived Concentration Guideline Level (DCGL) using RESRAD code is generally obtained for the reuse of the site and remaining buildings of the decommissioning of nuclear facilities. At this time, the evaluation first considers wide DCGL assuming homogenous contamination for the entire target site. The DCGL derived through this will be compared with the actual contamination measured at the Final Status Survey (FSS) stage to determine whether the site is compliance with criteria. Guidelines for Survey units are presented in MARSSIM and suggested in Class 1 through 3. Therefore, DCGL for the survey unit of a certain smaller area is established by applying a correction factor from wide DCGL, which is define as an Area Factor (AF). Therefore, this study reviewed the AF applied in overseas cases, reviewed the necessary factors for derivation, and compared them by applying factors to the preliminary experimental target area for domestic nuclear installations. The AF is the ratio of the dose from the base-case contaminated area to the dose from a smaller contaminated area with the same radioactive concentration. To this end, an unrestricted resident farmer scenario was applied as the site reuse scenario, which deals with all exposure pathways considered in the RESRAD. The potential exposure pathways considered in resident farmer scenarios are largely divided into external and internal exposures, which are based on NUREG/CR-5512. In addition, in order to calculate the AF, a change in the contaminated area occurs, and accordingly, a variable that varies according to the area, i.e., length parallel to aquifer flow (LCZPAQ), the contaminated fraction of plant food ingested (FPLANT), the contaminated fraction of meat and milk (FMEAT and FMILK), is accompanied. As the contamination area decreases, these variables decrease, and the criteria for reduction were reflected through overseas cases. In this study, three nuclides (C-14, Co-60, and Cs-137) were assumed as representative nuclides, and the area of the contaminated site was selected as 50,000 m2 and reduced at a certain rate. As a result, each nuclide showed different characteristics, but in general, AF increases as the area decreases. Compared to the area of this study, AF values were calculated to be smaller than those of overseas cases, but it was confirmed that the area of the values showed similar patterns. In addition, in the case of C-14, the slope of AF increased rapidly as the area decreased, while Co-60 and Cs-137 showed similar slopes.

The type of radioactive waste that may occur in the process of nuclear power plant dismantling can be classified into solid, liquid, gas, and mixed waste. The amount of these wastes must be defined in the Final Decommissioning Plan for approval of the licensing. Also, in the case of liquid radioactive waste, it is necessary to calculate the generation amount in order to treat radioactive waste at a Radioactive Waste Treatment Facility (RWTF) or on-site. In this regard, there is no Code and Standard for the amount of liquid radioactive waste generated during NPP are dismantled, but ANSI/NS-55.6 describes the amount of liquid radioactive waste generated from a light water reactor type NPP. This code is applied to nuclear power-related facilities such as domestic NPP and radioactive waste disposal facility. Therefore, this review intends to suggest an application plan for domestic NPP decommissioning through codes for liquid radioactive waste expected to generate during nuclear power plant decommissioning.

The type of radioactive waste that may occur in the process of NPP dismantling can be classified into solid, liquid, gas, and mixed waste. Most of the radioactive waste generated during the dismantling of a NPP is metal solid waste, but liquid radioactive waste is also a very important factor in terms of radiation environmental impact assessment. In the case of liquid radioactive waste, it is necessary to calculate the generation amount in order to design liquid radioactive waste processing system of Radioactive Waste Treatment Facility (RWTF). Depending on the amount of liquid radioactive waste generated, the type of liquid radioactive waste processing system included in the RWTF is different. In addition, in order to apply to the domestic RWTF, it is important to secure the site area occupied by the each system, the liquid radioactive waste treatment capacity of the system, and how to secure circulating water used for dilution and discharge of liquid radioactive waste. Therefore, this review aims to suggest an optimal method for the treatment system for liquid radioactive waste included in RWTF of Wolseong.

In case of Korea, unlike overseas nuclear power plants, adjacent units are located in permanently stopped nuclear power plants. Radioactive substances from airborne and liquid effluents are released into the environment from the NPP, and the radioactivity of the released substances must be reported to the regulatory authorities. Radioactive effluents are released into the environment not only in operation but also after permanent shutdown. Due to domestic conditions in which multiple units exist on the same site, it is necessary to consider radioactive effluents generated after permanent shutdown of NPPs. In particular, liquid effluent may have an increased tritium concentration due to draining the spent fuel pool. This paper summarizes the annual liquid emissions of PWR power plants that have been permanently shut down. The data was obtained from the Nuclear Regulatory Commission’s (NRC) annual radioactive effluent release report, which provides information on the annual emissions power plants into the environment. The liquid emissions of each plant were organized into an annual table, providing an overview of the amount of liquid released by each plant. This study aims to raise awareness about the potential environmental impact of permanently shut down nuclear power plants and the need for proper management of their liquid emissions. The findings of this study can used by operator, policymakers, and other stakeholders to make informed decisions regarding the decommissioning and management of nuclear power plants.

The type of radioactive waste that may occur in the process of nuclear power plant dismantling can be classified into solid, liquid, gas, and mixed waste. In addition, according to the level of radioactivity, it can be divided into high level, intermediate level, low level, and clearance level waste. In the case of solid radioactive waste, it is necessary to secure disposal suitability in order to deliver it to a disposal facility, so safe and efficient treatment of a large amount of radioactive waste generated during decommissioning is one of the most important issues. For the treatment of radioactive waste generated during decommissioning, technologies in various fields such as cutting, decontamination, melting, measurement, and packaging are required. Therefore, this study intends to present and application plan for decommissioning domestic nuclear power plants through overseas case studies for the treatment of radioactive waste expected to occur during nuclear power plant decommissioning.

The concept of clearance is to manage radioactive waste by incineration, reclamation, or recycling as non-radioactive waste, excluding those found to have a concentration of less than the allowable concentration of clearance. Among the types of waste subject to clearance, concrete is managed by recycling and landfill, metal by recycling and reuse, combustible materials by incineration, and soil by landfill. In Korea, clearance has been implemented in earnest since 2000, and the types and quantity of waste subject to clearance are increasing. For clearance, the nuclear-related operator submits its clearance plan to the regulatory body, and the regulatory body reviews the clearance plan and notifies the operator of its suitability. Since a significant amount of radioactive waste generated when decommissioning nuclear power plants is expected to be classified as clearance waste, this study will present clearance waste disposal measures for nuclear power plant through a review of overseas cases related to clearance.

The mobility of uranium (U) in the environment of a deep geological repository is controlled by various geochemical conditions and parameters. In particular, oxidation state of uranium is considered as a major factor to control the mobility of uranium in most of geological environments. In this study, therefore, we investigated the mobility of uranium in a deep geological repository by a natural analogue approach using a uranium deposit in the Ogcheon Metamorphic Belt (OMB). Uranium contents of rock samples from the study site ranged from 1.3 to 71 ppm (average 17.4 ppm). Uranium minerals found in the study site were mostly uraninite (UIVO2+x) and uranothorite ((UIV, Th)SiO4). The concentrations of U in the groundwater samples were very low (0.025~0.690 ppb) even though redox conditions are weakly oxidizing. Calculation results for U speciation in groundwater samples showed that major dissolved uranium species in the groundwater samples are mainly as calcium uranyl (UO2 2+) carbonate complexes such as Ca2UO2(CO3)3(aq) and CaUO2(CO3)3 2-. However, the activity ratios between 234U and 238U (AR(234U/238U)) showed U behavior in reducing conditions although the groundwater conditions were not reducing conditions and major dissolved U species were U(VI) species. Results from electron microscopic analyses for rock samples showed that major uranium minerals were U(IV) minerals such as uraninite and uranothorite. We could not identify other uranyl minerals and altered minerals from uraninite. This means that the geochemical condition of the study site has been maintained a reducing condition although the groundwater condition was a weakly oxidizing condition. Thus, the dissolution of uranium is strongly limited by the low solubility of uraninite. It is not obvious how the reducing condition of the study site has been maintained. Reducing agents such as pyrite, organic materials, and reducing bacteria might contribute to maintaining the reducing condition although further studies will be necessary. Results from this study imply that uranium mobility will be greatly limited by low dissolution of uraninite into groundwater if the reducing condition is well reserved. This limited mobility of uranium will be also contributed by low possibility of uraninite alteration into uranyl minerals which have a higher solubility than uraninite.

When the recycling technology of spent nuclear fuels (SNF) for future nuclear reactor systems and the treatment technology of SNF for disposing of in a disposal site use a molten salt such as LiCl-KCl eutectic as a processing medium one of the essential unit processes is a distillation process that remove the salt component mixed with fission products recovered. Especially, in case of Pyro-SFR recycling system the recovered nuclear fuel materials such as U, TRU and some of rare earths come from main three processes (electro-refining, electro-winning, and drawdown processes) for recycling of SNF. These recovered fuel materials contain large portion of molten salt or liquid cadmium which requires removal of them by distillation. In spent nuclear fuels discharged from PWR the portion of composing element is as follows. Uranium is about 95%, other actinides such as transuranic elements (TRU; Np, Pu, Am, Cm) is about 1%, the rare earths (lanthanides) is about 1%, and the other elements is about 3%. For example, americium (Am) in the recovered fuel materials has a problem that the reported loss of Am inevitably occurs during the vacuum salt distillation operation. A new segregation method of AMM (actinide metal mixture)–salt system is based on the difference in melting point of the actinide elements. It is possible to apply this segregation method to recovering other actinides from AMM with accompanied salt because of relatively large amount and lower melting point of a specific element in other actinides avoiding vacuum salt distillation. This new segregation method successfully tested using a surrogate element such as aluminum due to its similar melting point with a specific element. The segregation principle is solid-liquid separation, thus the solidified actinides mixture ingot can take out of a molten salt medium.

Instead of using expensive platinum, carbon anodes could potentially be utilized in the process of reducing oxides in LiCl-Li2O molten salt at a high cell potential. However, this high potential leads to the generation of a mixture of anodic gases containing toxic and corrosive gases such as chlorine (Cl2), oxygen (O2), carbon monoxide (CO), and carbon dioxide (CO2). To better understand this gas mixture, we conducted real-time analyses of the gases generated on the carbon anode during the TiO reduction reaction in the molten salt at 650°C, using a MAX-300-LG gas analyzer. Our results indicate that the ratio of CO/O2/CO2/Cl2 in the gas mixture is significantly influenced by the composition of the salt, and that removing the sources of oxygen ions in the salt increases the likelihood of generating toxic and corrosive Cl2 gas.

Pyroprocessing is a promising technique for the treatment of damaged fuel debris (corium) generated by severe nuclear accidents. The debris typically consists of (U, Zr)O2 originating from the UO2 fuel and Zr alloy-based cladding. By converting the corium to a metallic form, the principal components of the fuel can be recovered through subsequent electrorefining, allowing for long-term storage or final disposal. A study investigated the reduction of zirconium oxide compounds by Li metal as a reductant in molten LiCl salt. This research explored the feasibility of treating damaged nuclear fuel debris, which mainly consists of (U, Zr)O2. The results showed that ZrO2 was successfully reduced to Zr metal by Li metal in LiCl salt at 650C without the formation of Li2ZrO3. In particular, Zr metal was produced without the formation of Li2ZrO3 when LiCl salt containing a high concentration of Li metal was used. However, Zr metal was produced with Li2ZrO3 when LiCl salt containing both Li metal and Li2O was added. This suggests that the concentration of Li metal in the LiCl salt is an important factor in determining the formation of Li2ZrO3. The study also demonstrated that Li2ZrO3 was partially reduced to Zr metal by Li metal in LiCl salt. This finding suggests that Li metal may be effective in reducing other oxide compounds in molten LiCl salt, which could be useful in the treatment of corium. Overall, the research provides valuable insights into the feasibility of using pyroprocessing for the treatment of corium. The ability to recover and store the principal components of the fuel through electrorefining could have important implications for the long-term management of nuclear waste.

Uranium inventory in Boeun aquifer is facing the artificial reservoir that intended for supplying water to nearby cities (40-70 m apart) where, toxic radionuclides might mobile and enter the reservoir. In order to understand U mobility in the system, groundwater and fracture filling materials (FFMs) were analyzed for microbial signatures, C, O, Fe, S and U-series isotopes. The δ18O-H2O and 14C signatures suggested groundwater was originated from upland recharges dominantly and not affected by mixing with the surface water. However, the 234U/238U activity ratios (ARs) and 230Th/234U ARs in FFMs ranged from 0.93 to 1.67 and from 0.22 to 1.97, respectively, indicating that U was mobile along the fractures. In shallow FFMs, the U accumulations (~157 mg/kg) were found with Fe enrichments (~226798 mg/kg) and anomalies of δ56Fe and δ57Fe, implied U mobility in shallow depths was associated with Fe-rich environment. Also, in the shallow depths, Fe-oxidizers, Gallionella was prevailing in groundwater while Acidovorax was abundant near U ore depth. The Fe-rich environment in shallow depths was formed by pyrite dissolution, demonstrated using δ34S-SO4 and δ18O-SO4 distribution. Conclusively, the Fe-rich aquifer was capable of immobilizing the dissolved U through biotic and abiotic processes, without significant discharge into the nearby reservoir.

Transition metal oxides formed by a single or heterogeneous combination of transition metal ions and oxygen ions have various types of crystal structures, which can be classified as layered structures and non-layered structures. With non-layered structures, it is difficult to realize a two-dimensional structure using conventional synthesis methods. In this study, we report the synthesis of cobalt oxide into wafer-scale nanosheets using a surfactant-assisted method. A monolayer of ionized surfactant at the water-air interface acts as a flexible template for direct cobalt oxide crystallization below. The nanosheets synthesized on the water surface can be easily transferred to an arbitrary substrate. In addition, the synthesizing morphological and crystal structures of the nanosheets were analyzed according to the reaction temperatures. The electrochemical properties of the synthesized nanosheets were also measured at each temperature. The nanosheets synthesized at 70 °C exhibited higher catalytic properties for the oxygen evolution reaction than those synthesized at other temperatures. This work suggests the possibility of changing material performance by adjusting synthesis temperature when synthesizing 2D nanomaterials using a wide range of functional oxides, resulting in improved physical properties.