The objective of this study was to determine the ultrasonication-assisted extraction conditions that maximize the DPPH radical scavenging activity of extracts obtained from the stems of Lespedeza bicolor Turcz through the application of the Response Surface Methodology (RSM). Before delving into the analysis of extraction conditions using the RSM model, we conducted efficiency validation of ultrasonication-assisted extraction and executed single-factor experiments for ethanol concentration, extraction time, and extraction temperature. The data obtained from these single-factor experiments were employed to construct the Box-Behnken Design (BBD). In these results, in the single-factor experiments, it was evident that the parameters for ethanol concentration, extraction time, and extraction temperature exhibited quadratic trends. The single-factor experiments allowed us to discern the trends for each parameter leading to the maximum antioxidant capacity, and this data was subsequently applied to the BBD. Following the completion of initial experiments, a Response Surface Methodology (RSM) model was constructed based on Box-Behnken Design (BBD). According to the predictive model developed in this study, it was anticipated that performing ultrasonic-assisted extraction for 85.0412 minutes at an ethanol concentration of 32.573% and an extraction temperature of 51.5608°C will result in a DPPH radical scavenging activity of 79.7146%. The predictive results were statistically verified through a comparative analysis with actual measurements and ANOVA analysis, confirming the statistical significance of the model. The finding of this study underscore the significance of optimizing extraction conditions in the precise quantification of the antioxidant potential for economic advantages in both experimental and industrial contexts.

수서곤충은 일반적으로 하루살이목, 날도래목, 강도래목, 잠자리목, 노린재목, 딱정벌레목, 뱀잠자리목, 파리 목의 8개의 목(Order)에 속하며, 물을 기반으로 서식하는 곤충들을 의미한다. 국내 수서곤충의 연구는 도입기 (40-50년대), 형성기(60년대), 발전기(70-80년대), 도약기(90-2010년대)를 거쳐 왔다. 물을 기반으로 수질을 평가 하기 위해 수서곤충은 생태학적 연구가 먼저 시작되었으며, 이후 분류학적 연구가 시행되고 현재는 유전학적 및 응용학적 연구로 다양성을 증명하고 있다. 하지만, 물을 기반으로 하는 수서곤충의 서식처는 도시화로 인한 다양한 개발과 교란으로 감소하고 사라지고 있다. 서식처뿐만 아니라 수서곤충을 연구하는 학자들과 학생들도 정체기와 쇠퇴기를 거치고 있다. 국내 수서곤충의 다양성 종 목록은 1999년 처음으로 학술논단에서 491종이 정리되었으며, 이후 약 10년 만에 개정목록이 2011년에 988종으로 수정되었다. 이후 지속적인 국가생물다양성 확보 및 보전 전략에 따라 미발굴된 수서 파리류들이 대거 기록되면서 2018년에는 처음 기록보다 3배 이상 증가하 여 1,567종이 보고되었다. 이처럼 표면적으로는 수서곤충의 연구가 활성화되어 다양성은 늘어가고 있는 것처럼 보이지만 다양한 문제점들이 나타나고 있다. 수서곤충은 성충보다는 유충을 주로 연구하는 학문으로 많은 연구 자가 어려움을 겪고 있다. 단기간의 다양성 증진보다는 심도 있는 다양한 주제로 접근하는 방법이 필요할 것으로 보이며, 수서곤충의 다양성과 함께 서식처 회복 및 복원 방안도 함께 마련되어야 할 것으로 판단된다. 본 발표에서 는 국내 수서곤충의 다양한 연구주제와 피해 상황 및 활용방안을 소개하여 미래의 수서곤충의 방향성과 연구 방향을 심도 있게 논의하고자 한다.

Bangladeshi medicinal plants (BMP) have a history of traditional use in treating chronic inflammatory diseases, but a BMP bark’s antioxidant and anti-inflammatory effects remain largely unexplored. This study assessed methanolic extracts’ antioxidant and anti-inflammatory properties from the bark of 15 medicinal plant species native to Bangladesh. The methanol extracts of BMP bark were evaluated for their total antioxidant activity and ability to counteract inflammation induced by lipopolysaccharide (LPS) in RAW 264.7 macrophages. Among the 15 bark extracts from BMP, Albizia odoratissima (A. odoratissima), Engelhardia spicata (E. spicata), and Shorea robusta (S. robusta) showed the highest total phenolic contents and total antioxidant capacity by effectively scavenging free radicals. In particular, these three bark extracts significantly reduced the mRNA expression of LPS-induced inflammatory cytokines and enzymes inducible by inflammation in macrophages. Also, the mRNA expression of NADPH oxidase 2 was significantly suppressed by the three bark extracts in LPS-induced RAW 264.7 macrophages. These results suggest that out of the 15 bark extracts obtained from medicinal plants in Bangladesh, the extracts from A. odoratissima, E. spicata, and S. robusta exhibit substantial total antioxidant capacity by efficiently scavenging free radicals and also inhibit LPS-induced inflammation in macrophages.

In May 2023, a free-living marine nematode species from the genus Parapinnanema was identified in the subtidal zone of Ulleungdo Island, the East Sea, Korea. Specimens were collected using the Smith-McIntyre Grab. These specimens exhibited close similarities to Parapinnanema imbricatum from the sublittoral of Moneron Island, particularly in terms of general characteristics, such as the detailed structure of the buccal cavity, the complex and ringed structure of the cuticle, the copulatory apparatus, spinneret, and the female genital system. However, the Korean specimens of Parapinnanema imbricatum also displayed distinctive features compared to the original description, including a relatively elongated body (3,317-4,339 μm vs. 3,100-4,200 μm) and a narrower body width (66-77 μm vs. 71-85 μm). Additionally, the diameter of the head was relatively shorter (24-29 μm vs. 28-36 μm). This paper offers a comprehensive morphological description, along with illustrations and DIC photomicrographs, of P. imbricatum from Korean waters.

This study explores the impact of metal doping on the surface structure of spent nuclear fuels (SNFs), particularly uranium dioxide (UO2). SNFs undergo significant microstructural changes during irradiation, affecting their physical and chemical properties. Certain elements, including actinides and lanthanides, can integrate into the UO2 lattice, leading to non-stoichiometry based on their oxidation state and environmental conditions. These modifications are closely linked to phenomena like corrosion and oxidation of UO2, making it essential to thoroughly characterize SNFs influenced by specific element doping for disposal or interim storage decisions. The research employs X-ray diffraction (XRD), scanning electron microscopy (SEM), and Raman spectroscopy to investigate the surface structure of UO2 samples doped with elements such as Nd3+, Gd3+, Zr4+, Th4+, and ε-particles (Mo, Ru, Pd). To manufacture these samples, UO2 powders are mixed and pelletized with the respective dopant oxide powders. The resulting pellet samples are sintered under specific conditions. The XRD analysis reveals that the lattice parameters of (U,Nd)O2, (U,Gd)O2, (U,Zr)O2, and (U,Th)O2 linearly vary with increasing doping levels, suggesting the formation of solid solutions. SEM images show that the grain size decreases with higher doping levels in (U,Gd)O2, (U,Nd)O2, and (U,Zr)O2, while the change is less pronounced in (U,Th)O2. Raman spectroscopy uncovers that U0.9Gd0.1O2-x and U0.9Nd0.1O2-x exhibit defect structures related to oxygen vacancies, induced by trivalent elements replacing U4+, distorting the UO2 lattice. In contrast, U0.9Zr0.1O2 shows no oxygen vacancy-related defects but features a distinct peak, likely indicating the formation of a ZrO8-type complex within the UO2 lattice. ε-Particle doped uranium dioxide shows minimal deviations in surface properties compared to pure UO2. This structural characterization of metal-doped and ε-particle-doped UO2 enhances our understanding of spent nuclear fuel behavior, with implications for the characterization of radioactive materials. This research provides valuable insights into how specific element doping affects the properties of SNFs, which is crucial for managing and disposing of these materials safely.

Most of the C-14 produced is in the organic form, generated as methane (14CH4), methanol (14CH3OH), formaldehyde (14CH2O), and formic acid (14CO2H2). When analyzing C-14, it is transformed into the form of 14CO2, and its concentration is determined using LSC. Typical examples include the wet oxidation method, the combustion or Pyrolysis. The wet oxidation method uses strong acids and involves repeated operations, which generates large amounts of acid waste and secondary radioactive waste. The combustion method uses high temperatures, which requires an oxygen device. Pyrolysis also requires high temperature in a vacuum and catalysts. Catalysts are expensive because they are platinum-based. To compensate for these shortcomings, a C-14 analysis method using UV irradiation was developed. In this study, 100 mL of distilled water mixed with formic acid (CO2H2), potassium persulfate (K2S2O8), and silver nitrate (AgNO3) was irradiated with a 320-390 nm UV lamp to conduct a CO2 production reaction experiment. The UV range was measured using a photometer (UV Power puck II). The beaker was made of quartz in 150 mL size with three inlets : a temperature measurement, a sample inlet, and a collection tube connector. We changed the UV lamp used from a 450 W halogen lamp to a 100 W LED, which has a lower temperature and is safer. As a result of the experiment, CO2 bubbles were generated in the collection tube, due to the UV irradiation react, which uses oxidizer and catalysts. The maximum temperature of the solution irradiated with the LED UV lamp was less than 56°C. It confirmed the rate of bubble generation changed depending on the lamp distance, the amount of sample, oxidizer, and catalyst. In an experiment to confirm the reaction caused by heat, it was found that although a reaction occurred due to heat, the reaction was significantly lower than when using a UV lamp. The reproducibility experiment was conducted three times in total under the same conditions. It showed the same pattern. In the future, we plan to select mock samples, collect 14CO2 in Carbo- Sorb, and analyze them using LSC. The results of this research will be used as a technology to recover C-14 more safely and efficiently and will also be used to expand its application to the treatment of other wastes such as waste liquid and waste resin through simulated samples.

Molten chloride salts have received considerable research attention as potential nuclear fuel and coolant candidates for molten salt reactors. However, there are several challenges, especially for structural materials due to the selective dissolution of chromium (Cr) in the molten chloride salts environment. Understanding the compatibility of uranium (U), which is used as nuclear fuel in molten salt reactors, with Cr in molten chloride salts is critical for designing the molten salt reactor structure. Therefore, in this study, the cyclic voltammetry (CV) was used to investigate the electrochemical behaviors of U and Cr. The diffusion coefficients and formal potentials were obtained. The electrochemical properties of uranium and chromium were investigated by CV in molten NaCl-MgCl2 salt at 600°C. Tungsten rods for working and counter electrode, and Ag/AgCl for reference electrode were utilized in this experiment. UCl3 made from the chemical dissolution of U rods and CrCl2 (Sigma-Aldrich, 99.99%) were used. Diffusion coefficients (D) of U and Cr were calculated by measuring reduction peak current of U3+/U and Cr2+/Cr from CV curves and using the Berzins-Delahay equation; D (U3+/U) = 3.0×10-5 cm2s-1 and D (Cr2+/Cr) = 3.3×10-5 cm2s-1. The formal potentials were also calculated by using the reduction peak potential obtained from CV results; E0’ (U3+/U) = -1.173 V and E0’ (Cr2+/Cr) = -0.321 V. The ionization tendency was investigated by comparing each reduction peak potential. The reduction peak potential Ep,c was increasing order of Ep,c (U3+/U) < Ep,c (Cr2+/Cr) < Ep,c (U4+/U3+). It can be seen that in the presence of U4+ and Cr metals, the Cr in the alloy can dissolve into Cr2+, but in the presence of U3+ and Cr metals, the Cr in the alloy does not dissolve into Cr2+. By analyzing the CV curve, diffusion coefficients and formal standard potentials were obtained. The result of comparing reduction peak potentials suggests that the nuclear fuel using U4+ should be inhibited to prevent the selective dissolution of Cr.

This program aims to build a specialized and converged educational platform for the training of students in the back-end nuclear fuel cycle and cultivate integrated human resources encompassing majors, generations, and fields. To achieve this, we have established an infrastructure for integrated education and training in the radiochemistry and back-end nuclear fuel cycle and operated specialized educational courses linked with special lectures, experimental practices, and field trips. Firstly, to construct an integrated educational and training infrastructure for the back-end nuclear fuel cycle, we formed a committee of experts from both inside and outside the institution and built an advanced radiochemistry laboratory equipped with physical and chemical analysis instruments. Through a comprehensive educational program involving theory, experiments, and discussions, we have established an integrated curriculum across adjacent majors and interdisciplinary studies. We also operate short-term education and experimental training programs (e.g., summer and winter schools for the back-end nuclear fuel cycle). Secondly, the program has connected leading researchers domestically and internationally, as well as the next generation of scholars. The program offers long-term educational opportunities and internships targeting both undergraduate and graduate students. To support this, we continuously offer expert colloquiums and individual research internships. Through regular committee meetings and workshops, we focus on nurturing the integrated talents necessary for the back-end nuclear fuel cycle. Through this program, students from various fields are being trained as competent integrated human resources capable of addressing various issues in the back-end nuclear fuel cycle. It is expected that this will enable us to supply specialized technical personnel in the back-end nuclear field in line with mid-to-long-term demands.

Wolsong Unit 1, a domestic heavy water reactor nuclear power plant, was permanently shut down in December 2019. Accordingly, Wolsong Unit 1 plans to prepare a Final Decommissioning Plan (FDP), submit it to the government by 2024, receive approval for decommissioning, and begin full-scale decommissioning. One of the important tasks in the decommissioning of Wolsong Unit 1 is to determine the decommissioning strategy. It is necessary to decide on a decommissioning strategy considering various factors and variables, secure the technical background, and justify it. The selection of a decommissioning strategy is best achieved through the use of formal decisionmaking assistance techniques, such as considerations related to influencing factors. It is very important to understand the basic decommissioning strategy alternatives and whether sufficient consideration has been given to situations where only a single unit is permanently shut down in a multi-unit site like Wolsong Unit 1, while the remaining units are in normal operation. As a process for selecting a decommissioning strategy, first, all considerations that could potentially affect decommissioning presented in the KINS Decommissioning Safety Review Guidelines were synthesized, influencing factors to be used in the decision-making process were determined, and the concept was defined. In order to select the most appropriate decommissioning strategy by considering various evaluation attributes of possible decommissioning alternatives (immediate dismantling and delayed dismantling), the Wolsong Unit 1 decommissioning strategy was evaluated by reflecting the AHP decision-making technique.

The physicochemical similarities of hydrogen isotopes have made their separation a challenging task. Conventional methods such as cryogenic distillation, Girdler sulfide process, chromatography, and thermal cycling absorption have low separation factors and are energy-intensive. To overcome these limitations, research has focused on kinetic quantum sieving (KQS) and chemical affinity quantum sieving (CAQS) effects for selective separation of hydrogen isotopes. Porous materials such as metal-organic frameworks (MOF), covalent organic frameworks (COF), zeolites, carbon, and organic cages have been studied for hydrogen separation. In this study, we focus the enhancement for CAQS to provide the cations due to the chemical affinity between hydrogen isotope and unsaturated sites by cations in zeolite beads. Cation exchanged zeolite beads was synthesized with cobalt, copper, nickel, iron and silver in zeolite 4A beads. Synthesized cation exchanged zeolite was analyzed for the surface area and pore size in N2 and adsorption behaviors of hydrogen isotopes (D2/H2) for various cation exchanged zeolite beads using BET at 77 K. The study predicts the D2/H2 adsorption selectivity based on the results obtained with BET. These hydrogen isotope adsorption results will provide a foundation for future processes for tritium separation.

Pt/C catalysts were prepared using black carbon (CB), and evaluated for their potential application as a catalyst of liquid-phase catalystic exchange for tritium treatment. CB was treated with 10% H2O2 solution for 0 and 2 hours at 105°C, Ethylene glycol and 40wt% Pt were added to the dried treated sample to prepare a Pt/C catalyst. The physical and chemical properties of the prepared catalysts were evaluated by BET, XRD, elemental analysis (EA), and TEM analyses. As a result of BET analysis, the surface area of CB without 10% H2O2 was 237.2 m2·g-1, and after treatment with 10% H2O2, it decreased to 181.2 m2·g-1 for 2 hours. However, the internal surface area increased, indicating the possibility that more Pt could be distributed inside the CB treated with 10% H2O2. In the XRD analysis results, the presence of Pt was confirmed by observing the Pt peak in the prepared Pt/C catalyst, and it was also observed through TEM analysis that Pt was evenly distributed within the CB. The elemental analysis (EA) results showed that the ratio of S and N decreased and the ratio of O increased with increasing 10% H2O2 treatment time. The H2O2 treated carbon supported Pt catalysts and polytetrafluoroethylene were then loaded together on a foamed nickel carrier to obtain hydrophobic catalysts. Our hydrophobic Pt catalyst using H2O2 treated black carbon are expected to be usefully used in the tritium treatment system.

Radioactive waste (hereinafter referred to as mixed waste) containing hazardous substances (heavy metals, organic and inorganic waste liquids, asbestos, etc.) has been continuously generated from domestic nuclear power plants, nuclear facilities, and other industrial facilities, and heavy metals were released during the dismantlement of Kori Unit 1 and Wolseong Unit 1. Lead, cadmium, mercury, arsenic), asbestos, decontamination waste liquid (organic/inorganic waste liquid), etc. may be generated. Although hazardous waste related to the nuclear industry continues to be generated, only the regulation direction for hazardous substances is presented in the provisions related to hazardous substances in the delivery regulations for low and intermediate-level radioactive waste and the acceptance criteria for low and intermediate-level radioactive waste disposal facilities. In particular, because there is no clear definition of “hazardousness” and specific standards such as concentration and characteristics for classification of hazardous substances, as well as hazard removal procedures when the hazardousness of radioactive waste is confirmed, no hazardous substances have been delivered in Korea to date and many mixed wastes are stored at each generation facility or at the NPP. As a plan to improve delivery standards related to mixed waste is being prepared recently, it is believed that if the acceptance standards are revised accordingly, it will be possible to confirm the suitability for disposal of drums produced after the establishment of the acceptance standards in 2015. However, it is believed that securing disposal suitability for waste that was packed in 200L drums and compressed under super high pressure in the absence of specific technical standards and regulatory guidelines for the disposal of radioactive waste containing hazardous substances would still remain a difficult problem. In this report overseas acceptance standards related to hazardous waste were reviewed and a plan to secure the disposal suitability of 200 L drums compressed with of super high pressure was proposed.

Activated carbon (AC) is used for filtering organic and radioactive particles, in liquid and ventilation systems, respectively. Spent ACs (SACs) are stored till decaying to clearance level before disposal, but some SACs are found to contain C-14, a radioactive isotopes 5,730 years halflife, at a concentration greater than clearance level concentration, 1 Bq/g. However, without waste acceptance criteria (WAC) regarding SACs, SACs are not delivered for disposal at current situation. Therefore, this paper aims to perform a preliminary disposal safety examination to provide fundamental data to establish WAC regarding SACs SACs are inorganic ash composed mostly of carbon (~88%) with few other elements (S, H, O, etc.). Some of these SACs produced from NPPs are found to contain C-14 at concentration up to very-low level waste (VLLW) criteria, and few up to low-level waste (LLW) criteria. As SACs are in form of bead or pellets, dispersion may become a concern, thus requiring conditioning to be indispersible, and considering VLL soils can be disposed by packaging into soft-bags, VLL SACs can also be disposed in the same way, provided SACs are dried to meet free water requirement. But, further analysis is required to evaluate radioactive inventory before disposal. Disposability of SACs is examined based on domestic WAC’s requirement on physical and chemical characteristics. Firstly, particulate regulation would be satisfied, as commonly used ACs in filters are in size greater than 0.3 mm, which is greater than regulated particle size of 0.2 mm and below. Secondly, chelating content regulation would be satisfied, as SACs do not contain chelating chemicals. Also, cellulose, which is known to produce chelating agent (ISA), would be degraded and removed as ACs are produced by pyrolysis at 1,000°C, while thermal degradation of cellulose occurs around 350~600°C. Thirdly, ignitability regulation would be satisfied because as per 40 CFR 261.21, ignitable material is defined with ignition point below 60°C, but SACs has ignition point above 350°C. Lastly, gas generation regulation would be satisfied, as SACs being inorganic, they would be targeted for biological degradation, which is one of the main mechanism of gas generation. Therefore, SACs would be suitable to be disposed at domestic repositories, provided they are securely packaged. Further analysis would be required before disposal to determine detailed radioactive inventories and chemical contents, which also would be used to produce fundamental data to establish WAC.

Radioactive iodine-129, a byproduct of nuclear fission in nuclear power plants, presents significant environmental and health risks due to its high solubility in water and volatility. Iodine-129, with its half-life of 1.57×1017 years, necessitates safe management and disposal. Therefore, safely capturing and managing I-129 during spent nuclear fuel reprocessing is of paramount importance. To address these challenges, various glass waste forms containing silver iodide have been developed, such as borosilicate, silver phosphate, silver vanadate, and silver tellurite glasses. These glasses effectively immobilize iodine, but the high cost of silver raises affordability concerns. This study introduces CuI·Cu2O·TeO2 glass waste forms for iodine immobilization, a novel approach. The cost-effectiveness of copper, in contrast to silver, makes it an attractive alternative. The CuI·Cu2O·TeO2 glass waste forms were synthesized with varying CuI content (x) in (1-x)(0.3Cu2O·0.7TeO2) glass matrices. Xray diffraction (XRD) confirmed amorphous structures, and X-ray fluorescence (XRF) quantified composition. X-ray photoelectron spectroscopy (XPS) and Raman spectroscopy provided insights into structural properties. Durability assessments using a 7-day product consistency test (PCT-A) and inductively coupled plasma-mass spectrometry (ICP-MS) revealed compliance with U.S. glass regulations, making CuI·Cu2O·TeO2 glasses a promising choice for iodine immobilization in radioactive waste.

Domestic waste acceptance criteria (WAC) require flowable or homogeneous wastes, such as spent resin, concentrated waste, and sludge, etc., to be solidified regardless of radiation level, to provide structural integrity to prevent collapse of repository, and prevent leaching. Therefore, verylow level (VLL) spent resin (SR) would also require to be solidified. However, such disposal would be too conservative, considering IAEA standards do not require robust containment and shielding of VLL wastes. To prevent unnecessary cost and exposure to workers, current WAC advisable to be amended, thus this paper aims to provide modified regulation based on reviewed engineering background of solidification requirement. According to NRC report, SR is classified as wet-solid waste, which is defined as a solid waste produced from liquid system, thus containing free-liquid within the waste. NRC requires liquid wastes to be solidified regardless of radiation level to prevent free liquid from being disposed, which could cause rapid release of radionuclides. Furthermore, considering class A waste does not require structural integrity, unlike class B and C wastes, dewatering would be an enough measure for solidification. This is supported by the cases of Palo Verde and Diablo Canyon nuclear power plants, whose wet-solid wastes, such as concentrated wastes and sludge, are disposed by packaging into steel boxes after dewatering or incineration. Therefore, dewatering VLL spent resin and packaging them into structural secure packaging could satisfy solidification goal. Another goal of solidification is to provide structural support, which was considered to prevent collapse of soil covers in landfills or trenches. However, providing structural support via solidification agent (ex. Cement) would be unnecessary in domestic 2nd phase repository. As the domestic 2nd phase repository is cementitious structure, which is backfilled with cement upon closure, the repository itself already has enough structural integrity to prevent collapse. Goldsim simulation was run to evaluate radiation impact by VLL SR, with and without solidification, by modelling solidified wastes with simple leaching, and unsolidified wastes with instant release. Both simulations showed negligible impact on radiation exposure, meaning that solidifying VLL SR to delay leaching would be irrational. Therefore, dewatering VLL SR and packaging it into a secure drum (ex. Steel drum) could achieve solidification goals described in NRC reports and provide enough safety to be disposed into domestic repositories. In future, the studied backgrounds in this paper should be considered to modify current WAC to achieve efficient waste management.

The effect of various physicochemical processes, such as seawater intrusion, on the performance of the engineered barrier should be closely analyzed to precisely assess the safety of high-level radioactive waste repository. In order to evaluate the impact of such processes on the performance of the engineered barrier, a thermal-hydrological-chemical model was developed by using COMSOL Multiphysics and PHREEQC. The coupling of two software was achieved through the application of a sequential non-iterative approach. Model verification was executed through a comparative analysis between the outcomes derived from the developed model and those obtained in prior investigations. Two data were in a good agreement, demonstrating the model is capable of simulating aqueous speciation, adsorption, precipitation, and dissolution. Using the developed model, the geochemical evolution of bentonite buffer under a general condition was simulated as a base case. The model domain consists of 0.5 m of bentonite and 49.5 m of granite. The uraninite (UO2) was assigned at the canister-bentonite interface as the potential source of uranium. Assuming the lifetime of canister as 1,000 years, the porewater mixing without uranium leakage was simulated for 1,000 years. After then, the uranium leakage through the dissolution of uraninite was initiated and simulated for additional 1,000 years. In the base case model, where the porewater mixing between the bentonite and granite was the only considered process, the gypsum tended to dissolve throughout the bentonite, while it precipitated in the vicinity of bentonite-granite boundary. However, the precipitation and dissolution of gypsum only showed a limited effect on the performance of the bentonite. Due to the low solubility of uraninite in the reduced environment, only infinitesimal amounts of uranium dissolved and transported through the bentonite. Additional cases considering various environmental processes, such as seawater or cement porewater intrusion, will be further investigated.

The nuclide management process for reducing the environmental burden being developed by the Korea Atomic Energy Research Institute is performed in molten salts, resulting in contaminated salt wastes containing fission products such as Cs, Sr, Ba, and rare-earth nuclides. In addition, the spent fuel of a molten salt reactor (MSR) contains a variety of fission products, and a purification process may be required for the reuse of the salt and the separation and disposal of the fission products in the spent nuclear fuel. The melt-crystallization method is a technique used for the purification and separation of chemicals or metals based on the different melting points of the different substances. In a recent study, our group developed a reactive-crystallization method using Li2CO3 precipitation agent to precipitate metal corrosion from the reactor through a chlorination reaction by HCl and Cl2, which may occur in chloride molten salt, and successfully precipitated the metal precipitate and purified and recovered LiCl salt. In this study, reactive-crystallization method has been established for removing fission products and corrosive materials. Using the reactive crystallization method, white LiCl-KCl salt that was not discolored by metal corrosion was recovered through the crystallization plates, and fission products and metal elements were shown to be suppressed to several ppm in the purified salt. Consequently, high-purity salts were recovered with high nuclide and corrosive separation efficiencies. The reactive crystallization procedure can also be applied to other salt waste systems, such as MSR nuclear fuel treatment and molten salt chemistry for the elimination of corrosive substances.

아메리카동애등에 성충은 음식물 폐자원 등 유기물이 있는 곳에 알을 낳는 습성이 있다. 대부분의 농가는 음식 물폐자원을 가공한 단미사료(습식사료)를 유인배지로 활용하여 그 위에 플로랄폼(오아시스)를 놓고 알을 받는 다. 그러나 플로랄폼은 재사용이 불가하고 생분해되지 않는 환경폐기물로서 처리가 곤란하며 포름알데하이드, 카본 블랙 등의 발암물질을 함유한 것으로도 알려져 있다. 이에 본 연구는 먹이원 자체를 활용하여 폐기물이 발생하지 않는 친환경 산란받이를 개발하였으며 일회용으로 사용되는 플로랄폼을 대체하였다. 먹이원으로 활 용할 수 있는 습식사료와 건식사료를 주재료로 하여 제작하며, 습식사료(수분60~80%)와 건식사료(1~10%)를 1:0.5~1 비율로 혼합한 사료 혼합물과 보조첨가제와 물을 포함하여 제작한다. 친환경 산란받이는 기존 플로랄폼 대비 산란율이 34% 증가하였으며 구매비용 또한 75% 절감하였다.

담배거세미나방(Spodoptera litura), 열대거세미나방(Spodoptera frugiperda) 및 파밤나방(Spodoptera exigua) 은 광식성 해충이지만, 종특이적인 기주범위를 갖는다. 이들이 기주식물을 찾아가는 과정에 냄새감각이 어떤 역할을 하는지 알아보기 위해, 기주 및 비기주 식물에서 발산되는 46가지 휘발성 물질을 선정하여, 이들에 대한 나방 3종의 냄새반응을 GC-EAD(gas chromatography-electroantennogram detection)를 통해 확인하였다. 그 결과, 46가지의 식물유래화합물 중 9가지 물질이 3종의 나방 모두에서 냄새활성을 나타냈으며, 2가지 물질은 담배거세 미나방과 열대거세미나방에만 냄새활성을 나타냈고, 다른 몇 가지 물질은 담배거세미나방에만 냄새활성을 나 타냈다. 이 결과는 세 종 나방이 식물냄새물질 탐지를 위해 유사한 냄새감각세포를 가지며, 일부 종에서는 종특이 적인 냄새감각세포가 존재한다는 것을 보여준다. 이 결과를 바탕으로 냄새활성을 나타낸 물질들의 나방 3종에 대한 행동활성을 야외트랩실험을 통해 확인할 예정이다.

Sweet pepper(paprika) belongs to the genus Capsicum, and is one of the most important export product from Korea to Japan and Southeast Asia. So it is important to eradicate plant quarantine pests before export sweet pepper. Aphids, whiteflies and mites are major pests that can damage to sweet peppers. Fumigation is normally used to eradicate pests in plant quarantine, but phytotoxicity may can be appeared that affect the quality of the product. Low-temperature treatment, one of the most popular physical treatment, can reduce crop damage to preserve product quality, but it takes long time to kill pests, which can cause quality degradation. In this study, phytotoxicity of fumigants, phosphine(PH3), ethyl formate(EF) and PH3+EF on sweet peppers was investigated to use as basic data for physicochemical treatment. When treated with more than 35 mg/L of EF, phytotoxicity was occurred, and was not occurred with PH3. When low-temperature of 1.7 degrees treated for 15 days after fumigation, it seems to be no direct damage from low-temperature treatment. But quality of top of sweet pepper was decreased from 7 days after fumigation.