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.

토복령은 우수한 항균, 항산화, 항염증 효능을 가진 소재로 알려져 있다. 이러한 토복령(Smilax china)의 추출물의 기능성을 화장품에 적용하기 위한 기초연구로써 토복령에서 발견되는 플라보노이드인 quercetin, catechin, naringenin의 농도별 경피 투과 특성을 조사할 필요성이 있다. Marzulli의 정의에 적 용한 결과 케르세틴의 Kp 값은 0.1 mg/mL에서 "빠름"으로 분류되었고, 0.2 및 0.4 mg/mL에서 "보통"으로 분류되었다. 특히, 농도가 증가함에 따라 투과 속도가 감소하는 경향이 있었다. 나린제닌의 경우 Flux 값은 각각 0.1, 0.2 및 0.4 mg/mL 농도에서 0.69, 1.07 및 1.42 μg/hr/cm²이었으며, 해당 Kp 값은 각각 6.95, 5.34 및 3.56이었다. 나린제닌의 Kp 값은 모든 농도에서 "보통" 범주에 속하며, 케르세틴과 관찰된 것과 같이 농도가 높아짐에 따라 투과 속도가 감소하였다. 카테킨의 경우 Flux 값은 각각 0.1, 0.2 및 0.4 mg/mL 농도에서 0.75, 1.09 및 1.66 μg/hr/cm²이었으며, 해당 Kp 값은 각각 7.55, 5.46 및 4.16이었다. 카테킨의 Kp 값은 모든 농도에서 일관되게 "보통"으로 분류되었다. 여드름 저해능 및 항염증 효능이 우수 한 토복령 추출물의 유효성분인 quercetin, catechin, naringenin의 경피 투과 특성이 보통 이상으로 나타나 기능성 화장품에 사용할 수 있는 우수한 천연물 소재인 것을 확인할 수 있었다.

Henricia specimens were collected using a dual approach of trimix scuba diving and fishing nets. This inclusive collection encompasses the discovery of two species highlighted in this study and introduces and provides comprehensive descriptions for Henricia kinkasana and Henricia longispina aleutica. The descriptions offered in this study were derived from the thorough examinations of external morphological characteristics. The documentation provides detailed insight into key traits related to the abactinal and actinal skeletons and spines of these newly recorded species in Korea. This comprehensive examination contributes to our understanding of the distinct morphological characteristics defining each species within the genus Henricia.

최근 정원에 대한 관심이 증가함에 따라 자생 식물을 활용 하기 위한 연구가 진행되고 있다. 참식나무(Neolitsea sericea (Blume) Koidz).는 관상용뿐만 아니라 목재로도 이용할 수 있 다. 따라서 이 연구는 참식나무를 정원 식물 소재로 활용하기 위한 대량 증식 기술 개발을 위하여 종자의 발아 특성과 저장 특성에 대해 알아보고자 진행하였다. 2022년 11월 2일 제주 지역에서 채집한 종자를 실험에 사용하였다. 종자의 내·외부 형태 관찰을 한 결과, 종자가 탈리되는 시점에 배가 완전히 발 달한 형태를 가지고 있기 때문에 형태적휴면(morphological dormancy, MD)이 없음을 확인할 수 있었다. 수분흡수 실험 을 통하여 참식나무 종자는 72시간 이후 40% 정도의 수분흡수 율을 보여 내과피의 불투성에 의한 물리적휴면(PY, physical dormancy)이 없는 것으로 판단하였다. 채종 후 곧바로 무가 온 온실에 파종하면 이듬해 5월에 약 60%가 발아하였다. 저 온습윤처리 0, 2, 4, 8, 12주 후 25/15°C에 배양한 결과 최 종 발아율은 각각 0%, 0%, 21.1%, 32.6%, 34.7%였다. GA3 를 0, 10, 100, 1000mg·L-1 농도로 처리한 후 25/15°C에서 배양한 결과, 최종 발아율은 각각 0, 14.5, 22.7, 42.2%였다. 따라서 본 실험의 결과 8주 이상의 저온 처리 또는 GA3 처리 를 통하여 어느 정도 휴면을 타파시킬 수 있었다. 이러한 결과 를 통하여 종자는 intermediate PD를 가지는 것으로 판단 된다. 일부 발아하지 않은 종자는 deep PD를 가지고 있는 것 으로 판단하였다. 후숙 처리 실험의 결과 4주 만에 수분 함량 이 2% 이하로 감소하였고 종자는 전혀 발아하지 않았다. 따라 서 종자의 저장 특성 중 recalcittrant 특성을 지닌 난저장성 종자로 보인다.

Molten Salt Reactor, which employs molten salt mixture as fuel, has many advantages in reactor size and operation compared to conventional nuclear reactor. In developing Molten Salt Reactor, the behavior of fission product in operation should be preliminary evaluated for the correct design of reactor and its associated system including off-gas treatment. In this study, for 100 Mw 46 KCl- 54 UCl3 based Molten Salt Reactor with operating life time of 20 year, the fission product behavior was estimated by thermodynamic modeling employing FactSage 8.2. Total inventory of all fission product were firstly calculated using OpenMC code allowing depletion during neutronic calculation. Then, among all inventory, 46 element species from Uranium to Holmium were chosen and given to the input for equilibrium module of Factsage with its mass. In phase equilibrium calculation, for the correct description of solution phase, KCl-UCl3 solution database based on modified quasichemical model in the quadruplet approximation (ANL/CFCT-21/04) was employed and the coexisting solid phase was assumed to pure state. With the assumption of no oxygen and moisture ingress into reactor system, equilibrium calculation showed that 1% of solid phase and of gas phase were newly formed and, in gas phase, major species were identified : ZrCl4 (47%), Xe (33%), UCl4 (14%), Kr (5%), Ar (1%) and others. This result reveals that off-gas treatment of system should account for the appropriate treatment of ZrCl4 and UCl4 besides treatment of noble gas such as Xe and Kr.

In the decommissioning process of nuclear power plants, Ni-59, Ni-63 and Fe-55 present in radioactive waste are crucial radionuclides used as fundamental indicators in determining waste treatment methods. However, due to their low-energy emissions, the chemical separation of these two radionuclides is essential compared to others. Therefore, this study aims to evaluate the suitability of various pre-treatment methods for decommissioning waste materials by conducting characteristic assessments at each chemical separation stage. The goal is to find the most optimized pre-treatment method for the analysis of Ni-59, Ni-63 and Fe-55 in decommissioning waste. The comparative evaluation results confirm that the chemical separation procedures for Fe and Ni are very stable in terms of stepwise recovery rates and the removal of interfering radionuclides. However, decommissioning waste materials, which mainly consist of concrete, metals, etc., possess unique properties, and a significant portion may be low-radioactivity waste suitable for on-site disposal. Considering that the chemical behavior and reaction characteristics may vary at each chemical separation stage depending on the matrix properties of the materials, it is considered necessary to apply cascading chemical separation or develop and apply individual chemical separation methods. This should be done by verifying and validating their effectiveness on actual decommissioning waste materials.

The Korea Research Institute of Standards and Science has developed certified reference materials (concrete, soil, and metal radioactive liquid) for measuring gamma-emitting radionuclides to improve and maintain the quality assurance and quality control of the radioactivity measurement in decommissioning nuclear power plants. The raw materials that make up each CRM were mixed in an appropriate ratio with radionuclides. For certification and homogeneity assessment, 10 bottles were randomly selected, two sub-samples were collected from each bottle, and radionuclides were measured via HPGe gamma spectrometry. The results of the homogeneity tests using a one-way analysis of variance on the radionuclides in the CRMs fulfilled the requirements of ISO Guide 35. Coincidence summing and self-absorption correction were performed on measurement results by introducing the Monte Carlo efficiency transfer code and Monte Carlo N-Particle transport code. In concrete analysis, the reference values for five radionuclides (60Co, 241Am, 134Cs, and 137Cs) in the CRM were in the range of 15-40 Bq/kg, and the expanded uncertainty was within 10% (k = 2). In soil analysis, the reference values for the 137Cs and 60Co were 118.7 and 124.4 Bq/kg, and the expanded uncertainty was within 10% (k = 2). In metal radioactive liquid analysis, the reference values for 134Cs, 137Cs and 60Co in the CRM were in the range of 200-270 Bq/kg, and the expanded uncertainty was within 7% (k = 2).

The presence of organic components in spent scintillation liquid should be considered during all steps of radioactive waste processing for final disposal. Scintillation liquids often referred to as cocktails are generated form radiochemical analyses of radionuclides, which mainly consists of mixtures of liquid organic materials such as toluene and xylene. Typical features of these liquid organic materials are volatility, combustibility and toxicity. These are the reason why special attention must be paid to the management of liquid organic radioactive wastes. To select an appropriate waste management strategy and to design the treatment process of spent scintillation cocktails, it is required to investigate the nature of the waste such as specific radioactivity and moisture content. The analysis results of spent scintillation liquid generated at Wolsong nuclear power plants will be discussed. An overview of the technical approaches available for the treatment of organic radioactive waste will be additionally provided.

Concentrated effluent and spent ion exchange resins (IERs) from nuclear power plants (NPPs) were generated prior to the establishment of a disposal facility site and waste acceptance criteria have been temporarily stored at the NPPs because their suitability for disposal has not been confirmed. In particular, at the Kori Unit 1, which was the first to start the commercial operation in South Korea, the initially generated concentrated effluent and IERs are repackaged in large size of concrete containers and stored without provided regulation standard. The concentrated effluent is package as cementitious form in 200 L drums and repackaged in concrete containers, case of the IERs were solidified or dehydrated and repackaged in round concrete container. In this study, we review and propose a disposal plan for concentrated effluent and IERs repackaging drums that have not been confirmed to be suitable for disposal from the first operating nuclear power plant, Kori Unit 1, 2. First, the concentrated effluent was stored in four 200 L drums respectively, and then, it was again stored in concrete container and which was poured on top using grouted concrete. Therefore, the process was required by cutting concrete container for extracting the internal drums at first. Internal radioactive waste should be crushed to the suitable waste criteria and solidified, finally disposal in to the polymer concrete high integrity container (PC-HIC). IER was repackaged and disposal in square type of 200 L concrete drums respectively covered the cap. So, extracting the internal drums should be extracted after removing the cap of external concrete container. Cement solidification drums can be crushed and re-solidified or disposed in the PC-HIC. Stored IER after dehydrated can be disposal in PC-HIC. In conclusion, the container was used as a package that repackaging the concentrated effluent and IER was separated into two different types of waste depending on the level of contamination of radioactivity, the polluted area is disposed of as radioactivity contamination or the unspoiled area will be treated as self-disposal waste.

The bentonite buffer material is a crucial component for disposing of high-level radioactive waste (HLW). Several additives have been proposed to enhance the performance of bentonite buffer materials. In this study, unconfined compression tests were conducted on bentonite mixtures as well as pure bentonite buffer material. Joomunjin and silica sands were added at a 30% ratio, and graphite was added at 3% along with bentonite. The unconfined compression strength (UCS) and elastic modulus of pure bentonite were found to be 20% to 50% higher than those of bentonite mixtures under similar dry density and water content conditions. This decrease in strength can be attributed to the reduced cross-sectional area available for bearing the applied load in the bentonitemixture. Furthermore, the 3% graphite-bentonite mixture exhibited a 10% to 30% higher UCS and elastic modulus compared to the 30% sand-bentonite mixtures. However, since the strength properties of additive-bentonite mixtures are lower than those of pure bentonite, it is essential to evaluate thermohydraulic-mechanical functional criteria when considering the use of bentonite mixtures as buffer materials.

The PRIDE scale mechanical decladder is decladding apparatus for separating and recovering fuel material and cladding hull by horizontally slitting rod-cut. In order to enhance mechanical decladdng efficiency, the main requirements were considered as follows. Decladding of the fuel rods may be performed by rotation of three circular cutting blades inserted among the rollers arranged at 120° portion. In a mechanical decladder, a slitting assembly as a unit for slitting the cladding tube may include cutting blades for slitting and rollers for guiding extrusion of the cladding tube. Rotation of the cutting blades may be caused by the fuel rods being extruded from a plurality of rollers. Slitting intervals of rod-cuts having different diameters may be controlled by adding or removing a spacing plate between the cutting blade and a ranch bolt for fixing the slitting blade to the slitting assembly. An extrusion velocity with respect to the fuel rods may be controlled by a hydraulic pressure applied to the fuel rods. A force for cutting the fuel rods may be adjusted by controlling steel plates. Forces applied to a plurality of rollers may be generated by the hydraulic cylinder. The hydraulic pressure may be controlled by hydraulic pressure controller. The PRIDE scale mechanical decladder mainly consists of auto feeding module, hydraulic cylinder module and blade module. A load cell was installed between the hydraulic cylinder and the extrusion pin to measure the decladding force and slitting velocity, and a data acquisition system capable of obtaining data by using the RSC 232 was constructed. Also, the control panel can control the forward and backward movement of the extrusion pin, the hydraulic flow rate, and the hydraulic velocity. In the mechanical decladding test, 40 pieces of simulated rod-cuts were loaded in two auto feeding basket and slit by utilizing the 3-CUT blade modules in the housing, and hulls and simulated pellets were collected in the collection container. As a result, 80 pieces of simulated rodcut (brass pellets + Zry4 tube) were slit continuously without any problem. About 35 min was required to slit 80 rod-cuts and average decladding force was 260 kg. The decladding force of the ceramic simulated rod-cuts (castable) requires 25 kg less force than the brass pellets. Therefore, it is estimated that the spent fuel rod-cut can be fully split into three pieces using the mechanical decladder.