애긴노린재는 긴노린재과에 속하며 한국을 포함한 동아시아 국가의 다양한 곡물 및 관상용 식물의 주요 해충으로 여겨진다. 본 연구에서는 애긴노린재의 17,367 bp 미토콘드리아 유전체에서 13개의 protein-coding genes, 22개의 transfer RNA genes, 2개의 ribosomal RNA genes 과 non-coding A+T rich region를 확인하였다. G+C content는 23%로 나타났고 다른 긴노린재과와의 염기서열 유사성이 N. cymoides (94.5%), N. fuscovittatus (91.7%)으로 높은 것을 발견하였다. 애긴노린재의 미토콘드리아 유전체 정보는 향후 긴노린재과의 진화 연구와 해충 방제를 위한 정보로 널리 사용될 수 있다.

Biomass-derived porous carbon is an excellent scientific and technologically interesting material for supercapacitor applications. In this study, we developed biomass-derived nitrogen-doped porous carbon nanosheets (BDPCNS) from cedar cone biomass using a simple KOH activation and pyrolysis method. The BDPCNS was effectively modified at different temperatures of 600 °C, 700 °C, and 800 ℃ under similar conditions. The as-prepared BDPCNS-700 electrode exhibited a high BET surface area of 2883 m2 g− 1 and a total pore volume of 1.26 cm3 g− 1. Additionally, BDPCNS-700 had the highest electrical conductivity (11.03 cm− 1) and highest N-doped content among the different electrode materials. The BDPCNS-700 electrode attained a specific capacitance of 290 F g− 1 at a current density of 1 A g− 1 in a 3 M KOH electrolyte and an excellent longterm electrochemical cycling stability of 93.4% over 1000 cycles. Moreover, the BDPCNS-700 electrode had an excellent energy density (40.27 Wh kg− 1) vs power density (208.19 W kg− 1). These findings indicate that BDPCNS with large surface areas are promising electrode materials for supercapacitors and energy storage systems.

In this study, we have fabricated the phenolic resin (PR)/polyacrylonitrile (PAN) blend-derived core-sheath nanostructured carbon nanofibers (CNFs) via one-pot solution electrospinning. The obtained core-sheath nanostructured carbon nanofibers were further treated by mixed salt activation process to develop the activated porous CNFs (CNF-A). Compared to pure PAN-based CNFs, the activated PR/PAN blend with PR 20% (CNF28-A)-derived core-sheath nanostructured CNFs showed enhanced specific capacitance of ~ 223 F g− 1 under a three-electrode configuration. Besides, the assembled symmetric CNF28-A//CNF28-A device possessed a specific capacitance of 76.7 F g− 1 at a current density of 1 A g− 1 and exhibited good stability of 111% after 5,000 galvanostatic charge/discharge (GCD) cycles, which verifies the outstanding long-term cycle stability of the device. Moreover, the fabricated supercapacitor device delivered an energy density of 8.63 Wh kg− 1 at a power density of 450 W kg− 1.

This study was performed to evaluate the separation of Sr, Cs, Ba, La, Ce, and Nd using gas pressurized extraction chromatography (GPEC) with anion exchange resin for the quantitation of Neodymium. GPEC is a micro-scaled column chromatography system that provides a constant flow rate by utilizing nitrogen gas. It is overcome the disadvantages of conventional column chromatography by reducing the volume of elution solvent and shortening the analysis time. Here, we compared the conventional column chromatography and the GPEC method. The whole analysis time was decreased by nine times and radioactive wastes were reduced by five times using the GPEC system. Anion exchange resin 1-X4 (200~400 mesh size) was used. The sample was prepared at a 0.8 M nitric acid in methanol solution. The elution solvent was used at a 0.01 M nitric acid in methanol solution. Finally the eluate was analyzed by ICP-MS to determine the identification and recovery. In this case, we applied the natural isotopes of LREEs (139La, 140Ce, and 144Nd) and high activity nuclides (88Sr, 133Cs, and 138Ba) instead of radioactive isotopes for the preliminary test; as a result, unnecessary radioactive waste was not produced. The recoveries were 93.9%, 105.9%, 91.9%, 47.6%, 35.9%, and 79.9% of Sr, Cs, Ba, La, Ce, and Nd, respectively. The reproducibility of recoveries by GPEC were in the range 2.8%–10.9%.

The density of molten salts is the most important property in the development of molten salt reactor (MSR). The density value measured through the experiment is also very valuable as a gold standard for the validation of the prediction models based on molecular dynamics or other computational methods. To the best of our knowledge, the experimental density data of the ternary NaCl-MgCl2- UCl3 salt system as a MSR candidate fuel salt have never been reported previously. In this study, density measurement experiment of high-temperature molten salt of NaCl-MgCl2 and NaCl-MgCl2- UCl3 was conducted using a previously-developed density measurement system based on the maximum bubble pressure (MPB) method. As a result of the experiment, the density value of 62NaCl- 18MgCl2-20UCl3 molten salt at 873 K was 2.62 g/cm3. A density prediction value of 2.65 g/cm3 at 873 K was derived from the obtained results based on the rule of additivity of molar volume method. The predictred density of 62NaCl-18MgCl2-20UCl3 was consistent with the experimental value within 1%. The density measuring system used in this study is promising for the validation of other multicomponent molten salt systems.

Viscosity of molten salts is an essential property for the thermal hydraulic design and evaluation of molten salt reactor (MSR). Therefore, viscosity data is one of the fundamental physical property data required for safe process operation and countermeasures to severe accidents. In this study, based on our experience of developing a viscosity measurement system for high-temperature LiCl-KCl molten salt system, the viscosity of NaCl-MgCl2 and NaCl-MgCl2-UCl3 molten salts, which are considered promising salts in MSR, was measured. In order to investigate the physical properties of uranium in high-temperature NaCl-MgCl2 molten salt, a viscometer system for high-temperature viscosity measurement was specially designed. As a result of the measurement, the viscosity of the 58NaCl- 42MgCl2 molten salt was 2.73 cP at 838 K, 2.15 cP at 889 K, and 1.68 cP at 940 K. And the viscosity of 73NaCl-21MgCl2-6UCl3 molten salt was 3.79 cP at 877 K, 3.58 cP at 897 K, and 1.63 cP at 941 K. The repeatability of the measurement showed a precision of less than 3%. Although sufficientlyverified starting materials were not used, viscosity data were reported for the first time for NaCl- MgCl2-UCl3 molten salts.

Molten salts have gained significant attention as a potential medium for heat transfer or energy storage and as liquid nuclear fuel, owing to their superior thermal properties. Various fluoride- and chloride-based salts are being explored as potential liquid fuels for several types of molten salt reactors (MSRs). Among these, chloride-based salts have recently received attention in MSR development due to their high solubility in actinides, which has the potential to increase fuel burnup and reduce nuclear water production. Accurate knowledge of the thermal physical properties of molten salts, such as density, viscosity, thermal conductivity, and heat capacity, is critical for the design, licensing, and operation of MSRs. Various experimental techniques have been used to determine the thermal properties of molten salts, and more recently, computational methods such as molecular dynamics simulations have also been utilized to predict these properties. However, information on the thermal physical properties of salts containing actinides is still limited and unreliable. In this study, we analyzed the available thermal physical property database of chloride salts to develop accurate models and simulations that can predict the behavior of molten salts under various operating conditions. Furthermore, we conducted experiments to improve our understanding of the behavior of molten salts. The results of this study are expected to contribute to the development of safer and more efficient MSRs.

Tritium is a radioactive isotope of hydrogen with a half-life of about 12.3 years, and it is commonly found in the environment as a result of the production of Nuclear Power Plants. The World Health Organization (WHO) has established guidelines for the permissible levels of tritium in drinking water. The guideline value for tritium in drinking water is 10,000 Bq/L. It is important to note that the guideline value for tritium is not a legal limit, but rather a recommendation. National and local authorities may establish legal limits that are more restrictive than the WHO guideline value based on local conditions and risk assessments. The Australia and Finland have set a limit for tritium in drinking water at 76,103 Bq/L and 30,000 Bq/L respectively, which is more than three to seven times higher compare to guideline value of WHO. The United States Environmental Protection Agency (EPA) has set a maximum contaminant level (MCL) for tritium in drinking water at 20,000 picocuries per liter (pCi/L), which is equivalent to 740 Bq/L. The Health Canada has set a guideline value for tritium in drinking water at 7,000 Bq/L. Assuming drinking water corresponding to each tritium limit (or guideline value) for one year, the expected exposure dose is 0.01 mSv to 1 mSv. It means that the tritium in drinking water below the limits or guideline value does not pose a significant risk to human health.

Safety-related items in the decommissioning Nuclear Power Plants (NPPs) can largely consider safety for workers and residents. At this time, the effects of radioactive contamination on the Systems, Structures, and Components (SSCs) are caused by the performance of work related to Decontamination and Dismantlement (D&D) activities. Classification according to dismantling activities will be important, and the decay factor of radionuclides and the impact of contaminations due to plant characteristic (thermal and electrical capacity) in estimation of exposure dose from such activities will be considered compared to other overseas NPPs. Therefore, this study will consider some factors to consider for comparison with overseas cases in estimating worker exposure dose. To assess worker exposure doses, the classification of decommissioning activities must first be made. It should be classified including large components that can be generally considered, and the contents should be similar to compare with overseas cases. In case of decommissioned NPPs with prior experience, it is possible to predict worker’s exposure with respect to plant capacity, but this does not seem to have a specific correlation when reviewing the related data. Depending on the plant capacity, the occurrence of contamination of radioactive materials may have some correlation, but it cannot be determined that it has causality with the worker’s dose when dismantling. In addition, it is expected that the effects of workers’ exposure doses will vary depending on when the highly contaminated SSCs will be dismantled from permanent shut down. Therefore, the decay correlation coefficient for this high radiation dose works should be considered. If the high radiation dose work is performed before the base year, a correlation coefficient larger than 1 value will be applied, and in the opposite case, a value less than 1 will be applied. Whether or not to perform Full System Decontamination (FSD) is also an important consideration that affects worker dose, and correlation factors should be applied. In this study, the matters to be considered when estimating worker dose for dismantling NPPs were reviewed. This suggests factors to be reflected in the work classification and dose results for comparison with overseas NPP experiences. Therefore, when doing the workers’ dose estimation, it is necessary to derive a normalized doses considering each correlation factor when comparing with overseas cases along with dose estimation for the dismantling activities.