본 연구는 소양강댐 하류에서 서식하는 생태계교란 생물 종인 브라운송어와 그 먹이원으로 이용되는 저서성 대형무척추동물에 대한 파악을 위해 2022년부터 2023년까지 총 8회에 걸쳐 소양강댐 하류(St.1~St.3)와 지류 (St.4)에 대해 브라운송어와 공서종, 브라운송어의 위 내용물, 저서성 대형무척추동물의 종조성 및 기능군 분석을 실시하였다. 저서성 대형무척추동물의 경우, 하루살이목에서 가장 많은 분류군이 확인되었으며(27.1%), 그 중 붙는 무리(CL)와 헤엄치는 무리(SW)가 높은 비율을 차지하는 것으로 확인되었다. 브라운송어 채집 결과, 전장은 26∼246mm까지 총 105개체가 채집되었으며, 전장-체중 관계의 매개변수 b값이 3을 초과하여 안정적인 성장이 이루어지는 것으로 확인되었다. 브라운송어의 위 내용물에 대한 먹이원 분석 결과, 빙어(0.2%, TL: 246mm)와 육상곤충(2.7%, TL: 154mm, 183∼185mm)을 섭식한 개체는 매우 적었으며 상대적으로 전장이 큰 개체에서 확인 되었다. 대부분 수서곤충(73.8%)과 물 속에서 서식하는 비곤충류(23.3%)를 섭식하는 것으로 나타났다. 브라운 송어의 전장에 따른 먹이 섭식 패턴을 파악하기 위해 위 내용물에서 확인된 종들과의 상관분석을 실시한 결과, 브라운송어의 먹이원 중 유수성 환경 선호 종들의 경우 전장과 양의 상관관계(p<0.05)를 나타낸 반면, 모래 기질 이하의 흐름이 적은 서식처를 선호하는 종들의 경우 전장과 음의 상관관계(p<0.05)를 나타냈다.

M. pruinosa has been found on 98 families and 345 species of plants, indicating a wide host range. Since its first report in a persimmon orchard in Gimhae in 2009, it had spread to 126 municipalities over 12,429 ha. Because of powerful dispersal ability, it’s difficult to control with insecticides. Therefore, it is necessary to reduce the population density through the release of M. pruinosa’s natural enemy, N. typhlocybae. N. typhlocybae females prey on or externally parasitize the nymphs of M. pruinosa. This natural enemy was introduced from Italy in 2017 through an international cooperative project between National Institute of Agricultural Sciences and University of Padova for controlling the sporadic pest, Metcalfa pruinosa. This study reported the results of mass rearing N. typhlocybae indoors and releasing them outdoors from 2020 to 2023 in 21 regions nationwide, showing an average parasitism rate of 27.3% and an average production of 7.9 cocoons per host, suggesting the possibility of their establishment in Korea.

Velvet antler is widely used as a traditional medicine, and numerous studies have demonstrated its tremendous nutritional and medicinal values including immunity-enhancing effects. This study aimed to investigate different deer velvet extracts (Sample 1: raw extract, Sample 2: dried extract, and Sample 3: freeze-dried extract) for proximate composition, uronic acid, sulfated glycosaminoglycan, sialic acid, collagen levels, and chemical components using ultra-performance liquid chromatography-quadrupole-time-of-light mass spectrometry. In addition, we evaluated the cytotoxic effect of the deer velvet extracts on BV2 microglia, HT22 hippocampal cells, HaCaT keratinocytes, and RAW264.7 macrophages using the cell viability MTT assay. Furthermore, we evaluated acute toxicity of the deer velvet extracts at different doses (0, 500, 1000, and 2000 mg/kg) administered orally to both male and female ICR mice for 14 d (five mice per group). After treatment, we evaluated general toxicity, survival rate, body weight changes, mortality, clinical signs, and necropsy findings in the experimental mice based on OECD guidelines. The results suggested that in vitro treatment with the evaluated extracts had no cytotoxic effect in HaCaT keratinocytes cells, whereas Sample-2 had a cytotoxic effect at 500 and 1000 μg/mL on HT22 hippocampal cells and RAW264.7 macrophages. Sample 3 was also cytotoxic at concentrations of 500 and 1000 μg/mL to RAW264.7 and BV2 microglial cells. However, the mice treated in vivo with the velvet extracts at doses of 500–2000 mg/kg BW showed no clinical signs, mortality, or necropsy findings, indicating that the LD50 is higher than this dosage. These findings indicate that there were no toxicological abnormalities connected with the deer velvet extract treatment in mice. However, further human and animal studies are needed before sufficient safety information is available to justify its use in humans.

Most of the radioactive wastes generated during the nuclear fuel processing activities conducted by KEPCO Nuclear Fuel Co., Ltd. are classified as the categories of intermediate and low-level radioactive waste. These radioactive waste materials are intended for permanent disposal at a designated disposal site, adhering strictly to the waste acceptance criteria. To facilitate the safe transportation of radioactive waste to the disposal site, it is necessary to ensure that the waste drums maintain a level of criticality that complies with the waste acceptance criteria. This necessitates the maintenance of subcritical conditions, under immersion or optimal neutron moderation conditions. This paper presents a criticality safety assessment of concrete radioactive waste under the most conservative conditions of immersion and moderation conditions for waste drums. Specifically, In order to send radioactive waste, which is the subject of criticality analysis, to a disposal facility, pre-processing operations must be performed to ensure compliance with waste accepatance criteria. To meet the physical characteristics required by the accepance criteria, particles below 0.2 mm should not be included. Thus, a 0.3 mm sieve is used to separate particles lager than 0.3 mm, and only those particles are placed in drums. The drums should be filled to achieve a filling ratio of at least 85%. A criticality analysis was conducted using the KENO-VI of SCALE. The Criticality Safety Analysis Results of varying the filling ratio of concrete drums from 85% to 100% presented in an effective multiplication factor of 0.22484. Additionally, the effective multiplication factor presented to be 0.25384 under the optimal moderation conditions. This demonstrates full compliance with the USL and criticality technology standards set as 0.95.

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.

2022년 10월부터 2023년 5월까지 친환경 딸기재배 농가에서 점박이응애는 칠레이리애응애와 사막이리응애, 진딧물은 콜레마니진디벌을 대상으로 하는 천적처리구와 유기농자재를 사용하는 관행방제구에서 천적의 해 충 밀도억제 효과를 조사하였다. 천적처리구에서 점박이응애 성충 밀도는 잎당 1.5마리 이하, 알은 4개 이하로 관리되었고, 사막이리응애는 잎당 최대 0.4마리까지 증가하여 점박이응애 밀도억제에 많은 영향을 미친것으로 보인다. 반면, 관행방제구에서 점박이응애 밀도는 천적 방사구에 비하여 오히려 많은 발생량을 보였지만, 3월 9일부터는 사막이리응애의 증가와 유기농자재의 효과로 점박이응애 밀도는 급격하게 감소하였다. 진딧물 천적 처리구에서 진딧물 밀도는 1월부터 발생하였으며 3월 상순에 잎당 0.3마리까지 증가하였으나 이후 감소하였고, 콜레마니진디벌은 진딧물 발생이 많지 않아 3.9마리/㎡(2회) 방사하는 데 그쳤다. 한편, 관행방제구 포장의 진딧 물은 유기농자재의 영향으로 거의 발생하지 않았다.

Mass production of high-quality carbon nanotubes (CNTs) remains a challenge, requiring the development of new wetimpregnated catalyst suitable for the catalytic chemical vapor deposition (CCVD) of CNTs in a fluidized bed reactor. For the successful development of a new catalyst, a highly robust system to synthesize CNTs must be established. Here, we systematically investigated the robustness of CNT synthesis by CCVD using a wet-impregnated catalyst. We statistically tested four factors that could potentially affect the robustness of CNT synthesis system, focusing on carbon yield and IG/ID. First, we tested the effect of vacuum baking before CNT growth. F test and CV equality test concluded that vacuum baking recipe did not significantly reduce the variability of the CNT synthesis. Second, we tested the batch-to-batch variation of catalysts. The results of t test and one-way analysis of variance indicate that there is significant difference in carbon yield and IG/ID among catalysts from different batches. Third, we confirmed that there is spatial non-uniformity of wet-impregnated catalysts within a batch when they are produced in large scale. Finally, we developed a multi-step heating recipe to mitigate the temperature overshooting during the CNT synthesis. The multi-step recipe increased the mean of carbon yield, but did not influence the variability of CNT synthesis. We believe that our research can contribute to the establishment of a robust CNT synthesis system and development of new wet-impregnated catalysts.

Removing CO2 gas to address the global climate crisis is one of the most urgent agendas. To improve the CO2 adsorption ability of activated carbon, nitrogen plasma surface treatment was conducted. The effect of nitrogen plasma treatment on the surface chemistry and pore geometry of activated carbon was extensively analyzed. The porosity and surface groups of the activated carbon varied with the plasma treatment time. By plasma treatment for a few minutes, the microporosity and surface functionality could be simultaneously controlled. The changed microporosity and nitrogen groups affected the CO2 adsorption capacity and CO2 adsorption selectivity over N2. This simultaneous surface etching and functionalization effect could be achieved with a short operating time and low energy consumption.

According to the Nuclear Safety and Security Commission (NSSC) Notice No. 2021-26 “Delivery Regulations for the Low- and Intermediate Level Radioactive Waste (LILW)”, the activity of 3H, 14C, 55Fe, 58Co, 60Co, 59Ni, 63Ni, 90Sr, 94Nb, 99Tc, 129I, 137Cs, 144Ce, and gross alpha must be identified. Currently, the scaling factor of the dry active waste (DAW) for LILW is applied as an indirect evaluation method in Korea. The analyses are used the destructive methods and 55Fe, 60Co, 59Ni, 63Ni, 90Sr, 94Nb, 99Tc, and 137Cs, which are classified as nonvolatile nuclides, are separated through sequential separation and then measured by gamma detector, liquid scintillation counter (LSC), alpha/beta total counter (Gas Proportional Counter, GPC), and ICP-MS. We will introduce how to apply the existing nuclide separation method and improve the measurement method to supplement it.

Radiation dose rates for spent fuel storage casks and storage facilities of them are typically calculated using Monte Carlo calculation codes. In particular, Monte Carlo computer code has the advantage of being able to analyze radiation transport very similar to the actual situation and accurately simulate complex structures. However, to evaluate the radiation dose rate for models such as ISFSI (Independent Spent Fuel Storage Installation) with a lot of spent fuel storage casks using Monte Carlo computational techniques has a disadvantage that it takes considerable computational time. This is because the radiation dose rate from the cask located at the outermost part of the storage facility to hundreds of meters must be calculated. In addition, if a building is considered in addition to many storage casks, more analysis time is required. Therefore, it is necessary to improve the efficiency of the computational techniques in order to evaluate the radiation dose rate for the ISFSI using Monte Carlo computational codes. The radiation dose rate evaluation of storage facilities using evaluation techniques for improving calculation efficiency is performed in the following steps. (1) simplified change in detailed analysis model for single storage cask, (2) create source term for the outermost side and top surface of the storage cask, (3) full modeling for storage facilities using casks with surface sources, (4) evaluation of radiation dose rate by distance corresponding to the dose rate limit. Using this calculation method, the dose rate according to the distance was evaluated by assuming that the concrete storage cask (KORAD21C) and the horizontal storage module (NUHOMS-HSM) were stored in the storage facility. As a result of calculation, the distance to boundary of the radiation control area and restricted area of the storage facility is respectively 75 m / 530 m (KORAD21C case), and 20 m / 350 m (NUHOMS-HSM case).

The facile production of high-purity mesophase pitch has been a long-standing desire in various carbon industries. Recently, polymer additives for mesophase production have attracted much attention because of their convenience and efficiency. We propose polyvinylidene fluoride (PVDF) as a strong candidate as an effective additive for mesophase production. The mesophase content and structural, chemical, and thermal properties of pitches obtained with different amounts of added PVDF are discussed. The influence of PVDF decomposition on mesophase formation is also discussed. We believe that this work provides an effective option for mesophase pitch production.

The decommissioning of nuclear power plant (NPP) generates large amount of waste. Since the most of the concretes are slightly surface contaminated, the accurate characterization and regionspecific surface decontamination are important for the efficient waste management. After the effective surface decontamination and separation, most of the concrete waste from decommissioning of NPP can be classified as a clearance waste. Various surface characterization and decontamination technologies are suggested. The mechanical technologies are simple and offers direct application. The laser-based technologies offer efficient separation and surface contamination. The high price, however, hesitates the application of the process. The nitro-jet technology, which is based on the evaporation of liquid nitrogen, allows the effective decontamination. However, the high price and uncertainty of large are application hinders the practical application in NPP decommissioning. In this paper, various technologies for characterization, handling, treatment, etc., will be discussed. The advantages and disadvantages of the technologies will be discussed, in terms of practical applications.

Molten salt immersion technique has been tested with several Sr oxides, SrZrO3, SrMoO4 and U2SrOy, and MgCl2 based molten salts for the Sr nuclide separation. Reaction time, temperature, and salt composition were varied to effectively separate Sr in chloride forms. ICP-OES, XRD, and SEM analysis were conducted for the conversion efficiency and structure and morphology analysis. It is confirmed that all experiments of SrZrO3 with MgCl2 at 800°C for reaction time 5, 10, 20 hours showed higher conversion efficiency than 99% and in LiCl-KCl-MgCl2 and NaCl-MgCl2 molten salts at 500°C or 600°C, conversion efficiency higher than 97% was obtained. SrMoO4 in MgCl2 immersion experiments for 10 hours showed higher conversion efficiency than 99% when the molar ratio of salt/oxide powder is 7. U2SrOy was also tested with MgCl2 molten salt at 800°C and higher efficiency than 99% and mainly MgUO4 were produced as a reaction product.