Throughfall (TF)—as a diffusive hydrological water flux—significantly affects ecohydrological and biogeochemical processes within forest ecosystems. Recent investigations have revealed the impact on TF generation processes within unmanaged coniferous plantations of under-canopy structures, particularly those laden with dead branches, as well as upper-canopy structures. However, spatiotemporal variations in TF in such plantations remain unexplored. We investigated these variations in TF in a 33-year-old unmanaged Japanese cypress (Chamaecyparis obtusa Endl.) plantation, laden with dead branches, with a high stand density (SD) of 2,500 stems ha−1. Over a two-year period (May 2017 to May 2019), we conducted weekly TF measurements using 28 manual-type TF collectors. We compared the present TF ratio and canopy water storage capacity (S) with those of previous investigations conducted on Japanese cypress plantations. Moreover, we assessed key indices contributing to spatiotemporal TF variations (canopy cover: CC and distance to the nearest stem: TFd) and potentially influential dead branch indices (number of dead branches: TFdb and vertical spacing length on a stem: TFs) to elucidate TF spatial patterns. The results showed that the TF ratio was notably lower than that in previous studies (n = 13), with SD (r = –0.92, p < 0.001) and S (r = –0.87, p < 0.001) emerging as key influential factors among other stand-structure parameters. Spatial TF patterns exhibited a decreasing trend as the gross rainfall (GR) increased. Temporal stability was not significantly associated with CC (r = 0.120, p = 0.544), TFd (r = 0.068, p = 0.731), TFdb (r = 0.211, p = 0.281), or TFs (r = 0.206, p = 0.292) for any of the TF collectors. These findings underscore the important role of GR in determining the spatial variation of TF. Collectively, our results contribute to an enhanced understanding of TF spatiotemporal heterogeneity in unmanaged Japanese cypress plantations with dead branches.

The demand for transportation is increasing due to the continuous generation of radioactive wastes. Especially, considering the geographical characteristics of Korea and the location characteristics of nuclear facilities, the demand for maritime transportation is expected to increase. If a sinking accident happens during maritime transportation, radioactive materials can be released into the ocean from radioactive waste transportation containers. Radioactive materials can spread through the ocean currents and have radiological effects on humans. The effect on humans is proportional to the concentration of radioactive materials in the ocean compartment. In order to calculate the concentration of radioactive materials that constantly flow along the ocean current, it is necessary to divide the wide ocean into appropriate compartments and express the transfer processes of radioactive materials between the compartments. Accordingly, this study analyzed various ocean transfer evaluation methodologies of overseas maritime transportation risk codes. MARINRAD, POSEIDON, and LAMER codes were selected to analyze the maritime transfer evaluation methodology. MARINRAD divided the ocean into two types of compartments that water and sediment compartments. And it was assumed that radionuclides are transfered from water to water or from water to sediment. Advection, diffusion, and sedimentation were established as transfer process for radionuclides between compartments. MARINRAD use transfer parameters to evaluate transer processes by advection, diffusion, and sedimentation. Transfer parameters were affected by flow rate, sedimentation rate, sediment porosity, and etc. POSEIDON also divided the ocean into two types that water and sediment compartment, each compartments was detaily divided into three vertical sub-compartment. Advection, diffusion, resuspension, sedimentation, and bioturbation were established as transport processes for radionuclides between compartments. POSEIDON also used transfer parameters for evaluating advection, diffusion, resuspension, sedimentation, and bioturbation. Transfer parameters were affected by suspended sediment rates, sedimentation rates, vertical diffusion coefficients, bioturbation factors, porosity, and etc. LAMER only considered the water compartment. It divided the water compartment into vertical detailed compartments. Diffusion, advection and sedimentation were established as the nuclide transfer processes between the compartments. To evaluated the transfer processes of nuclides for diffusion and advection, LAMER calculated the probability with generating random position vectors for radionuclides’ locations rather than deterministic methods such as MARINRAD’s transfer parameters or POSEIDON’s transfer rates to evaluate transfer processes. The results of this study can be used as a basis for developing radioactive materials’ ocean transfer evaluation model.

The occurrence of shear failure in a rock mass, resulting from the sliding of joint surfaces, is primarily influenced by the surface roughness and contact area of these joints. Furthermore, since joints serve as crucial conduits for the movement of water, oil, gas, and thermal energy, the aperture and geometric complexity of these joints have a significant impact on the hydraulic properties of the rock mass. This renders them critical factors in related industries. Therefore, to gain insights into the mechanical and hydraulic behavior of a rock mass, it is essential to identify the key morphological characteristics of the joints mentioned above. In this study, we quantified the morphological characteristics of tensile fractures in granitic rocks using X-ray CT imaging. To accomplish this, we prepared a cylindrical sample of Hwang-Deung granite and conducted splitting tests to artificially create tensile fractures that closely resemble rough joint surfaces. Subsequently, we obtained 2D sliced X-ray CT images of the fractured sample with a pixel resolution of approximately 0.06 mm. By analyzing the differences in CT numbers of the rock components (e.g., fractures, voids, and rock matrix), we isolated and reconstructed the geometric information of the tensile fracture in three dimensions. Finally, we derived morphological characteristics, including surface roughness, contact area, aperture, and fracture volume, from the reconstructed fracture.

This study was conducted to calculate the damage of Italian ryegrass (IRG) by abnormal climate using machine learning and present the damage through the map. The IRG data collected 1,384. The climate data was collected from the Korea Meteorological Administration Meteorological data open portal.The machine learning model called xDeepFM was used to detect IRG damage. The damage was calculated using climate data from the Automated Synoptic Observing System (95 sites) by machine learning. The calculation of damage was the difference between the Dry matter yield (DMY)normal and DMYabnormal. The normal climate was set as the 40-year of climate data according to the year of IRG data (1986~2020). The level of abnormal climate was set as a multiple of the standard deviation applying the World Meteorological Organization (WMO) standard. The DMYnormal was ranged from 5,678 to 15,188 kg/ha. The damage of IRG differed according to region and level of abnormal climate with abnormal temperature, precipitation, and wind speed from -1,380 to 1,176, -3 to 2,465, and -830 to 962 kg/ha, respectively. The maximum damage was 1,176 kg/ha when the abnormal temperature was -2 level (+1.04℃), 2,465 kg/ha when the abnormal precipitation was all level and 962 kg/ha when the abnormal wind speed was -2 level (+1.60 ㎧). The damage calculated through the WMO method was presented as an map using QGIS. There was some blank area because there was no climate data. In order to calculate the damage of blank area, it would be possible to use the automatic weather system (AWS), which provides data from more sites than the automated synoptic observing system (ASOS).

Natural radionuclides-containing substances (NORM) contain natural radionuclides and cause radiation exposure. In Korea, safety management measures were needed to deal with and dispose of radon mattresses containing monazite in relation to such NORM. However, there is no clear safety management system related to NORM waste in Korea. In order to manage this reasonably and systematically, it is necessary to investigate and analyze standards and management measures related to the treatment and disposal of NORM waste. Therefore, this study investigated and analyzed the exemption and clearance level of NORM waste regulations in international organizations and foreign countries. IAEA GSR Part 3, 2013/59/Euratom, ANSI/HPS N13.53, CRCPD SSRCR Part N, and ARPANSA Publications 15 safety management regulations were analyzed to investigate safety management standards for NORM waste. The exemption and clearance level in international organizations and foreign countries were compared and analyzed based on radioactive concentration and dose. In addition, the management measures proposed for each literature were also investigated. As a result of the analysis, IAEA GSR Part 3 applied 1 mSv as a regulatory exemption level, 1 Bq/g for uranium and thorium series as a clearance level, and 10 Bq/g for K-40 nuclides. The IAEA recommends a differential approach to the potential and scale of exposure. The EU applied 1 Bq/g to uranium and thorium families and 10 Bq/g to K-40 nuclides for both regulatory exemption and clearance levels. The EU recommended that it be managed in proportion to the scale and likelihood of exposure as a result of the action. It is analyzed that this is similar to the IAEA’s management plan. In the United States, there was no single federal government radioactive concentration and dose for NORM management. The management plan differed in management status and level from state to state, and K-40 was excluded from regulation unless it was intentionally enriched. In the case of Australia, the radioactive concentration of uranium and thorium was 1 Bq/g as a standard for regulatory exemption and 1 mSv as a dose. As a management plan, it was suggested to dispose of waste by means of accumulation, dilution/dispersion, and reclamation. It was also suggested that the scale of exposure, like international organizations, take into account the possibility. The results of this study are believed to be used as basic data for presenting domestic NORM waste treatment and disposal methods in the future.

Natural uranium-contaminated soil in Korea Atomic Energy Research Institute (KAERI) was generated by decommissioning of the natural uranium conversion facility in 2010. Some of the contaminated soil was expected to be clearance level, however the disposal cost burden is increasing because it is not classified in advance. In this study, pre-classification method is presented according to the ratio of naturally occurring radioactive material (NORM) and contaminated uranium in the soil. To verify the validity of the method, the verification of the uranium radioactivity concentration estimation method through γ-ray analysis results corrected by self-absorption using MCNP6.2, and the validity of the pre-classification method according to the net peak area ratio were evaluated. Estimating concentration for 238U and 235U with γ-ray analysis using HPGe (GC3018) and MCNP6.2 was verified by 􀟙-spectrometry. The analysis results of different methods were within the deviation range. Clearance screening factors (CSFs) were derived through MCNP6.2, and net peak area ratio were calculated at 295.21 keV, 351.92 keV(214Pb), 609.31 keV, 1120.28 keV, 1764.49 keV(214Bi) of to the 92.59 keV. CSFs for contaminated soil and natural soil were compared with U/Pb ratio. CSFs and radioactivity concentrations were measured, and the deviation from the 60 minute measurement results was compared in natural soil. Pre-classification is possible using by CSFs measured for more than 5 minutes to the average concentration of 214Pb or 214Bi in contaminated soil. In this study, the pre-classification method of clearance determination in contaminated soil was evaluated, and it was relatively accurate in a shorter measurement time than the method using the concentrations. This method is expected to be used as a simple pre-classification method through additional research.

Oral squamous cell carcinoma (OSCC), which accounts for approximately 90% of oral cancers, has a high rate of local recurrence and a poor prognosis despite improvements in treatment. Exosomes released from OSCC cells promote cell proliferation and metastasis. Although it is clear that the biogenesis of exosomes is mediated by the endosomal sorting complex required for transport (ESCRT) machinery, the gene expression pattern of ESCRT, depending on the cell type, remains elusive. The exosomal release from the human OSCC cell lines, HSC-3 and HSC-4, and their corresponding gefitinib-resistant sub-cell lines, HSC-3/GR and HSC-4/GR, was assessed by western blot and flow cytometry. The levels of ESCRT machinery proteins, including Hrs, Tsg101, and Alix, and whole-cell ubiquitination were evaluated by western blot. We observed that the basal level of exosomal release was higher in HSC-3/GR and HSC-4/GR cells than in HSC-3 and HSC-4 cells, respectively. Long-term gefitinib exposure of each cell line and its corresponding gefitinib-resistant sub-cell line differentially induced the expression of the ESCRT machinery. Furthermore, whole-cell ubiquitination and autophagic flux were shown to be increased in gefitinib-treated HSC-3 and HSC-4 cells. Our data indicate that the expression patterns of the ESCRT machinery genes are differentially regulated by the characteristics of cells, such as intracellular energy metabolism. Therefore, the expression patterns of the ESCRT machinery should be considered as a key factor to improve the treatment strategy for OSCC.

High-entropy alloys (HEAs) are attracting attention because of their excellent properties and functions; however, they are relatively expensive compared with commercial alloys. Therefore, various efforts have been made to reduce the cost of raw materials. In this study, MIM is attempted using coarse equiatomic CoCrFeMnNi HEA powders. The mixing ratio (powder:binder) for HEA feedstock preparation is explored using torque rheometer. The block-shaped green parts are fabricated through a metal injection molding process using feedstock. The thermal debinding conditions are explored by thermogravimetric analysis, and solvent and thermal debinding are performed. It is densified under various sintering conditions considering the melting point of the HEA. The final product, which contains a small amount of non-FCC phase, is manufactured at a sintering temperature of 1250oC.

In general, after the decommissioning of nuclear facilities, buildings on the site can be demolished or reused. The NSSC (Nuclear Safety and Security Commission) Notice No. 2021-11 suggests that when reusing the building on the decommissioning site, a safety assessment should be performed to confirm the effect of residual radioactivity. However, in Korea, there are currently no decommissioning experiences of nuclear power plants, and the experiences of building reuse safety assessment are also insufficient. Therefore, in this study, we analyzed the foreign cases of building reuse safety assessment after decommissioning of nuclear facilities. In this study, we investigated the Yankee Rowe nuclear power plant, Rancho Seco nuclear power plant, and Hematite fuel cycle facility. For each case, the source term, exposure scenario, exposure pathway, input parameter, and building DCGLs were analyzed. In the case of source term, each facility selected 9~26 radionuclides according to the characteristics of facilities. In the case of exposure scenario, building occupancy scenario which individuals occupy in reusing buildings was selected for all cases. Additionally, Rancho Seco also selected building renovation scenario for maintenance of building. All facilities selected 5 exposure pathways, 1) external exposure directly from a source, 2) external exposure by air submersion, 3) external exposure by deposited on the floor and wall, 4) internal exposure by inhalation, and 5) internal exposure by inadvertent ingestion. For the assessment, we used RESRAD-BUILD code for deriving building DCGLs. Input parameters are classified into building parameter, receptor parameter, and source parameter. Building parameter includes compartment height and area, receptor parameter includes indoor occupancy fraction, ingestion rate, and inhalation rate, and source parameter includes source thickness and density. The input parameters were differently selected according to the characteristics of each nuclear facility. Finally, they derived building DCGLs based on the selected source term, exposure scenario, exposure pathway, and input parameters. As a result, it was found that the maximum DCGL was 1.40×108 dpm/100 cm2, 1.30×107 dpm/100 cm2, and 1.41×109 dpm/100 cm2 for Yankee Rowe nuclear power plant, Rancho Seco nuclear power plant, and Hematite fuel cycle facility, respectively. In this study, we investigated the case of building reuse safety assessment after decommissioning of the Yankee Rowe nuclear power Plant, Rancho Seco nuclear power plant, and Hematite fuel cycle facility. Source terms, exposure scenarios, exposure pathways, input parameters, and building DCGLs were analyzed, and they were found to be different depending on the characteristics of the building. This study is expected to be used in the future building reuse safety assessment after decommissioning of domestic nuclear power plants. This work was

For safe management of spent nuclear fuels, they should be delivered to repository or waste disposal site. As the amount of spent nuclear fuel transportation is expected to increase in the future due to the provision of an intermediate storage facility, the necessity to secure transportation cask is emerging. In order to secure the spent nuclear fuel transportation cask, it is necessary to analyze the regulatory processes for domestic and foreign spent nuclear fuel transportation cask. In this study, the regulatory processes for domestic and foreign spent nuclear fuel transportation cask was analyzed. In this study, the IAEA, US, and Korea spent nuclear fuel transportation cask regulatory processes were analyzed. The domestic and foreign spent nuclear fuel transportation cask regulatory processes consist of design phase, manufacturing phase, and operation phase. In the design stage, the transport requirements are designed in accordance with the safety requirements of international organizations and countries. The application to be submitted when applying for approval should include a safety analysis report, evidence proving compliance with safety requirements et al. In the manufacturing stage, it is a stage to check whether the safety requirements are satisfied before the first use after manufacturing the transportation cask. Inspections include welding inspection, leakage inspection, shielding inspection, and thermal inspection. In the operation stage, it is a stage of periodically performing inspections for continuous maintenance of the package when the transportation cask is used. The inspection items to be performed are similar to the manufacturing stage and typically include performance inspection of components and leakage inspection. In this study, domestic and foreign spent nuclear fuel transportation cask regulatory processes were analyzed. It was found that the domestic and foreign spent nuclear fuel transportation cask regulatory processes consist of the design phase, the manufacturing phase, and the operation phase. The results of this study can be used as basic data for policy decision-making for the spent nuclear fuel cask.

The aim of this study was to investigate the effect of isolated lactic acid bacteria (LAB) on the quality of high moisture rye silage. Rye forage (Secale cereale L.) was harvested at the heading stage (27.3% of dry matter (DM)) and cut into approximately 3-5 cm lengths. Then, the forage divided into 4 treatments with different inoculants: 1) No additives (CON); 2) Lactobacillus brevis strain 100D8 at a 1.2 x 105 colony-forming unit (cfu)/g of fresh forage (LBR); 3) Leuconostoc holzapfelii strain 5H4 at a 1.0 x 105 cfu/g of fresh forage (LHO); and 4) Mixture of LBR and LHO (1:1 ratio) applied at a 1.0 x 105 cfu/g of fresh forage (MIX). About 3 kg of forage from each treatment was ensiled into a 20 L mini-bucket silo in quadruplicate for 100 days. After silo opening, silage was collected for analyses of chemical compositions, in vitro nutrient digestibilities, fermentation characteristics, and microbial enumerations. The CON silage had the highest concentrations of neutral detergent fiber and acid detergent fiber (p = 0.006; p = 0.008) and a lowest in vitro DM digestibility (p < 0.001). The pH was highest in CON silage, while lowest in LBR and MIX silages (p < 0.001). The concentrations of ammonia-N, lactate, and acetate were highest in LBR silage (p = 0.008; p < 0.001; p < 0.001). Propionate and butyrate concentrations were highest in CON silage (p = 0.004; p < 0.001). The LAB and yeast counts were higher in CON and LHO silages compare to LBR and MIX silages (p < 0.001). However, the mold did not detect in all treatments. Therefore, this study could conclude that L. brevis 100D8 and Leu. holzapfelii strain 5H4 can improve the digestibility and anti-fungal activity of high moisture rye silage.