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.

When the parent radionuclide decays, the progeny radionuclide is produced. Accordingly, the dose contribution of the progeny radionuclide should be considered when assessing dose. For this purpose, European Commission (EC) and International Atomic Energy Agency (IAEA) provide weighting factors for dose coefficient. However, these weighting factors have a limitation that does not reflect the latest nuclide data. Therefore, in this study, we analyzed the EC and IAEA methodology for derivation of weighting factor and used the latest nuclide data from ICRP 107 to derive weighting factors for dose coefficient. Weighting factor calculation is carried out through 1) selection of nuclide, 2) setting of evaluation period, and 3) derivation based on ICRP 107 radionuclide data. Firstly, in order to derive the weighting factor, we need to select the radionuclides whose dose contribution should be considered. If the half-life of progeny radionuclides sufficiently short compared to the parent radionuclide to achieve radioactive equilibrium, or if the dose coefficient is greater of similar to that of the parent radionuclide and cannot be ignored, the dose contribution of the progeny radionuclide should be considered. In order not to underestimate the dose contribution of progeny radionuclides, the weighting factors for the progeny nuclides are taken as the maximum activity ratio that the respective progeny radionuclides will reach during a time span of 100 years. Finally, the weighting factor can be derived by considering the radioactivity ratio and branch fraction. In order to calculate the weighting factor, decay data such as the half-life of the radionuclide, decay chain, and branch fraction are required. In this study, radionuclide data from ICRP 107 was used. As a result of the evaluation, for most radionuclides, the weighting factors were derived similarly to the existing EC and IAEA weighting factors. However, for some nuclides, the weighting factors were significantly different from EC and IAEA. This is judged to be a difference in the half-life and branch fraction of the radionuclide. For example, in the case of 95Zr, the weighting factor for 95mNb showed a 35.8% difference between this study and previous study. For ICRP 38, when 95Zr decays, the branch fraction for 95mNb is 6.98×10-3. In contrast, for ICRP 107, the branch fraction is 1.08×10-2, a difference of 54.7%. Therefore, the weighting factor for the dose coefficient based on ICRP 107 data may differ from existing studies depending on the half-life and decay information of the nuclide. This suggests the need for a weighting factor based on the latest nuclide data. The results of this study can be used as a basis for the consideration of dose contributions for progeny radionuclides in various dose assessments.

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.

As the demand for nuclear power increases as a means to achieve carbon neutrality, concerns about nuclear proliferation have also grown. Consequently, significant researches have conducted to enhance nuclear non-proliferation resistance. Among these research, nuclear material attractiveness is a methodology used to evaluate how appealing a particular material is for potential use in nuclear weapons, based on the characteristics of that material. Existing nuclear material attractiveness assessments focused on materials like U, Pu, and TRU, which could be directly used in the production of nuclear weapons. However, these assessments did not consider how the properties of nuclear materials change throughout the nuclear fuel cycle, with each facility process. This study assumed a scenario of the nuclear fuel cycle of graphite reduction reactors and analyzed including enrichment facilities and PUREX. This study used the FOM (Figure-Of-Merit) method developed by LANL (Los Alamos National Laboratory) for evaluating the nuclear material attractiveness. The FOM formula consists of three parameters such as critical mass, heat content, and dose The critical mass of targe materials and the dose evaluation were conducted using the Monte Carlo N-Particle code. The heat content was calculated using the ORIGEN code embedded in the Scale code. In particular, if U-238 is dominant in the facility’s materials, such as mining and refining facilities, and critical mass evaluation is unpractical. Therefore, 1SQ (Significant Quantity) of that uranium was assumed as the critical mass value for the FOM evaluation, even though 1SQ is not identical to the critical mass As a result of this study, the attractiveness of Pu produced by PUREX among all nuclear fuel cycle facilities was 2.7616, which was the most attractive to be diverted to nuclear weapons. Through this study, it was shown that the proliferation risk of the nuclear facilities in the nuclear fuel cycle and risk of diversion among those facilities.

Bombyx mandarina (Lepidoptera: Bombycidae), the presumed ancestor of B. mori, has long been a subject of study to illustrate the geographic relationships in connection with origin of B. mori. We report 97 mitochondrial genome (mitogenome) sequences of B. mandarina collected from Korea and Japan. Phylogenetic and population genetic analyses showed that all individuals of B. mandarina collected in Korean localities formed a strong group together with all individuals originated from northern China (mainly north of the Qinling-Huaihe line) and some of southern China. This group was placed as the sister group to B. mori strians suggesting that this group had been served as an immediate progenitor for B. mori.

This study was aimed to isolate bacterial inoculants producing chitinase and evaluate their application effects on corn silage. Four corn silages were collected from four beef cattle farms to serve as the sources of bacterial inoculants. All isolates were tested against Fusarium graminearum head blight fungus MHGNU F132 to confirm their antifungal effects. The enzyme activities (carboxylesterase and chitinase) were also measured to isolate the bacterial inoculant. Based on the activities of anti-head blight fungus, carboxylesterase, and chitinase, L. buchneri L11-1 and L. paracasei L9-3 were subjected to silage production. Corn forage (cv. Gwangpyeongok) was ensiled into a 10 L mini silo (5 kg) in quadruplication for 90 days. A 2 × 2 factorial design consists of F. graminearum contamination at 1.0104 cfu/g (UCT (no contamination) vs. CT (contamination)) and inoculant application at 2.1 × 105 cfu/g (CON (no inoculant) vs. INO (inoculant)) used in this study. After 90 days of ensiling, the contents of CP, NDF, and ADF increased (p<0.05) by F. graminearum contamination, while IVDMD, acetate, and aerobic stability decreased (p<0.05). Meanwhile, aerobic stability decreased (p<0.05) by inoculant application. There were interaction effects (p<0.05) on IVNDFD, NH3-N, LAB, and yeast, which were highest in UCT-INO, UCT-CON, CT-INO, and CT-CON & INO, respectively. In conclusion, this study found that mold contamination could negatively impact silage quality, but isolated inoculants had limited effects on IVNDFD and yeast.

As nuclear power plants are operated in Korea, low and intermediate-level radioactive wastes and spent nuclear fuels are continuously generated. Due to the increase in the amount of radioactive waste generated, the demand for transportation of radioactive wastes in Korea is increasing. This can have radiological effect for public and worker, risk assessment for radioactive waste transportation should be preceded. Especially, if the radionuclides release in the ocean because of ship sinking accident, it can cause internal exposure by ingestion of aquatic foods. Thus, it is necessary to analyze process of internal exposure due to ingestion. The object of this study is to analyze internal exposure by ingestion of aquatic foods. In this study, we analyzed the process and the evaluation methodology of internal exposure caused by aquatic foods ingestion in MARINRAD, a risk assessment code for marine transport sinking accidents developed by the Sandia National Laboratory (SNL). To calculate the ingestion internal exposure dose, the ingestion concentrations of radionuclides caused by the food chain are calculated first. For this purpose, MARINRAD divide the food chain into three stages; prey, primary predator, and secondary predator. Marine species in each food chain are not specific but general to accommodate a wide variety of global consumer groups. The ingestion concentrations of radionuclides are expressed as an ingestion concentration factors. In the case of prey, the ingestion concentration factors apply the value derived from biological experiments. The predator's ingestion concentration factors are calculated by considering factors such as fraction of nuclide absorbed in gut, ingestion rate, etc. When calculating the ingestion internal exposure dose, the previously calculated ingestion concentration factor, consumption of aquatic food, and dose conversion factor for ingestion are considered. MARINRAD assume that humans consume all marine species presented in the food chain. Marine species consumption is assumed approximate and conservative values for generality. In the internal exposure evaluation by aquatic foods ingestion in this study, the ingestion concetration factor considering the food chain, the fraction of nuclide absorbed in predator’s gut, ingestion rate of predator, etc. were considered as influencing factors. In order to evaluate the risk of maritime transportation reflecting domestic characteristics, factors such as domestic food chains and ingestion rate should be considered. The result of this study can be used as basis for risk assessment for maritime transportation in Korea.

In Korea, the construction of dry storage facilities for spent nuclear fuel is being promoted through the 2nd basic plan for high-level radioactive waste management. When operating dry storage facilities, exposure dose assessment for workers should be performed, and for this, exposure scenarios based on work procedures should be derived prior. However, the dry storage method has not yet been sufficiently established in Korea, so the work procedure has not been established. Therefore, research is needed to apply it domestically based on the analysis of spent nuclear fuel management methods in major overseas leading countries. In this study, the procedure for receiving and storing spent nuclear fuel in a concrete overpack-based storage facility was analyzed. Among the various spent nuclear fuel management systems, the metal overpack-based HI-STAR 100 system and the concrete overpackbased HI-STORM 100 system are quite common methods in the United States. Therefore, in this study, work procedures were analyzed based on each final safety analysis report. First, the HI-STAR 100 overpack enters the facility and is placed in the transfer area. Remove the impact limiter of the overpack and install the alignment device on the top of the overpack. Place the HI-TRAC, an on-site transfer device, on top of the alignment unit and remove the lids of the two devices to insert the canister into the HI-TRAC. When the canister transfer is complete, reseat the lid to seal it, and disconnect the HI-TRAC from the HI-STAR 100. Raise the canister-loaded HI-TRAC over the alignment device on the top of the HI-STORM 100 overpack and remove the lids of the two devices that are in contact. Insert the canister into the HI-STORM 100 and reseat the lid. The HI-STORM 100 loaded with spent nuclear fuel is transferred to the designated storage area. In this study, the procedure for receiving and storing spent nuclear fuel in a concrete overpack-based storage facility was analyzed. The main procedure was the transfer of canisters between overpacks, and it was confirmed that HI-TRAC was used in the work procedure. The results of this study can be used as basic data for evaluating the exposure dose of operating workers for the construction of dry storage facilities in Korea.

Milling facilities, which belong to the front end of the nuclear fuel cycle, are essential steps for utilizing uranium in nuclear power generation. These milling facilities currently provide the International Atomic Energy Agency (IAEA) with the location and annual production capacity of the facility through the Additional Protocol (AP, INFCIRC/540) and grant IAEA inspectors on-site sampling authority to apply safeguards to the facility. However, since milling facilities process a large amount of nuclear material and the product uranium ore concentrate (UOC) is bulk material, the absence of accounting for the facility could pose a potential risk of nuclear proliferation. Therefore, this study proposes technical approach that can be utilized for safeguards in milling facilities. Since the half-life of uranium isotopes is much longer than that of its daughter, they reach the secular equilibrium condition. However, after milling process, the fresh tailing showed the break of that secular equilibrium. As time goes on, they newly reach another secular equilibrium condition. Based on this observation, this study discussed the feasibility of the ratio method in safeguards purpose. The scenario applied in this study was 1% of uranium mill tailing. It was observed that the U-238/Th-234 and U- 238/Pa-234m ratios in fresh milling tails varied as a function of time after discharging, particularly during the first one year. This change can be worked as a significant signature in terms of safeguards. In conclusion, the ratio method in mill tails could be applicable for safeguards of nuclear milling facility.

Currently, low and intermediate-level radioactive wastes and spent nuclear fuels are continuously generated in Korea. For the disposal of the radioactive wastes, the transport demand is expected to increase. Prior to transportation, it is necessary to evaluate the radiation risk of transportation to confirm that is not high. In Korea, there is no transportation risk assessment code that reflects domestic characteristics. Therefore, foreign assessment codes are used. In this study, before developing the overland transportation risk assessment code that reflects domestic characteristics, we analyzed the radiation risk assessment methodology in transportation accident codes developed in other countries. RADTRAN and RISKIND codes were selected as representative overland transportation risk assessment codes. For the two codes we analyzed accident scenarios, exposure pathways, and atmospheric diffusion. In RADTRAN, the user classifies accident severity for possible accident scenarios, and the user inputs the probability for each accident severity. On the other hand, in the case of RISKIND, the accident scenarios are classified and the probabilities are determined according to the NRC modal study (LLNL, 1987) in consideration of the cask impact velocity, cask impact angle, and fire temperature. In the case of RISKIND, the accident scenarios are applied only to transportation of spent nuclear fuel, and cannot be defined for low and intermediate-level radioactive waste. However, in the case of RADTRAN, since the severity and probability of accidents are defined by user, it can be applied to low and intermediate-level radioactive wastes. As the exposure pathways considered in transportation accident, both RADTRAN and RISKIND consider external exposure (cloudshine and groundshine), and internal exposure (inhalation, resuspension inhalation and ingestion). In the case of RADTRAN, additionally, external exposure due to loss of shielding (LOS) is considered. Atmospheric diffusion calculation is essential to determine the extent to which radioactive materials are diffused. In both RADTRAN and RISKIND, atmospheric diffusion calculations are based on Gaussian diffusion model. Users must input Pasquill stability class, release height, heat release, wind speed, temperature and mixing height, etc. Additionally, RADTRAN can input weather information relatively simply by inputting only the Pasquill stability class fraction and selecting the US average weather option. This study results will be used as a basis for developing radioactive waste overland transportation risk assessment code that reflects domestic characteristics.

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

The parent and daughter nuclides in a radioactive decay chain arrive at secular equilibrium once they have a large half-life difference. The characteristics of this equilibrium state can be used to estimate the production time of nuclear materials. In this study, a mathematical model and algorithm that can be applied to radio-chronometry using the radioactive equilibrium relationship were investigated, reviewed, and implemented. A Bateman equation that can analyze the decay of radioactive materials over time was used for the mathematical model. To obtain a differential-based solution of the Bateman equation, an algebraic numerical solution approach and two different matrix exponential functions (Moral and Levy) were implemented. The obtained result was compared with those of commonly used algorithms, such as the Chebyshev rational approximation method and WISE Uranium. The experimental analysis confirmed the similarity of the results. However, the Moral method led to an increasing calculation uncertainty once there was a branching decay, so this aspect must be improved. The time period corresponding to the production of nuclear materials or nuclear activity can be estimated using the proposed algorithm when uranium or its daughter nuclides are included in the target materials for nuclear forensics.

RADTRAN is a code that assesses the radiation risk of radioactive material transportation. RADTRAN assumes that the package is a point source or a line source regardless of package type and corrects the external dose rate using a shape factor which depends on the critical dimension of the package. However, the external dose rate calculated using a shape factor may be different from the actual external dose rate. Therefore, it is necessary to analyze the effect of the shape factor on the external dose rate. In this study, the effect of the shape factor on the external dose rate in RADTRAN was analyzed by comparison with MCNP. This study analyzed change in external dose rate depending on the distance from the package and the critical dimension. The distance from the package was in the range of 1–800 m. The shape of the package was assumed to be cylindrical with a radius of 1 m, and the critical dimensions of the package were assumed to be 2, 4, and 8 m. Attenuation and build-up in the air were not considered to consider only the effect on the shape factor. When simulating the exposure situation using MCNP, the package was assumed to be a volume source, and flux by distance from the package was calculated using F5 tally. The dose rate at 1 m from the package was normalized to 2 mSv·hr−1. As a result of the analysis, the external dose rates of the package were higher in RADTRAN than in MCNP. For the critical dimension of 2, 4, and 8 m, when the distance from package is 1–10 m, the RADTRAN was 1.83, 4.08, and 5.27 times higher on average than MCNP, respectively. And when the distance from the package was 10–100 m and 100–800 m, RADTRAN was 1.10, 2.02, 3.01 times and 1.04, 1.92, 2.43 times higher than MCNP, respectively. It was found that the larger the distance from the package is and the smaller the critical dimension of the package is, the less conservatively RADTRAN assessed. It is because the shape of the package gets closer to the point source as the distance from the package increases, and the shape factor decreases as the critical dimension of the package decreases. The result of this study can be used as the basis for radiation risk assessment when transporting radioactive materials.