The aim of this study was to investigate dry matter productivity and nitrate nitrogen content in the growth stages of barnyard millet (Echinochloa esculenta) cultivated for feed, which was treated with different nitrogen fertilization levels. An early variety of barnyard millet (cv. Shirohie) was used for the test, and the different treatments with nitrogen fertilizer were as follows: 50% (N-40 kg/ha, T1), 100% (N-80 kg/ha, reference amount, T2), 150% (N-120 kg/ha, T3), 200% (N- 160 kg/ha, T4), 250% (N-200 kg/ha, T5), and 300% (N-240 kg/ha, T6). Sowing was done on May 13, 2021 and plants were harvested for four stage; vegetative stage, elongation stage, heading stage, and milk stage. The length of the millet increased significantly as the amount of nitrogen fertilization increased during the harvest period (p<0.05), but the difference was insignificant during the milk stage (p>0.05). Moreover, barnyard millet dry matter yield increased significantly as the levels of nitrogen fertilization increased (p<0.05), but there was no significant difference in dry matter yield among nitrogen fertilization levels during the heading stage (p>0.05). Chlorophyll also was significantly higher in T5 (250%) at all harvesting times, whereas nitrate nitrogen content was highest at the vegetative stage, gradually decreased as growth progressed, and lowest at the milk stage. Finally, as the nitrogen fertilization levels increased, the nitrate nitrogen content was significantly higher in all treatment groups (p<0.05). Therefore, our results suggest that the most appropriate nitrogen fertilizer levels is between 150%–200%, considering the dry matter yield, feed ingredients and nitrate nitrogen content in barnyard millet for feed.

This study aims to establish a modified analytical method with sensitivity and reliability for streptomycin (STP) and dihydrostreptomycin (DHS) of residues level in pig tissues, plasma and urine by LC-MS/MS on the basis of previous studies. The mass parameters of quantitative and qualitative ions for STP and DHS were optimized using multiple reaction monitoring in positive mode. The separation of compounds was conducted using BEH Amide column according to material’s characteristics. The analytes in plasma were extracted with only organic solvents. In muscle and kidney, KH2PO4 buffer solution containing 2% CCl3COOH and EDTA-Na was used as extraction solvent. The WCX cartridege was selected as SPE cartridge in considering high recoveries for STP and DHS. The analytes in urine were extracted by organic solvents with acid and addition of EDTA. The limits of detection (LODs) in STP and DHS ranged 0.45~3.66 μg/kg and 0.22~0.78 μg/kg, respectively. The limits of quantification (LOQs) were 1.35~11.10 μg/kg in STP and 0.66~2.36 μg/kg in DHS. The recoveries (%) were 94.29~104.5% in STP and 92.32~108.45% in DHS except for plasma with lower values (61.45/68.5%, respectively). In the precision evaluation, the coefficient of variation (CV, %) of STP showed <10.50% on intra-day and <18.04% on inter-day. The CV (%) of DHS showed <8.42% on intra-day, whereas <17.98% on inter-day. The modified method is reliable for continuous residual monitoring in pig to ensure food safety for consumer’s health. In addition, this method could be used in study relation to residue depletion and pharmacokinetics of veterinary drug.

Canine parvovirus type-2 (CPV-2) is a major etiological agent causing gastrointestinal enteritis in domestic and wild carnivores. Since the emergence of CPV-2 in the late 1970s, subtypes CPV-2a, CPV-2b, and CPV-2c have spread worldwide. CPV-2 prevalence differed according to region and season. This study aims to investigate the prevalence of CPV-2 infection in Korea. Samples were collected from 536 dog feces in animal shelters and 225 necropsied intestinal tissues of dog carcasses submitted in the Animal and Plant Quarantine Agency (APQA) for diagnostic purposes from 2016 to 2020 in Korea. Among the 761 samples, 181 (23.8%) were positive for the following subtypes: CPV-2a (n=138), CPV-2c (n=16), CPV-2b (n=14), and CPV-2 (n=2). Feline parvovirus (n=2) and co-infection with CPV-2a and CPV-2c (n=1) were also detected. There was no significant difference in the regional distribution of CPV-2 in Korea, which is prevalent in winter. This result shows the prevalence of CPV-2 according to various environments in Korea and will be useful in establishing an effective prevention strategy against CPV-2 that reflects the situation in Korea with continuous monitoring.

To rationalize the protection of spent nuclear fuel transport storage cask, we intend to investigate the status of domestic and foreign safety regulations and related technologies to develop sabotage scenarios and analyze the protection performance and radiation impact of transport storage cask. It is essential to conduct an aircraft collision safety evaluation on spent nuclear fuel transportation and storage casks in Korea due to changes in laws and regulations related to nuclear power plant design and demand for enhanced safety. Domestic and foreign research on the protection performance of spent nuclear fuel transport storage cask was based on 9.11 events, and the results of all studies show that the speed of the aircraft and leakage of nuclear materials are insignificant. The Sandia National Laboratory (SNL) calculates Aerosol emissions from spent fuel damage in the event of sabotage and calculates Source Term based on the Durbin-Luna model. In this paper, radiation sensitivity analysis was performed due to damage to the carrier according to the size of the accident, assuming that there was a hole enough to basket from the external shell among the collision scenarios identified for domestic cask models.

As the decommissioning of nuclear power plants increases, there is an increasing interest in the amounts of radioactive waste. Especially, the radiation dose limit for packaging of radioactive wastes shall not exceed 2 mSv·h−1 and 0.1 mSv·h−1 on contact and at 2 m, respectively in South Korea. The DEMplus provides various environmental geometry and all properties such as materials, absorptions, and reflections and the estimation of the radiation dose rates is based on the radiation interactions of the designed 3D geometry model. With the consideration of the radiation dose rate by using DEMplus and its strategy of packaging plan, the radiation shielding was optimized and estimated in this paper. The modular shielded containers (MSC) with shielding inserted were used for radioactive wastes that require shielded packaging. In order to verify the accuracy of the estimated radiation dose rate by using DEMplus, the estimated results were compared with those obtained using MicroShield. The trends of the estimated radiation dose rates using DEMplus and the estimation of MicroShield were similar to each other. The results of this study demonstrated the feasibility of using DEMplus as a means of estimating the radiation dose limit in packaging plan of the radioactive waste.

In thermal cutting process, gas flow injected from the nozzle has a significant effect on the cutting materials. The gas flow is difficult to observe gases are transparent, therefore, in this study, Schlieren method was adopted to visualize the gas flow inside the kerf. The kerf shape was inserted between two slices of transparent glass in order to imitate the real cutting environment. In order to get the flow characteristics, a high speed camera was equipped and the image processing was applied to compare the before and after injection images. As a result, the method for visualizing gas flow was successfully developed and also expected to be applied to the analysis of purging gas in various welding environments.

In this study, an aerosol process was introduced to produce CaCO3. The possibility of producing CaCO3 by the aerosol process was evaluated. The characteristics of CaCO3 prepared by the aerosol process were also evaluated. In the CaCO3 prepared in this study, as the heat treatment proceeded, the calcite phase disappeared. The portlandite phase and the lime phase were formed by the heat treatment. Even if the CO2 component is removed from the calcite phase, there is a possibility that the converted CO2 component could be adsorbed into the Ca component to form a calcite phase again. Therefore, in order to remove the calcite phase, carbon components should be removed first. The lime phase was formed when CO2 was removed from the calcite phase, while the portlandite phase was formed by the introducing of H2O to the lime phase. Therefore, the order in which each phase formed could be in the order of calcite, lime, and portlandite. The reason for the simultaneous presence of the portlandite phase and the lime phase is that the hydroxyl group (OH−) introduced by H2O was not removed completely due to low temperature and/or insufficient heating time. When the sufficient temperature (900°C) and heating time (60 min) were applied, the hydroxyl group (OH−) was removed to transform into lime phase. Since the precursor contained the hydrogen component, it could be possible that the moisture (H2O) and/or the hydroxyl group (OH−) were introduced during the heat treatment process.

Uranium-235, used for nuclear power generation, has brought radioactive waste. It could be released into the environment during reprocessing or recycling of the spent nuclear fuel. Among the radioactive waste nuclides, I-129 occurs problems due to its long half-life (1.57×107 y) with high mobility in the environment. Therefore, it should be captured and immobilized into a geological disposal system through a stable waste form. One of the methods to capture iodine in the off-gas treatment process is to use silver loaded zeolite filter. It converts radioactive iodine into AgI, one of the most stable iodine forms in the solid state. However, it is difficult to directly dispose of AgI itself in an underground repository because of its aqueous dissolution under reducing condition with Fe2+. It must be immobilized in the matrix materials to prevent release of iodine as a result of chemical reaction. Among the matrix glasses, silver tellurite glass has been proposed. In this study, additives including Al, Bi, Pb, V, Mo, and W were added into the silver tellurite glass. The thermal properties of each matrix for radioactive iodine immobilization were evaluated. The glasses were prepared by the melt-quenching method at 800°C for 1 h. Differential scanning calorimetry (DSC) was performed to evaluate the thermal properties of the glass samples. From the study, the glass transition temperature (Tg) was increased by adding additives such as V2O5, MoO3, or WO3 in the silver tellurite glass. The relative electro-static field (REF) values of V2O5, MoO3, and WO3 are about three times higher than that of the glass network former, TeO2. It could provide sufficient electro-static field (EF) to the TeO2 interacting with the non-bridging oxygen forming Te-O-M (M = V, Mo, W) links. Therefore, the addition of V2O5, MoO3, or WO3 reinforced the glass network cohesion to increase the Tg of the glass. The addition of MoO3or WO3 in the silver tellurite glass increased Tg and crystallization temperature (Tc) with remaining the glass stability.

Radioactive Cesium is fission products of spent nuclear fuelwith high heat generating nuclide, having a 30 years half-life. Particularly, it is important to make stable waste form because Cs-137 have high solubility and mobility at ground water. The ceramic waste form has higher thermal and structural stability and lower solubility than glass and cement waste form. Various ceramic waste forms for Cs immobilization have been researched such as aluminosilicate (CsAlSi2O6), phosphate (CsZr2(PO4)3), titanate (CsxAlxTi8-XO16) and CsZr0.4W1.5O6. Cs pollucite is incorporated radio-Cesium to aluminosilicate framework by inorganic ion-exchange with zeolite. Therefore, it is an extremely stable structure. In previous study, we are prepared Cs pollucite pellet with various ratio of Cs precursor/matrix materials, and attempted to evaluate applicability as ceramic waste form. Cs pollucite is produced by mixing Mullite and SiO2 obtained by heat treatment Kaolinite with Cs2CO3 in ratios of 0.5, 0.6, 0.7, 0.8. Optimized ratio was 0.5 revealed single pollucite phase and the others exhibited CsAlSiO4 phase with pollucite. Cs pollucite of ratio 0.5 was pelletized under various conditions and evaluated performance as waste form. herein, the pellets were cracked on surface and edges broken. Therefore, Cs pollucite having high ratio of matrix materials contained Si and Al was prepared and pelletized, and then waste form was evaluated. The Cs pollucite powder is ratio of Cs precursor/matrix materials were 0.1, 0.2, 0.3, 0.4. Pollucite powder was mixed with 1.5, 2.0wt% Polyvinyl alcohol as binder, and dried at 70°C for overnight. Afterward, these powders obtained were pressed using punch-die apparatus at 50, 100 bar for 1 hour and the pellets with about dia. 25 mm and height 10 mm was acquired. These pellets were sintered at 1,400°C for 5 hours. Subsequently, the waste forms were evaluated physicochemical test such as compression strength, thermal conductivity, thermal expansion and leaching properties analysis.

Many countries plan to dispose of spent nuclear fuel through deep geological disposal system. In Korea, a plan is being established for the construction of a deep disposal facility to dispose of highlevel radioactive waste (or spent nuclear fuel). For construction of a deep geological repository, the NSSC (Nuclear Safety and Security Commission) stipulate that detailed technical standards for location, structure, and disposal system of deep geological repository are determined and announced by the Nuclear Safety and Security Commission Notification. Therefore, the regulatory body should carry out the process of regulatory review whether the technical standards developed by the implementer are suitable for the IAEA’s recommendations and guidelines and domestic conditions. In this process, there are many difficulties and uncertainties in terms of time and cost to independently develop safety factors in Korea by referring to the IAEA reports. So, this study intends to investigate and analyze regulatory cases for important safety factors through cases of overseas leading countries in deep geological disposal project. There are two regulatory cases intensively investigated in this study. The first is a regulatory case of regulatory bodies and external experts on the safety case, and the second is a regulatory review case in the process of site selection factor selection. In case of regulatory review of safety case, Sweden and France were selected as the representative target countries. In Sweden, safety cases such as SR-97, SR-Can, and SR-Site have been developed and there are cases of active regulatory review by regulatory agencies in the RD&D process. In France, several safety cases based on sedimentary rocks were developed and the OECD/NEA IRT (International Review Team) was inquired for review for each safety case. The site selection process is divided into a preliminary site selection stage, a site investigation stage, and a site selection and application stage. In each stage, evaluation to select a safe site is carried out using allocated siting factors of that stage. The IAEA SSG-14 report describes aspects that implementers consider in the site selection process and, with this reference, many countries are developing various siting factors and assessment methodologies in consideration of their domestic bedrock condition and geological positions. As a representative example, in Japan which is highly affected by earthquakes and igneous activities, the siting factor is classified into EF (Evaluation Factors) and FF (Favoulable Factors). So, site assessment is conducted preferentially using EF related to earthquakes and igneous activity.

To reduce the environmental burden caused by the disposal of spent nuclear fuel and maximize the utilization of the repository facility, waste burden minimization technology is currently being developed at the Korea Atomic Energy Research Institute (KEARI). The technology includes a nuclide management process that can maximize disposal efficiency by selectively separating and collecting major nuclides in spent nuclear fuel. In addition, for efficient storage facility utilization, the short-term decay heat generated by spent nuclear fuel must be removed from the waste stream. To minimize the short-term thermal load on the repository facility, it is necessary to separate heat generating nuclides such as Cs-137 and Sr-90 from the spent fuel. In particular, Sr-90 must be separated because it generates high heat during the decay process. KAERI has developed a technology for separating Sr nuclides from Group II nuclides separated through the nuclide management process. In this study, we prepared Sr ceramic waste form, SrTiO3, by using the solid-state reaction method for long-term storage for the decay of separated Sr nuclides and evaluated the physicochemical properties of the waste form. Also, the radiological and thermal characteristics of the Sr waste form were evaluated by predicting the composition of Sr nuclides separated through the nuclide management process, and the estimation of centerline temperature was carried out using the experimental thermal data and steady state conduction equation in a long and solid cylinder type waste form. These results provided fundamental data for long-term storage and management of Sr waste.