Following the previous study, which investigated the pharmacological properties of the Technekitty injection (Tc-99m), the toxicity of a single intravenous administration of the Technekitty injection (Tc-99m) and the side effects that may occur at the diagnostic dose were confirmed. The Technekitty injection (Tc-99m) was administered intravenously once at a dose of 0, 0.67, 2.0, and 6.0 mCi/kg to 5 male and female rats per group. Mortality, general symptom observation, and weight measurement were performed for 2 weeks, followed by observation of autopsy findings. There were no deaths, and no statistically significant weight change was observed. No abnormal systemic signs related to the Technekitty injection (Tc-99m) were observed. These results confirmed that Technekitty injection (Tc-99m) can be safely administered intravenously at doses up to 6.0 mCi/kg. Additionally, technetium-99m at an average dose of 2 mCi (74 MBq) has been verified as a diagnostic dose without adverse effects, allowing the Technekitty injection (Tc-99m) to be used safely without side effects at this dosage. This study demonstrates that the Technekitty injection (Tc-99m) has a wide safety margin, supporting its potential for clinical application. Moreover, these findings align with the nonclinical safety standards for radiopharmaceuticals, reinforcing its utility in veterinary medicine. The Technekitty injection (Tc-99m) is expected to be applicable for clinical diagnosis as a veterinary drug in Korea.

Thyroid scanning using technetium-99m (99mTc) is the gold standard for diagnosing feline hyperthyroidism. In cats with an overactive thyroid, a thyroid scan is the most appropriate imaging technique to detect and localize any hyperfunctional adenomatous thyroid tissue. In this study, the pharmacological properties of the Technekitty injection (Tc-99m), developed as a diagnostic agent for feline hyperthyroidism using 99mTc as an active ingredient, were tested in FRTL-5 thyroid follicular cell line and ICR mice. The percentage of cell uptake of the Tc-99m in FRTL-5 thyroid cells was 0.182 ± 0.018%, which was about 6 times higher compared to Clone 9 hepatocytes. This uptake decreased by 38.2% due to competitive inhibition by iodine (sodium iodide). In tissue distribution tests by using ICR mice, the highest distribution was observed in the liver, kidneys, spleen, lungs, and femur at 0.083 hours after administration, and this distribution decreased as the compound was excreted through the kidneys, the primary excretory organ. Maximum distribution was confirmed at 1 hour in the small intestine, 6 hours in the large intestine, and 2 hours in the thyroid gland. Additionally, the total amount excreted through urine and feces over 48 hours (2 days) was 78.80% of the injected dose, with 37.70% (47.84% of the total excretion) excreted through urine and 41.10% (52.16% of the total excretion) through feces. In conclusion, the Tc-99m has the same mechanism of action, potency, absorption, distribution, metabolism, and excretion characteristics as 99mTc used for feline hyperthyroidism in the United States, Europe, and other countries, because the Technekitty injection (Tc-99m) contains 99mTc as its sole active ingredient. Based on these results, the Technekitty injection (Tc-99m) is expected to be safely used in the clinical diagnosis of feline hyperthyroidism.

After the permanent shut down of Kori Unit 1, various decommissioning activities will be implemented, including decontamination, segmentation, waste management, and site restoration. During the decommissioning period, waste management is among the most important activities to ensure that the process proceeds smoothly and within the expected timeframe. Furthermore, the radioactive waste generated during the operation should be sent to a disposal facility to complete the decommissioning project. Square and cylindrical concrete re-package drums were generated during the 1980s and 1990s. The square, containing boron concentrates, and cylindrical, containing spent resin, concrete re-package drums have been stored in a radioactive waste storage building. Homogeneous radioactive waste, including boron concentrates, spent resin, and sludge, should be solidified or packaged in high-integrity containers (HICs). This study investigates the sequential segmentation process for the separation of contaminated and non-contaminated regions, the re-packaging process of segmented or crushed cement-solidified boron concentrate, and re-packaging in HICs. The conceptual design evaluates the re-packaging plan for the segmented and crushed cement-solidified waste using HICs, which is acceptable in a disposal facility, and the quantity of generated HICs from the treatment process.

Valgus Kolbe, 1909 is a small genus of Cetoniinae, with 20 described species worldwide. In Korea, only one species, Valgus koreanus Sawada, have been recorded. Species of this genus have been known that they are commonly associated with termite colonies. They feed on the wall of termite burrows in logs or standing dead trees. In this study, we report a new species of this genus, Valgus gwangneungensis sp. nov.. We provide a key to the species of Valgus, description of the new species and photographs of habitus and male aedeagus.

The large rectangular and cylindrical concrete drums are stored in nuclear power plant (NPP) for a long time. At the early stage of NPP operation, the treatment technology of boron concentrates and spent resin was not well developed, when compared to current system. Since the waste acceptance criteria (WAC) of the disposal facility was not established, the boron concentrates and spent resins were packaged in 200 L drum. Some of the 200 L drums, which contain relatively high dose rate radioactive waste, were stored in large concrete drum. The concrete drum offers superior shielding effect and allows reduction of radiation exposure to workers. The WAC requires various characteristics: radiological characteristics, physical characteristics, chemical characteristics, etc. The non-destructive method allows the rapid evaluation and estimation of the concrete structure. Also, it is expected that the large concrete exhibits integrity after the measurements. In this paper, the non-destructive method to understand the large rectangular and cylindrical drum is systematically studied. The advantage and disadvantage of the non-destructive methods were compared in this paper. In addition, the optimized methodology to characterize the radioactive waste containing large rectangular and cylindrical drum will be discussed in this paper.

Radioactive cesium is a heat generated and semi-volitile nuclide in spent nuclear fuel (SNF). It is released gasous phase by head-end treatment which is a pretreatment of pyroprocessing. One of the capturing methods of gasous radioactive cesium is using zeolite. After ion-exchanged zeolite, it is transformed to ceramic waste form which is durable ceramic structure by heat treatment. Various ceramic wasteforms for Cs immobilization have been researched such as cesium aluminosilicate (CsAlSi2O6), cesium zirconium phosphate (CsZr2(PO4)3), cesium titanate (CsxAlxTi8-xO16, Cs2TiNb6O18) and CsZr0.5W1.5O6. The cesium pollucite is composed to aluminosilicate framework and cesium ion incorporated in matrix materials lattices. Many researchers are reported that the pollucite have high chemical durability. In this study, the Cesium pollucite was fabricated using mixtures of aluminosilicate denoted Absorbent product (AP) and Cs2CO3 by calcination and pelletized by cold pressing. The characterization of fabricated pollucite powder and pellets was analyzed by XRD, TGA, SEM, SEMEDS and XRF. The chemical durability of pollucite powder was evaulated by PCT-A and ICP-MS and OES. Thus, the optimal pressure condition without breaking the pellets which is low Cs2O/AP ratio and pelletizing pressure was selected. The long-term leaching test was performed using MCC-1 method for 28 days with the fabricated pollucite pellets. The leachate of leaching test was allard groundwaster and Deionized water and replaced 5 contact periods which is 3 hours, 3 days, 7 days, 14 days and 28 days and analyzed by ICPMS. The leaching rate was shown two stages. The first stage was rapid and relatively large amount of nuclides were leached. The leaching rate was decreased in the second stage. The fractional release rate of this study was shown same trend. These results were similar to previous studies.

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.