This study presents a rapid and quantitative radiochemical separation method for Nb isotopes in radioactive waste samples from the nuclear power plant with anion exchange resin after Fe coprecipitation. After radionuclides were leached from the radioactive waste samples with concentrated HCl and HNO3, the Nb isotopes were coprecipitated with Fe after filtering the leaching solution with 0.45 micron HA filter, while the Sr, Tc and Ni isotopes were in the solution. The Nb isotopes were separated in HCl medium with anion exchange resin. The purified Nb isotopes were measured using a low level liquid scintillation counter after installing quenching curve with standard Nb-94 isotopes. The separation method for Nb isotopes investigated in this study was applied to neutron dosimeter samples from the nuclear power plant after validating the Nb activity concentration with gamma spectrometry system.

It is likely to occur internal exposure for workers in Nuclear Power Plants (NPPs) due to the intake of radionuclide. To assess the internal exposure dose the measurement of activity for remain radionuclide is necessary. The Whole Body Counters (WBCs) are commonly used for measurement of remain radionuclide activity in human body. Korea Hydro & Nuclear Power Co., Ltd. (KHNP) conduct performance test of WBCs in all NPPs for every year to confirm the performance of equipment. The performance test is conducted using unknown sources and the participants of the comparison test submit the radionuclide and activity of the unknown sources measured by WBC as a result. The performance indicator and criteria for WBC recommended in the American National Standards Institute (ANSI) N13.30 report published in 2011 are applied. The performance indicator is Root Mean Squared Error (RMSE) and criteria is 0.25 or less. The results of performance test performed in 2022 for all WBC is meet the ANSI N13.30 criteria. And the RMSE values are confirmed from 0.01 to 0.23. This means that the residual radioactivity measurement results using WBC are reliable.

In this research, KPS manufactured Full System Decontamination (FSD) equipment, which is consisted of Oxidizing Agent Manufacturing System (OAMS), Chemical Injection System (CIS), RadWaste Treatment System (RWTS), Chemical Waste Decomposition & Treatment System (CWDS) and conducted demonstration test to prepare Decontamination and Decommissioning (D&D) project of Kori nuclear power plant in Korea. Each equipment of FSD was modularized due to the limited size of equipment hatch of Kori nuclear power plant. To simulate the expected circumstances in nuclear power plant such as usage of heater or position of each equipment, additional equipment was used. The chemical concentration and flow rate of process water for FSD were used as mentioned in the previous study by KHNP CRI. FSD was conducted for three cycles and each cycle was consisted of oxidation, reduction, chemical decomposition and purification. Oxidation and reduction process were conducted at 90°C. Chemical decomposition and purification process were conducted at 40°C due to the damage of UV lamp and IX by the heat. Total volume of process water for FSD demonstration test was 2.5 m2. KPS conducted decontamination performance review by calculating thickness reduction and weight loss of installed specimen. Operational review was conducted as if FSD test was conducted in the field based on the result of demonstration test. One of the most prioritized features is the workers’ safety. Also, the appropriate position of equipment needs to be considered to meet the required specification of component.

In nuclear power plant, there were many contaminated tanks dispose of radioactive fluid waste. These tanks are made of stainless-steel, and corrosion can occur when tanks are exposed to radioactive fluid waste containing moisture for a long time. Therefore, those sludge waste including radionuclide should be collected, solidified, and disposed of. If sludge can be melted, sludge can be easily solidified. However, melting points of sludge components (Fe2O3, NiO, Cr2O3) are very high as 1565, 1955, and 2435 , respectively. Therefore, melting sludge is difficult. If a solidification auxiliary material such as cement or asphalt is used to help solidify, solidification can easily occur, but cement and asphalt are vulnerable to heat. Vitrification using glass material can be solidification method, but the waste loading ratio of glass material is higher than 50%. High waste loading ratio is weakness in terms of volume reduction of waste. In this study, ferro frit powder (Na2O, K2O, CaO, Al2O3, B2O3, SiO2, ZnO) is used as solidification auxiliary material. When ferro frit powder mixed with sludge material are melted in sludge material, melted ferro frit powder can stick sludge material and can solidify sludge material without melting. Sludge can be solidified by using ferro frit powder with a smaller waste loading ratio than the vitrification method. However, since the waste loading ratio of the solidification auxiliary material is small, if ferro frit powder is not uniformly distributed between sludge powder, solidification may not be performed properly. Although the mixing ratio between sludge and ferro frit in solidified sludge is same, when the distribution of ferro frit powder in sludge is non-homogeneous, the difference in chemical and physical characteristics as compressive strength and leaching resistance can be observed in solidified sludge. As the ferro frit mixing ratio in the site where ferro frit exists was relatively high, the melting point of the mixed powder (sludge+ferro frit) decreased, and the mixed powder could not maintain its shape and melted. In the case of the area where ferro frit does not exist, since only the stainless-steel oxide sludge exists, sludge was not melted, and the shape was maintained. However, it was confirmed that the leaching resistance was lowered by visually observing the color change of the leachate within a short period of time (about 2 hours) when solidified sludge was immersed in the leachate.

Lubricant oil waste contaminated with radioactive materials generated at nuclear facilities can be disposed of as industrial waste in accordance with self-disposal standards if only radioactive materials are removed. Lubricant oil used in nuclear facilities consists of oil of 75-85% and additives of 15-25%, and lubricant oil waste contains heavy metals, carbon, glycol, etc. In addition, lubricant oil waste from nuclear facilities contains metallic gamma-ray emission radionuclides including Co-60, Cs-137 and volatile beta-ray emission radionuclides such as C-14 and H-3, which are not present in lubricant oil waste from general industries and these radionuclides must be eliminated according to the Atomic Energy Act. In general industries, the wet treatment technologies such as acid-white soil treatment, ion purification, thin film distillation, high temperature pyrolysis, etc. are used as the refining technology of lubricant oil waste, but it is difficult to apply these technologies to nuclear industrial sites due to restrictions related with controlling the generation of secondary radioactive waste in sludge condition containing radionuclides of metal components, and limiting the concentration of volatile radioactive elements contained in refined oil to be below the legal threshold. In view of these characteristics, the refinement system capable of efficiently refining and treating lubricant oil waste contaminated with radioactive materials generated in nuclear facilities has been developed. The treatment process of this R&D system is as follows. First, the moisture in the radioactive lubricant oil waste pretreated through the preprocessing system is removed by the heated evaporating system, and the beta-emission radionuclides of H-3 and C-14 can be easily removed in this process. Second, the heated lubricant oil waste by the heated evaporating system is cooled through the heat exchanging system. Third, the particulate matters with gamma-ray emission radionuclides are removed through the electrostatic ionizing system. Forth, the lubricant oil waste is stored in the storage tank and the purified lubricant oil waste is discharged to the outside after sampling and checking from the upper, middle and lower positions of the lubricant oil waste stored in the storage tank. Using this R&D system, it is expected that the amount of radioactive waste can be reduced by efficiently refining and treating lubricant oil waste in the form of organic compounds contaminated with radioactive materials generated in nuclear facilities.

Kori-1, the nuclear power plants in South Korea, first started operation in April 1978 and was suspended permanently in 2017. The saturation rate time of spent nuclear fuel generated by major nuclear power plants operating in Korea are getting closer. If we fail to dispose spent nuclear fuel, which is equivalent to high-level radioactive waste, the nuclear power plants will have to be shutdown. High-level radioactive waste is permanently disposed through a deep geological disposal system because it contains long-term half-life nuclides and emits high energy. To select the deep geological disposal site and construct the disposal facilities, it is necessary to establish appropriate regulatory policies accordingly. The status of database construction in OECD-NEA, NRC, SITEX, and IAEA, which provides safety regulations for deep geological disposal system, stipulates each requirement for dismantling nuclear power plants. However, details such as specific figures are not specified, and guidelines for the disposal of high-level radioactive wastes are not clearly distinguished. In Korea, the CYPRUS program, an integrated database system, has been developed to support comprehensive performance evaluation for high-level waste disposal. However, due to several difficult situations, maintenance and upgrades have not been performed, so the research results exist only in the form of raw data and the new research results have not been reflected. Other than that, there is no preemptive basis for regulating the deep geological disposal system. With real-time database, we can develop a regulatory system for the domestic deep disposal system by systematically analyzing the regulatory condition and regulatory case data of international organizations and foreign leading countries. The database system processed and stored primary data collected from nuclear safety reports and other related data. In addition, we used relational database and designed table to maximize time and space efficiency. It is provided in the form of a web service so that multiple users can easily find the data they want at the same time. Based on these technologies, this study established a database system by analyzing the legal systems, regulatory standards, and cases of major foreign leading countries such as Sweden, Finland, the United States, and Japan. This database aims to organize data for each safety case component and further prepare a safety regulatory framework for each stage of development of disposal facilities suitable for the domestic environment.

CYPRUS is a web-based waste disposal research comprehensive information management program developed by the Korea Atomic Energy Research Institute over three years from 2004. This program is stored as existing quality assurance documents and data, and the research results can be viewed at any time. In addition, it helps to perform all series of tasks related to the safety evaluation study of the repository in accordance with the quality assurance system. In the future, it is necessary to improve the user convenience by clarifying the relationship between FEP and scenarios and upgrading output functions such as visualization and automatic report generation. This purpose of this study is to research and develop the advanced program of CYPRUS. This study is based on building FEP, DIM and scenario databases. It is necessary to develop an algorithm to analyze and visualize the FEP, DIM and scenario relationship. This project is an integrated information processing platform for DB management and visualization considering user convenience. The first development goal is to build long-term evolutionary FEP, DIM, and scenarios as a database. The linkage by FEP item was designed in consideration of convenience by using a mixed delimiter of letters and numbers. This design provides information on detailed interactions and impacts between FEP items. Scenario data lists a series of events and characteristic change information for performance evaluation in chronological order. In addition, it includes information on FEP occurrence and mutual nutrition by period, and information on whether or not the repository performance is satisfied by item. The second development goal is to realize the relationship analysis and visualization function of FEP and scenario based on network analysis technique. Based on DIM, this function analyzes and visualizes interactions between FEPs in the same way as PID, RES, etc. In addition, this function analyzes FEP and DIM using network analysis technique and visualizes it as a diagram. The developed platform will be used to construct and visualize the FEP DB covering research results in various disposal research fields, to analyze and visualize the relationship between core FEP and scenarios, and finally to construct scenarios and calculation cases that are the evaluation target of the comprehensive performance evaluation model. In addition, it is expected to support the knowledge exchange of experts based on the FEP and scenario integrated information processing platform, and to utilize the platform itself as a part of the knowledge transfer system for knowledge preservation.

Nuclear power plant decommissioning generates significant concrete waste, which is slightly contaminated, and expected to be classified as clearance concrete waste. Clearance concrete waste is generally crushed into rubble at the site or a satellite treatment facility for practical disposal purposes. During the process, workers are exposed to radiation from the nuclides in concrete waste. The treatment processes consist of concrete cutting/crushing, transportation, and loading/unloading. Workers’ radiation exposure during the process was systematically studied. A shielding package comprising a cylindrical and hexahedron structure was considered to reduce workers’ radiation exposure, and improved the treatment process’s efficiency. The shielding package’s effect on workers’ radiation exposure during the cutting and crushing process was also studied. The calculated annual radiation exposure of concrete treatment workers was below 1 mSv, which is the annual radiation exposure limit for members of the public. It was also found that workers involved in cutting and crushing were exposed the most.

This study assessed the changes in the fiber properties of virgin and recovered fibers from lab-scale and pilot-scale depolymerization reactors based on the thermal air oxidation-resistance characteristics. Lab-scale and pilot-scale depolymerization reactors had different depolymerization volumes. Results showed that the lab-scale and pilot-scale peak solvent temperatures were 185 °C and 151 °C, respectively. The lab-scale had highest solvent temperature rate increase because of the small depolymerization volume and the dominant role of the cavitation volume. The structural properties of the recovered and virgin fibers were intact even after the depolymerization and after the pretreatment and oxidation-resistance test. We observed 1.213%, 1.027% and 0.842% weight loss for the recovered (lab-scale), the recovered (pilot-scale) and virgin fibers because of the removal of impurities from the surface and chemisorbed gases. Further, we observed 0.8% mass loss of the recovered fibers (lab-scale) after the oxidative-onset temperature because of the “cavitation erosion effect” from the dominant of the cavitation bubbles. The “cavitation erosion effect” was subdued because of the increased depolymerization volume in the pilot-scale reactor. Therefore, negligible impact of the pilot-scale mechanochemical recycling process on the structure and surface characteristics of the fibers and the possibility of reusing the recovered fibers recycling process were characteristic. Representative functional groups were affected by the thermal oxidation process. We conducted HPLC, HT-XRD, TGA– DSC, XPS, SEM, and AFM analysis and provided an extensive discussion of the test thereof. This study highlighted how misleading and insufficient small-lab-scale results could be in developing viable CFRP depolymerization process.

Two tree frogs, Dryophytes suweonensis and Dryophytes japonicus, inhabiting Korea, are morphologically similar and share the same habitats. Therefore, they are identified mainly through their calls, especially for males. Dryophytes suweonensis is registered as an endangered (IUCN: EN grade) and protected species in South Korea. Thus, it is necessary to develop a method to rapidly identify and discriminate the two species and establish efficient protection and restoration plans. We identified significant genetic variation between them by sequencing a maternallyinherited mitochondrial 12S ribosomal DNA region. Based on the sequence data, we designed a pair of primers containing 7 bp differences for high resolution melting (HRM) analysis to rapidly and accurately characterize their genotypes. The HRM analysis using genomic DNA showed that the melting peak for D. suweonensis was 76.4±0.06°C, whereas that of D. japonicus was 75.0±0.05°C. The differential melt curve plot further showed a distinct difference between them. We also carried out a pilot test for the application of HRM analysis based on immersing D. suweonensis in distilled water for 30 min to generate artificial environmental DNA (eDNA). The results showed 1.10-1.31°C differences in the melting peaks between the two tree frog samples. Therefore, this HRM analysis is rapid and accurate in identifying two tree frogs not only using their genomic DNA but also using highly non-invasive eDNA.

Heterogeneous nuclear ribonucleoprotein A2/B1 (hnRNPA2/B1) is an N6-methyladenosine (m6A) RNA modification regulator and a key determinant of premRNA processing, mRNA metabolism and transportation in cells. Currently, m6A reader proteins such as hnRNPA2/B1 and YTHDF2 has functional roles in mice embryo. However, the role of hnRNPA2/B1 in porcine embryogenic development are unclear. Here, we investigated the developmental competence and mRNA expression levels in porcine parthenogenetic embryos after hnRNPA2/B1 knock-down. HhnRNPA2/B1 was localized in the nucleus during subsequent embryonic development since zygote stage. After hnRNPA2/B1 knock-down using double stranded RNA injection, blastocyst formation rate decreased than that in the control group. Moreover, hnRNPA2/B1 knock-down embryos show developmental delay after compaction. In blastocyste stage, total cell number was decreased. Interestingly, gene expression patterns revealed that transcription of Pou5f1, Sox2, TRFP2C, Cdx2 and PARD6B decreased without changing the junction protein, ZO1, OCLN, and CDH1. Thus, hnRNPA2/B1 is necessary for porcine early embryo development by regulating gene expression through epigenetic RNA modification.

To assist farmers in shifting from subsistence to commercial agriculture, the government of Laos has promoted farmer organizations throughout the country. The farmer production groups have gradually transformed and emerged into strong cooperatives or associations. The purpose of this study was to examine the factors affecting farmers’ participation in FPGs. Data were collected from primary and secondary sources. A total of 140 respondents participated in a field survey in three kumb ban (village groups) in the Xepon District. Descriptive statistics and binary logistics regression models were used to analyze the data. The results show that the education level of household head, farmland size, access to credit, and perceived trust for farmer production group membership are significantly different from those who are non-members of the groups. However, non-members of FPGs face more constrain in accessing the market, as compared to the members of FPGs. The empirical results from the Binary Logit Model indicate that education level of headed household, farmland size, access to credit, perceived trust, and constraint to market are factors that significantly determine farmers’ participation in FPGs. This study contributes to the empirical literature by providing a better understanding of farmer participation in the groups. The government of Laos should consider and promote access to credit, farmer learning, and market integration to improve farmers' participation in FPGs.

Vitamin K1 (VK1) has been widely used as a coumarin antagonist and for the treatment of hemorrhagic disease in veterinary practice. However, the potential mechanism of adverse reaction after VK1 injection has been not fully elucidated. In this study, two cases of anaphylactic reactions after subcutaneous VK1 injection were presented, and then an experimental study was performed to further characterize the anaphylactic reactions. Two dogs developed anaphylactic reactions after subcutaneous VK1 injections and were promptly treated with antihistamines and glucocorticoids, after which abnormal signs related to anaphylaxis disappeared. Subsequently, a study was undertaken to ascertain the nature of the adverse reactions to subcutaneous VK1 injection. Six healthy laboratory beagle dogs received subcutaneous VK1 administrations once daily for eight days. They were monitored for clinical signs after each injection, and blood samples were collected for the measurement of plasma histamine and immunoglobulin E concentrations using enzyme-linked immunosorbent assay. All six dogs showed mild angioedema after the VK1 injections. The dogs also displayed clinical signs including sneezing, coughing, skin reddening, excess salivation, pawing the ground, and somnolence on days 4, 6, and 8. Plasma histamine and immunoglobulin E concentrations were significantly increased by the repeated injections. In summary, this study describes anaphylactic reactions resulting from subcutaneous VK1 administration in dogs. Clinicians should be aware that the repeated subcutaneous injection of VK1 can trigger an anaphylactic reaction in dogs.

This study presents a rapid and quantitative sequential separation method for H-3 and C-14 isotopes with distillation apparatus in environmental samples released from nuclear facilities. After adding 200 mg of granulated potassium permanganate and 500 mg of sodium hydroxide in 100 mL of sample solution, the sample solution was heated until approximately 10 mL of distillate, and the distillate fraction was removed. The sample solution was heated again until a minimum 10 mL of additional distillate was collected. 10 mL of distillate was transferred to the LSC vail and the measurement sample for H-3 was made by adding 10 mL of Ultima Gold LLT to the LSC vial. After adding 2.5 g of potassium persulfate, 2 mL of 1M silver nitrate and 15 mL of concentrated nitric acid to the remained sample solution, the sample solution was heated for 90 minutes and C-14 isotopes were adsorbed into 10 mL of Carbo-Sorb solution in glass vial. The measurement sample for C-14 was made by adding 10 mL of Permafluor to the C-14 fraction in glass vial. The purified H-3 and C-14 samples were measured by the liquid scintillation counter after quenching correction. The average recoveries of H-3 and C-14 with CRM were measured to be 96% and 85%, respectively. The sequential separation method for H-3 and C-14 investigated in this study was applied to activated charcoal filter produced from nuclear power plants after validating the reliability by result of proficiency test (KOLAS-KRISS, PT-2021-51).

In 2005, groundwater contamination due to unplanned releases of radioactive materials from the US. Nuclear Power Plants (NPPs) such as Braidwood and Indian Point was confirmed. The following year, in 2006, The Nuclear Regulatory Commission (NRC) established a task force team to investigate the history of unplanned release of all NPP in the US. As a results 217 events of unplanned release including leaks and spills were identified in the US NPPs. The NRC regulates the radioactivity concentration of off-site groundwater by setting a reporting levels (RLs), and if exceeds the RLs, the licensee must report within 30 days. When the off-site groundwater is used as drinking water or non-drinking water, the RLs for tritium in groundwater are 740 Bq·L−1 or 1,110 Bq·L−1, respectively. Whereas the NRC does not set the RLs for on-site groundwater. The Nuclear Energy Institute (NEI) issued the guidance document “Industry groundwater protection initiative” NEI 07-07 in 2007. And the members of the NEI promised with regulatory body and local governments to implement groundwater monitoring/protection program according to the NEI 07-07. The document states that when the on-site groundwater is used as drinking water, the RL (740 Bq·L−1) for off-site groundwater will be applied and the licensee voluntarily reported to the NRC. And also, NPPs are setting the Investigation Level (IL) below the RP and the IL is various among the NPPs. The IL is the standard by which detailed investigations are implemented when the level (radioactivity concentration) is exceeded.