돼지의 체세포 핵이식(Somatic cell nuclear transfer,SCNT)은 인간에게 약리적 효과가 있는 단백질, 이종 간 장기이식(xenotransplantation)에 사용되는 장기, 질병 연 구 목적의 모델 동물을 제공한다. 특히 형질전환 돼지를 활용한 심장 이식이 세계 최초로 성공한 후 형질전환 돼 지 생산의 안정화는 다음 연구를 위한 중요한 점으로 대 두되고 있으나, 미니돼지의 체세포 핵이식 배아의 생산 효율은 아직 낮은 실정이다. 형질전환의 성공은 양질의 SCNT 배아 생산에서 시작되어야 한다. 이러한 SCNT 배 아의 생산 효율을 향상할 수 있는 요인 중에는 공여 세포 의 형태가 있으며, 성공적인 공여 세포의 생산을 위해서 는 종축에 따른 세포의 특성을 파악하여야 하고, 혈액형 의 차이에서 발생하는 문제점 해결을 위해 OO 타입의 선 별이 필요하다. 본 연구에서는 지속적인 계대 배양을 통 하여 공여 세포로 사용되는 미니돼지의 태아섬유아세포의 계대 배양 조건을 확립하고자 한다. 또한 미니돼지의 혈 액형을 PCR 기반으로 분석하여 분류하고 OO 타입의 선 별을 통하여 이종 간 이식에 용이하게 공여 세포의 조건 을 확립하였다. 이후 sgRNA(single guide RNA)를 사용하 여 CRISPR-Cpf1로 GGTA1(α-1,3 galactosyl-transferase) 유전자를 knock-out 한 미니돼지의 생산으로, 급성면역반 응을 유발하는 Gal(1,3)Gal epitope이 제거된 미니돼지의 세포 주를 구축 및 체세포 핵이식을 통해 GGTA1 knock-out 미니돼지를 생산하였으며, 이러한 연구는 이후 체세포 핵이식 및 이종 간 장기이식에 중요한 기초자료로 사용될 것이라고 생각된다.

Screening of Bacillus subtilis strains capable of producing extracellular hydrolases, including lipolytic enzyme was carried out. Within the scope of this study, total 61 strains were isolated from various domestic specific environmental samples such as seawater, soil, Jeju ranch, hot spring, salt farm, and fermented food and identified as Bacillus subtilis based on 16S rRNA sequences. Among the isolates, 4 strains had extracellular lipase activity. The growth profile of the strains revealed that 4 strains showed well growth at 50℃, 3 strains at 55℃, and pH 7.0. All strains could tolerate salinity up to 5%(w/v), 3 strains up to 10%(w/v), and only one strain showed growth at 15%(w/v). Additionally, amylolytic and proteolytic activities were detected in these strains. The highest lipolytic, amylolytic, and proteolytic activity was detected in Bacillus subtilis YS-YR 5A. These results demonstrate the potential application of extracellular hydrolase-producing Bacillus subtilis strains, especially strain YS-YR 5A, as fermentation starter to enhance the functionality and multiplication of functional natural products in the food/medicine/cosmetics/bio industries.

In this study, superior carbon nanotubes (CNT) were chemically modified with itaconic acid (IA) and a polyaniline (PANI) composite was formed and used to remove methylene blue (MB) dye from an aqueous solution. The capacity of CNTs modified with IA (IA/CNT) and composited with PANI (PANI/CNT) to remove MB dye from an aqueous solution was compared and investigated. The effects of parameters such as pH (3–10), adsorbent dose (0.8–8 g/L), initial dye concentration (10–100 mg/L), and temperature (25–55 °C) on MB adsorption were investigated. IA/CNT and PANI/CNT adsorbents were characterized by analyzes such as Fourier Transform Infrared Spectroscopy (FT-IR), Field Emission Scanning Electron Microscopy (FE-SEM), transmission electron microscope (TEM), and Brunauer, Emmett and Teller (BET). It was determined that the isotherm data fit the Langmuir isotherm model. The maximum adsorption capacity (qmax) of PANI/CNT and IA/CNT calculated according to this model (at 25 °C) was 12.78 and 32.78 mg g− 1, respectively. Thermodynamic analysis results showed that the adsorption was exothermic, feasible, and spontaneous. It can be said that the possible mechanism of MB on PANI/CNT and IA/CNT adsorbents occurs with the participation of π–π interaction, electrostatic attraction and hydrogen bonding.

Kori unit 1, the first PWR (Pressurized Water Reactor) in Korea, was permanent shut down in 2017. In Korea, according to the Nuclear Safety Act, the FDP (Final Decommissioning Plan) must be submitted within 5 years of permanent shutdown. According to NSSC Notice, the types, volumes, and radioactivity of solid radioactive wastes should be included in FDP chapter 9, Radioactive Waste Management, Therefore, in this study, the types depending on generation characteristics and radiological characterization methods and process of solid radioactive waste were analyzed. Solid radioactive waste depending on the characteristics of the generation was classified into reactor vessel and reactor vessel internal, large components, small metals, spent nuclear fuel storage racks, insulation, wires, concrete debris, scattering concrete, asbestos, mixed waste, soil, spent resins and filters, and dry active waste. Radiological characterization of solid radioactive waste is performed to determine the characteristics of radioactive contamination, including the type and concentration of radionuclides. It is necessary to ensure the representativeness of the sample for the structures, systems and components to be evaluated and to apply appropriate evaluation methods and procedures according to the structure, material and type of contamination. Therefore, the radiological characterization is divided into concrete and structures, systems and components, and reactor vessel, reactor vessel internal and bioshield concrete. In this study, the types depending on generation characteristics and radiological characterization methods and process of solid radioactive waste were analyzed. The results of this study can be used as a basis for the preparation of the FDP for the Kori unit 1.

As an initial part of Kori-1 & Wolsung-1 Unit decommissioning planning, a characterization plan is developed to define the nature, extent and location of contaminants, determine sampling locations and protocols, determine quality assurance objectives for characterization, and define documentation requirements. The actual characterization of a facility is an iterative process that involves initial sampling according to the characterization plan, field management (such as labeling, packaging, storing, and transport) of the samples, laboratory analysis, conformance to the data quality objectives (DQOs), and then identifying any additional sampling required, refining the DQOs, and modifying the characterization plan accordingly. The final product of the facility characterization is a document that describes the type, amount, and location of contaminants that will require consideration and removal during the decommissioning operations sufficient to prepare a decommissioning plan. In this study, implementing a characterization plan, developed in accordance with this standard, will result in obtaining or deriving the above information.

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.

As the importance of radioactive waste management has emerged, quality assurance management of radioactive waste has been legally mandated and the Korea Radioactive Waste Agency (KORAD) established the “Waste Acceptance Criteria for the 1st Phase Disposal Facility of the Wolsong Lowand Intermediate-Level Waste Disposal Center (WAC)”, the detailed guideline for radioactive waste acceptance. Accordingly, the Korea Atomic Energy Research Institute (KAERI) introduced a radioactive waste quality assurance management system and developed detailed procedures for performing the waste packaging and characterization methods suggested in the WAC. In this study, we reviewed the radioactive waste characterization method established by the KAERI to meet the WAC presented by the KORAD. In the WAC, the characterization items for the disposal of radioactive waste were divided into six major categories (general requirements, solidification and immobilization requirements, radiological, physical, chemical, and biological requirements), and each subcategories are shown in detail under the major classification. In order to satisfy the characterization criteria for each detailed item, KAERI divided the procedure into a characterization item performed during the packaging process of radioactive waste, a separate test item, and a characterization item performed after the packaging was completed. Based on the KAERI’s radioactive waste packaging procedure, the procedure for characterization of the above items is summarized as follows. First, during the radioactive waste packaging process, the characterization corresponding to the general requirements (waste type) is performed, such as checking the classification status of the contents and checking whether there are substances unsuitable for disposal, etc. Also, characterization corresponding to the physical requirements is performed by checking the void fraction in waste package and visual confirmation of particulate matter, substances containg free water, ect. In addition, chemical and biological requirements can be characterized by visually confirming that no hazardous chemicals (explosive, flammable, gaseous substances, perishables, infectious substances, etc.) are included during the packaging process, and by taking pictures at each packaging steps. Items for characterization using separate test samples include radiological, physical, and chemical requirements. The detailed items include identification of radionuclide and radioactivity concentration, particulate matter identification test, free water and chelate content measurement tests, etc. Characterization items performing after the packaging is completed include general requirements such as measuring the weight and height of packages and radiological requirements such as measurements of surface dose rate and contamination, etc. All of the above procedures are proceduralized and managed in the radioactive waste quality assurance procedure, and a report including the characterization results is prepared and submitted when requesting acceptance of radioactive waste. The characterization of KAERI’s radioactive waste has been systematically established and progressed under the quality assurance system. In the future, we plan to supplement various items that require further improvement, and through this, we can expect to improve the reliability of radioactive waste management and activate the final disposal of KAERI’s radioactive waste.

The decommissioning of Korea Research Reactor Units 1 and 2 (KRR-1&2), the first research reactors in South Korea, began in 1997. Approximately 5,000 tons of waste will be generated when the contaminated buildings are demolished. Various types of radioactive waste are generated in large quantities during the operation and decommissioning of nuclear facilities, and in order to dispose of them in a disposal facility, it is necessary to physico-chemically characterize the radioactive waste. The need to transparently and clearly conduct and manage radioactive waste characterization methods and results in accordance with relevant laws, regulations, acceptance standards is emerging. For radioactive waste characterization information, all information must be provided to the disposal facility by measuring and testing the physical, chemical, and radiological characteristics and inputting related documents. At this time, field workers have the inconvenience of performing computerized work after manually inputting radioactive waste characterization information, and there is always a possibility that human errors may occur during manual input. Furthermore, when disposing of radioactive waste, the production of the documents necessary for disposal is also done manually, resulting in the aforementioned human error and very low production efficiency of numerous documents. In addition, as quality control is applied to the entire process from generation to treatment and disposal of radioactive waste, it is necessary to physically protect data and investigate data quality in order to manage the history information of radioactive waste produced in computerized work. In this study, we develop a system that can directly compute the radioactive waste characterization information at the field site where the test and measurement are performed, protect the stored radioactive waste characterization data, and provide a system that can secure reliability.

Discontinuities exert great influence on the thermal, hydraulic, and mechanical behavior of rock mass. Rock joint is one of the most frequently encountered discontinuities in many engineering applications, such as tunnel, rock slope and repository for high level radioactive waste. Therefore, the effects of rock joint should be thoroughly investigated in various aspects. Rock joint has gone through many geological processes and its behavior can be characterized by many properties. Among them, geometric properties, such as joint roughness, aperture, and contact area can affect mechanical and hydraulic properties and vice versa. Therefore, accurate understanding and characterization of the geometric properties are of importance. Generally, the geometric properties of a joint are obtained or estimated using the surface height or elevation, which could be measured by various contact or noncontact methods. Then, the coordinates of the surfaces are used to calculate several parameters, for instance roughness indexes and mechanical aperture, in a quantitative manner. This paper is a part of SKB task force project that aims to evaluate the geometric properties of rock joints and to analyze the hydromechanical behavior within a rough joint considering the properties. Four pairs of joint surfaces were laser-scanned in order to obtain coordinates of the surfaces and then the coordinates were used to calculate the roughness, directional roughness, aperture, and spatial correlations. At the same time, fluid flow within a rough joint were simulated by a commercial FEM code, considering the variation of aperture space due to normal load. Flowrate, flow path, and channelization were investigated in an aperture scale. Since rock mass consists of several joints and/or joint sets, characterization of a single rock joint can be utilized for analyzing the behavior of rock mass as a reference.

Micro-electronic gas sensor devices were developed for the detection of carbon monoxide (CO), nitrogen oxides (NOx), ammonia (NH3), and formaldehyde (HCHO), as well as binary mixed-gas systems. Four gas sensing materials for different target gases, Pd-SnO2 for CO, In2O3 for NOx, Ru-WO3 for NH3, and SnO2-ZnO for HCHO, were synthesized using a sol-gel method, and sensor devices were then fabricated using a micro sensor platform. The gas sensing behavior and sensor response to the gas mixture were examined for six mixed gas systems using the experimental data in MEMS gas sensor arrays in sole gases and their mixtures. The gas sensing behavior with the mixed gas system suggests that specific adsorption and selective activation of the adsorption sites might occur in gas mixtures, and allow selectivity for the adsorption of a particular gas. The careful pattern recognition of sensing data obtained by the sensor array made it possible to distinguish a gas species from a gas mixture and to measure its concentration.