Radioactive iodine-129, a byproduct of nuclear fission in nuclear power plants, presents significant environmental and health risks due to its high solubility in water and volatility. Iodine-129, with its half-life of 1.57×1017 years, necessitates safe management and disposal. Therefore, safely capturing and managing I-129 during spent nuclear fuel reprocessing is of paramount importance. To address these challenges, various glass waste forms containing silver iodide have been developed, such as borosilicate, silver phosphate, silver vanadate, and silver tellurite glasses. These glasses effectively immobilize iodine, but the high cost of silver raises affordability concerns. This study introduces CuI·Cu2O·TeO2 glass waste forms for iodine immobilization, a novel approach. The cost-effectiveness of copper, in contrast to silver, makes it an attractive alternative. The CuI·Cu2O·TeO2 glass waste forms were synthesized with varying CuI content (x) in (1-x)(0.3Cu2O·0.7TeO2) glass matrices. Xray diffraction (XRD) confirmed amorphous structures, and X-ray fluorescence (XRF) quantified composition. X-ray photoelectron spectroscopy (XPS) and Raman spectroscopy provided insights into structural properties. Durability assessments using a 7-day product consistency test (PCT-A) and inductively coupled plasma-mass spectrometry (ICP-MS) revealed compliance with U.S. glass regulations, making CuI·Cu2O·TeO2 glasses a promising choice for iodine immobilization in radioactive waste.

South It is necessary to develop the future technologies to improve the sustainability and acceptability of nuclear power plants generation. Currently, our company is preparing to build the dry storage facility on-site in accordance with the basic plan for managing high-level radioactive waste announced by the government in 2021. However, studies on technologies for the volume reduction of spent nuclear fuel to increase the efficiency of on-site spent fuel dry storage facilities are very not enough. Accordingly, in this study, the storage efficiency and appropriateness for the SF volume reduction processing technologies such as SF oxide processing technology and consolidation technology are evaluated. Finally, the goal is to develop the optimized technologies to improve the storage efficiency of spent nuclear fuel. As a result in this study is followings. [Safety] After removing volatile fission products (Xe, Kr, I, etc.), Xe, Kr, etc. are removed during storage of the sintered structures. UO2 has a high melting point of approximately 1,000°C after cesium (Cs) has been removed, and heat can be removed by natural convection. [Economy]1999 DUPIC unit facility unit price reference, 2020 standard 328 $/kg estimated. A Comprehensive Approach Considering the Whole System is needed. Benefit from replacement and continuous operation of metal storage containers. Changes in economic efficiency obtained in conjunction with fluctuations in electricity prices and disposal. [Waste filter] A separated solidification facility high-level waste filter is required, and overseas outsourcing must be considered. [Waste cladding]. Cannot be accommodated in low-level disposal site. This reason is why the Ni nuclides occur to be in bulk. [Metal structural material] It is possible to reduce the initial volume by 7.6% or more when compressed or melted, but the technology needs to be advanced. [Oxide blocks] Larger size and density are expected to improve storage and disposal efficiency. [Facilities operation waste] Expected to be able to be disposed of at mid-to-low level decommissioning sites in Gyeongju city. [Solidified volatile nuclides and activated metals] Expected to improve storage efficiency when used volume is reduced and stored, such as outsourced reprocessing. [Oxide block] Radioactivity and decay heat are estimated to be reduced by half during oxide treatment. 75% reduction in volume and 40% reduction in storage area compared to used nuclear fuel before treatment. [Merits/Shortages] Improvement of storage and disposal efficiency empirical research such as large-capacity [real-scale] oxide block production is required. Oxide processing facilities are likely to be classified as post-use nuclear fuel processing facilities. It is determined that additional documents such as a Radiation Environmental Report (RER) must be submitted. Existence of possible external leaks of glass, highly mobile radionuclides from the point of view of nuclear criticality and heat removal. Acceptancy requirements of citizens in the process of creating additional sites for oxide treatment facilities. Considering social public opinion, it is necessary to secure the acceptability such as residents’ opinions convergence. Characteristics of high nuclear non-propagation compared to other processing technologies involving chemical processing. Also, Expectation of volume reduction effect for spent nuclear fuel itself. Volume reduction methods for solid waste and gaseous waste are required.

초파리를 당뇨병을 연구하는데 사용하는 경우가 많으나 당뇨병이 유전적 요인에 작용하여 다음 세대에 미치 는 영향을 다루고 있는 연구는 적다. 따라서 본 연구에서는 초파리에 고당도의 먹이를 통해 당뇨병을 유발하고, 류신(leucine)의 당뇨병에 대한 치료효과를 확인하고 다음 세대에 미치는 영향을 알아보고자 했다. 본 연구에서 는 Drosophila melanogaster를 초파리 모델에 사용했다. 먹이에 고농도의 설탕, 고농도의 설탕과 류신을 첨가하여 당뇨의 발병과 당뇨병의 치료효과를 확인했다. 당뇨병의 발병을 확인하기 위해 초파리의 체내의 포도당을 측정 하여 대조군에 비해서 고농도의 설탕을 섭취한 경우에 포도당의 농도는 증가했고, 류신을 섭취한 경우에는 포도 당이 대조군에 비해서도 급격히 떨어지는 것을 확인했다. 고농도의 설탕이 포함된 먹이 조건을 유지하고 교배를 통해 자손을 얻고자 하였다. 그 결과, 고농도의 설탕을 섭취한 경우 자손의 몸의 크기가 0.1mm 정도 감소하고 무게 또한 감소했다. 그러나 고농도의 설탕과 류신을 섭취한 경우 자손을 얻지 못했다. 당뇨병을 유발시킨 초파리 를 이용해 류신과 운동의 효과를 동시에 확인했다. 당뇨병이 유발된 경우 운동만으로는 포도당의 감소에 영향을 주지 못하였으나, 류신 운동을 병행한 경우에는 포도당이 감소했고 수컷 초파리에서 잘 관찰되었다. 운동과 류신 섭취를 병행하는 실험에서는 당뇨병을 유발한 초파리의 자손을 사용했기 때문에 크기는 당뇨병을 유발하지 않은 것보다 작았다. 따라서 초파리의 당뇨병 모델을 통해서 당뇨병이 유전적으로 전달되며 Leucine에 의한 치료 는 성충에 수행되는 것이 효과적임을 알 수 있었다.

Titanium, which has excellent strength and toughness characteristics, is increasingly used in the aerospace field. Among the titanium alloys used for body parts, more than 80 % are Ti-6Al-4V alloys with a tensile strength of 931 MPa. The spark plasma sintering (SPS) method is used for solidification molding of powder manufactured by the mechanical milling (MM) method, by sintering at low temperature for a short time. This sintering method avoids coarsening of the fine crystal grains or dispersed particles of the MM powder. To improve the mechanical properties of pure titanium without adding alloying elements, stearic acid was added to pure titanium powder as a process control agent (PCA), and MM treatment was performed. The properties of the MM powder and SPS material produced by solidifying the powder were investigated by hardness measurement, X-ray diffraction, density measurement and structure observation. The processing deformation of the pure titanium powder depends on the amount of stearic acid added and the MM treatment time. TiN was also generated in powder treated by MM 8 h with 0.50 g of added stearic acid, and the hardness of the powder was higher than that of Ti-6Al-4V alloy when treated with MM for 8 h. When the MM-treated powder was solidified in the SPS equipment, TiC was formed by the solid phase reaction. The SPS material prepared as a powder treated with MM 8 h by adding 0.50 g of stearic acid also formed TiN and exhibited the highest hardness of Hv1253.

The decommissioning of Korea’s nuclear power facilities is expected to take place starting with the Kori Unit 1 followed by the Wolsong Unit 1. In Korea, since there is no experience of decommissioning, considerations of site selection for the waste treatment facilities and reasonable selection methods will be needed. Only when factors to be considered for construction are properly selected and their effects are properly analyzed, it will be possible to operate a treatment facility suitable for future decommissioning projects. Therefore, this study aims to derive factors to be considered for the site selection of treatment facilities and present a reasonable selection methodology through evaluation of these factors. In order to select a site for waste treatment facilities, three virtual locations were applied in this study: warehouse 1 to warehouse 3. Such a virtual warehouse could be regarded as a site for construction warehouses, material warehouses, annexed building sites, and parking lots in nuclear facilities. If the selection of preliminary sites was made in the draft, then it is necessary to select the influencing factors for these sites. The site of the treatment facility shall be suitable for the transfer of the waste from the place where the dismantling waste is generated to the treatment facility. In addition, in order for construction to take place, interference with existing facilities and safety should not be affected, and it should not be complicated or narrow during construction. Considering the foundation and accessibility, the construction of the facility should be economical, and the final dismantling of the facility should also be easy. In order to determine one final preferred plan with three hypothetical locations and five influencing factors, there will be complex aspects and it will be difficult to maintain consistency as the evaluation between each factor progresses. Therefore, we introduce the Analytic Hierarchical Process (AHP) methodology to perform pairwise comparison between factors to derive an optimal plan. One optimal plan was selected by evaluating the three virtual places and five factors of consideration presented in this study. Given the complexity and consistency of multiple influencing factors present and prioritizing them, AHP tools help users make decisions easier by providing simple and useful features. Above all, it will be most important to secure sufficient grounds for pairwise comparison between influencing factors and conduct an evaluation based on this.

The intermediate level spent resins waste generated from water purification for the the moderator and primary heat transport system during operaioin of heavy water reactor (HWR). Especially, moderator resins contain high level activity largely because of their C-14 content. So spent resins are considered as a problematirc solid waste and require special treatment to meet the waste acceptance criteria for a disposal site. Various methods have been studied for the treatment of spent resins which include thermal, destructive, and stripping methods. In the case of solidification methods, cement, bitument or organic polymers were suggested. In the 1990s, acid stripping using nitric acid and thermal treatment methods were actively investigated in Canada to remove C-14 nuclide from waste resin. In Japan, thermal distructive method was studied in the 1990s. Since 2005, KAERI developed acid stripping method using phosphate salt. However, acid stripping method are not suitable due to large amounts of 2nd waste containing acid solution with various nuclides. To solve this probelm, KAERI has been suggested the microwave treatment method for C-14 selective removal from waste resin in the 2010s. Pilot scale demonstration tests using radioactive waste resin generated from Wolsung unit 1 and unit 2 were successfully conducted and 95% of C-14 was selectively removed from the radioactive waste resin. In recent years, price of C-14 source is dramatically increased due to market growth of C-14 utilization and exclusive supply chain depending on China and Russia. High purity of C-14 were captured in HWR waste resin. Interest of C-14 recovery research from HWR waste resin is currently increased in Canada. In this study, microwave method is suggested to treat HWR waste resin with C-14 recovery process. Additionally, status of waste resin management and research trends of HWR waste resin treatment are introduced.

A large amount of small and medium-sized metal waste is generated during the decommissioning of nuclear power plants (NPPs). Metal waste is mostly contaminated with low-level radioactive, so it needs decontamination for self-disposal and recycling. A large amount of Organic Decontamination Liquid Waste during decontamination will be generated. The generated organic liquid waste is low in concentration, so the decomposition efficiency is low in the decomposition process. A conditioning process is necessary to concentrate at a high concentration. For effective treatment for Organic Decontamination Liquid Waste, the composition of organic liquid waste and conditioning process were analyzed. Organic acids, metal ions, radioactive nuclides, surfactants, etc. are present in the Organic Decontamination Liquid Waste, and suspended solids are sometimes generated by various reactions. According to previous studies, the concentration of organic acids including surfactants obtained results from several tens of ppm to a maximum of 1,000 ppm, so the maximum value of 1,000 ppm was assumed. For the composition and total amount of metal ions, the average value (52.7wt% Fe, 16.3wt% Ni, 15.1wt% Cr, 15.9wt% Mn) of the distribution of metal species removed by the actual decontamination process is applied, and the total amount is 1,000 ppm was assumed. As for the radionuclides, only 60Co and 137Cs, which are expected to be mainly present, were considered, and 60Co was assumed to be 2,000 Bq/g and 137Cs to be 360 Bq/g by referring to the literature. The amounts of suspended solids were assumed to be 500 ppm by referring to the characteristics of the liquid waste generated in the decontamination process of the NPPs. Based on the estimated value, a reaction formula was established and a simulated Organic Decontamination Liquid Waste was prepared. As a result of measurement using an analysis device, the composition of the estimated and simulated Organic Decontamination Liquid Waste had similar values. The conditioning and treatment process largely consists of pretreatment, conditioning, decomposition processes. Organic Decontamination Liquid Waste goes through a pretreatment process to remove impurities with large particles. In the conditioning process, treated water that has passed through the UF/RO membrane system is discharged into the environment. At this time, Concentrated water goes through a decomposition process for processing the Organic Decontamination Liquid Waste, and is discharged to the environment through a secondary RO membrane system. The conditioning process is the low-concentration Organic Decontamination Liquid Waste in the UF membrane system is forming a micelles in an RO membrane system, concentrating it to a high concentration and then go through a recirculation process in the UF membrane system. An experiment was conducted to confirm whether the concentration of surfactants occurred during the conditioning process. As a result of the experiment confirmed that the highly concentrated surfactant formed micelles and was filtered out in the UF membrane system.

Most of the spent nuclear fuel generated by domestic nuclear power plants (NPPs) is temporarily stored in wet storage which is spent fuel pool (SFP) at each site. Currently, in case of Kori Unit 2, about 93.6% of spent nuclear fuel is stored in SFP. Without clear disposal policy determined for spent nuclear fuel, the storage capacity in each nuclear power plant is expected to reach saturation within 2030. Currently, the SFP stores not only spent fuel but also various non-fuel assembly (NFA). NFA apply to all device and structures except for fuel rods inserted in nuclear fuel assembly. The representative NFA is control element driving mechanism (CEDM), in-core instrument (ICI), burnable poison, and neutral resources. Although these components are irradiated in the reactor, they do not emit high-temperature heat and high radiation like nuclear fuel, so if they are classified as intermediate level waste (ILW) and low level waste (LLW) and moved outside the SFP, positive effects such as securing spent fuel storage space and delaying saturation points can be obtained. Therefore, this study analyzes the status of spent fuel and Non Fuel Assembly (NFA) storage in SFP of domestic nuclear power plants. In addition, this study predict the amount of spent fuel and NFA that occur in the future. For example, this study predicts the percentage of current and future ICIs and control rods in the SFP when stored in the spent fuel storage rack. In addition, the positive effects of moving NFA outside the SFP is analyzed. In addition, NFA withdrawn from SFP is classified as ILW & LLW according to the classification criteria, and the treatment, storage, and disposal methods of NFA will be considered. The study on the treatment, storage, and disposal methods of NFA is planned to be conducted by applying the existing KN-12 & KN-18 containers and ILW & LLW containers being developed for decommissioning waste.

Decommissioning of Nuclear Power Plant (NPP) projects in South Korea starts with permanent shutdown of Kori unit 1 and Wolsung unit 1. It is important to establish a treatment and disposal method for radioactive waste generated during the decommissioning of the nuclear power plants. Large quantities of the wastes during decommissioning of NPP are generated in a short period of time and the wastes have various types and characteristics. For efficient decommissioning of NPP process, the radioactive waste is classified by types and each treatment method and packaging concept is presented respectively in this paper. Radioactive waste generated during decommissioning of NPP is classified into reactor vessel, reactor internals, metals, Dry Active Waste (DAW), concreate, spent fuel storage rack, spent resin and spent filter, etc., and the packaging concept for each type should be established to meet the waste acceptance criteria. Major waste acceptance criteria requirements include nuclides concentration, filling rate, free water, surface radiation does rate and weight. Radioactive waste containers can be classified into packaging containers, transport containers, and disposal containers. The packaging container is used to contain, transport, and store radioactive waste within the radiation control area, and a control number has been assigned as a radioactive waste drum after the final treatment has been completed. The transport container is used for transporting radioactive waste filled-containers from a radiation control area through an uncontrolled area. In this paper, the concept of disposal of dismantled radioactive waste and packaging methods were reviewed in comprehensive consideration of domestic radioactive waste transport and storage regulations, permanent disposal environment, and development status of large containers.

Kori unit 1 was permanently shut down in 2007 and is currently awaiting approval for decommissioning and dismantling (D&D). The wastes generated during decommissioning is estimated to be approximately 14,500 of 200 L drums. In this study, the treatment process of decommissioning wastes will be reviewed through the case of the US Zion nuclear power station (ZNPS). Zion unit 1 and 2 received an operating license in 1973 and were permanently shut down and the spent nuclear fuel was transferred to the pool in 1998. The decommissioning was carried out according to the following five steps; (1) safe storage (SAFSTOR) dormancy, (2) preparation for decommissioning, (3) establishment of independent spent fuel storage installation (ISFSI) and transfer of the spent fuel and greater than class C radioactive materials, (4) decommissioning operations and (5) site restoration. The total volume of waste generated during decommissioning was expected to be approximately 1.7×105 m3. This is far above the Kori unit 1 waste estimation because ZNPS has a history of accidents and includes soil waste. Wastes were treated differently according to their properties and locations.

Inorganic and organic ion exchange materials were generally applied to liquid processes in nuclear reactor. In the case of heavy-water reactor (HWR), zeolite, active carbon, anion resin, and cation resin were used to treat liquid processes such as reactor primary coolant cleanup and liquid radioactive waste management system. Then, used ion exchangers were stored at storage tanks. Various kinds of nuclides were adsorbed in ion exchange materials. Especially, C-14, long half-life nuclide, was highly concentrated in anion resin, and waste resin was treated as intermediated level radioactive waste (ILW). Thermal and non-thermal methods such as pyrolysis, incineration, catalytic extraction, acid digestion, and wet oxidation have been studied for treating spent resin. However, destructive methods are not suitable due to massive off gas waste containing radioactive species. To solve this problem, various kinds of processes were developed such as acid stripping, PLO process, activity stripping, thermal treatment, and etc. In this study, microwave method is suggested to treat HWR waste resin. C-14 nuclide was selectively removed from waste resin without decomposition of main structure in waste resin. Radioactive waste resin generated from Wolsung HWR unit 1 and unit 2 was treated using microwave method and 95% of C-14 was successfully removed from the radioactive waste resin.

동물플랑크톤이 식물플랑크톤을 선택적으로 섭식하는 특성에 대한 이해는 수생태계 먹이사슬 내의 물질 이동에 중요하다. 하지만 해부를 통한 위내용물 추출 방법은 소형 요각류를 대상으로 적용하기에는 적절하지 않고, 유전자가 유실되거나 위내용물이 아닌 개체 외부의 유전자로 인해 오염될 가능성이 존재한다. 본 연구에서 호내 식물 플랑크톤 조성 및 기타 환경이 상이한 두 지점을 선정하여 모든 지점에서 지속적으로 출현하는 기수성 요각류인 Sinocalanus tenellus를 대상으로 위내용물의 유전자 분석을 수행하였다. 요각류 개체 외부의 DNA를 제거하는 데 2.5%로 희석한 시판용 표백제(차아염소산나트륨 5.4%) 에 2분간 처리하여 증류수로 2회 세척한 뒤 유전자를 추출하였다. 추출된 유전자는 23S rRNA을 증폭하여 서열분석을 실시하였다. Capillary sequencing 분석 결과, 원수와 처리수 및 요각류 위내용물에서 다양한 분류군 (규조강, 녹조 강, 남조강, 와편모조강, 은편모조강, 황갈조강)에 속하는 식물플랑크톤이 검출되었으며, 새만금호 내 시공간 차이에 따라 상이한 경향을 보였다. 현미경을 이용하여 동정한 식물플랑크톤 군집 조성의 경우 규조강이 우점한 반면, 동일한 원수의 유전자 분석 (capillary sequencing) 결과에서는 주로 녹조강, 남조강 및 와편모조강이 우점하여 다소 상반된 경향을 나타냈다. 본 연구에서 적용한 위내용물 분석에 특화된 외부 유전자 제거 전처리 방법은 농도와 처리시간 조절 등의 응용방법에 따라 다양한 동물플랑크톤 분류군에 적용이 가능할 것으로 사료된다.

본 연구는 갈색날개매미충의 기계유유제 처리 방법별 부화율과 산란에 의한 사과 열매의 품질 변화에 대해 조사하였다. 기계유유제 처리 효과를 보면 기계유유제 20배를 도포한 것이 평균 0.57%로 가장 적은 부화율을 보였고, 분무한 가지에서는 평균 1%의 부화율을 보였다. 기계유유제를 50배 도포 처리시 부화율이 약 35%를 보인반면 분무처리는 약 77%를 보여 편차를 고려하면 무처리와 차이가 없는 것으로 보인다. 홍로와 후지 품종에서 갈색날개매미충이 산란한 결과지와 산란되지 않은 결과지에 이듬해 사과 열매가 결실되어도 과실의 품질 차이는 통계적 유의성은 없었다. 또한 갈색날개매미충의 산란에 의한 가지의 부러짐도 없었고, 결과지 생육도 통계적 유의성은 없었다.