단단한 자구를 가진 적색 비모란선인장 ‘Gangjeok’ 품종 은 ‘Isaek’품종을 모본으로, ‘Suyeon’ 품종을 부본으로 하여 2018년에 교배하여 육성하였다. 교배 후 획득한 종자는 조직 배양실에서 기내파종하여 획득한 유묘를 기내에서 삼각주선 인장에 접목하여 ‘1802001’ 등 20계통을 양성하였다. 2019 년에 기내에서 양성한 20계통을 온실에서 삼각주선인장 대목 에 접목하여 재배하면서 ‘1812005’ 계통을 1차 선발하였다. 2020년부터 2022년까지 3차에 걸쳐서 특성을 검정한 후, 농 산물직무육성품종 심의회에서 최종 선발하여 ‘Gangjeok’으 로 명명하였다.‘Gangjeok’ 품종은 편원형의 적색 구를 가진 다. 혹(tubercle)이 돌출된 형태의 모구는 8.4개의 능(rip)을 가지며, 3.5mm 짧은 회색 가시가 발생한다. 정식 10개월 후 ‘Gangjeok’ 품종의 직경은 46.1mm이며, 자구는 평균 18.3 개 발생한다. 2022년 육성계통 평가회에서 ‘Gangjeok’ 품종 은 높은 기호도 점수 4.0을 받았다.

주방세제의 진딧물 살충효과와 분국화 생육에 미치는 영향 을 알아보기 위해 실험을 수행하였다. 복숭아혹진딧물과 목화 진딧물을 대상으로 주방세제 100, 200, 400배액 단용처리와 주방세제 400배액 100mL에 소주 10mL 또는 20mL를 혼용 하여 분무 처리하였다. 그 결과 주방세제 400배 이하의 모든 농도에서 복숭아혹진딧물은 85%, 목화진딧물은 90%의 살충 률을 보였다. 주방세제와 소주 혼용처리에 의한 살충효과 상 승은 보이지 않았다. 하우스안에서 진딧물이 많이 발생한 분 국화 잎에 400배액 이하의 농도로 3일 간격으로 2회 처리함 으로써 90% 이상의 진딧물이 감소되는 효과를 얻을 수 있었 다. 또한 분국화를 재배하면서 3~4일 간격으로 5주간 지속적 으로 주방세제를 처리했을 때, 대조구에서는 90% 잎에서 진 딧물이 발생한데 비해, 모든 주방세제 단독 처리구에서는 발 생율이 15% 이하로 낮게 유지되었다. 또한 국화의 생체중, 초 장, 엽수는 주방세제 농도가 높아질수록 조금씩 낮아지는 경 향이었으나, 초폭이나 분지수에서는 차이가 없었다. 그러나 주방세제 200배 이하의 농도에서는 국화의 잎이 갈변되거나 꽃잎 끝이 백화되는 약해가 관찰되었다.

This study assessed the utility of netted melon ‘Top Earl’s’ and cantaloupe melon ‘Alex’ as functional fruits by analysing their moisture content, vitreous sugar, folic acid, citric acid, and beta-carotene levels. High-performance liquid chromatography (HPLC) was used to analyse the free sugar, folic acid, citric acid, and beta-carotene levels. The moisture content was not significantly different between ‘Top Earl’s’ and ‘Alex.’ The glucose, sucrose, and fructose contents were three, two, and one-and-a-half fold higher in ‘Alex’ than in ‘Top Earl’s.’ Moreover, citric acid was approximately three times higher in ‘Alex’ than that in Top Earl’s.’ However, the folic acid content was higher in ‘Top ‘Earl’s’ than ‘Alex,’ and the amount was 124 μg / 100 g FW and 112 μg / 100 g FW respectively. ‘Beta-carotene was undetectable in ‘Top Earl’s,’ whereas it was 1000 μg / 100 g FW in ‘Alex.’ β-carotene, a substance that is converted in the body into vitamin A and acts as an antioxidant, is an important component in healthy food. These results suggested that the cantaloupe melon ‘Alex’ has a higher free sugar content and functional ingredients, such as antioxidants, including citric acid and beta carotene, than the netted melon ‘Top Earl’s.’

Porous carbons are considered promising for CO2 capture due to their high-pressure capture performance, high chemical/ thermal stability, and low humidity sensitivity. But, their low-pressure capture performance, selectivity toward CO2 over N2, and adsorption kinetics need further improvement for practical applications. Herein, we report a novel dual-templating strategy based on molten salts (LiBr/KBr) and hydrogen-bonded triazine molecules (melamine–cyanuric acid complex, MCA) to prepare high-performance porous carbon adsorbents for low-pressure CO2. The comprehensive investigations of pore structure, microstructure, and chemical structure, as well as their correlation with CO2 capture performance, reveal that the dual template plays the role of porogen for multi-hierarchical porous structure based on supermicro-/micro-/meso-/ macro-pores and reactant for high N/O insertion into the carbon framework. Furthermore, they exert a synergistic but independent effect on the carbonization procedure of glucose, avoiding the counter-balance between porous structure and hetero-atom insertion. This enables the preferred formation of pyrrolic N/carboxylic acid functional groups and supermicropores of ~ 0.8 nm, while retaining the micro-/meso-/macro-pores (> 1 nm) more than 60% of the total pore volume. As a result, the dual-templated porous carbon adsorbent (MG-Br-600) simultaneously achieves a high CO2 capture capacity of 3.95 mmol g− 1 at 850 Torr and 0 °C, a CO2/ N2 (15:85) selectivity factor of 31 at 0 °C, and a high intra-particle diffusivity of 0.23 mmol g− 1 min− 0.5 without performance degradation over repeated use. With the molecular scale structure tunability and the large-scale production capability, the dual-templating strategy will offer versatile tools for designing high-performance carbon-based adsorbents for CO2 capture.

Graphene-derived materials are an excellent electrode for electrochemical detection of heavy metals. In this study, a MnO2/ graphene supported on Ni foam electrode was prepared via ultrasonic impregnation and electrochemical deposition. The resulting electrode was used to detect Pb(II) in the aquatic environment. The graphene and MnO2 deposited on the Ni foam not only improved active surface area, but also promoted the electron transfer. The electrochemical performance towards Pb(II) was evaluated by cyclic voltammetry (CV) and square wave anodic stripping voltammetry (SWASV). The prepared electrode exhibited lower limit of detection (LOD, 0.2 μM (S/N = 3)) and good sensitivity (59.9 μAμM−1) for Pb(II) detection. Moreover, the prepared electrodes showed good stability and reproducibility. This excellent performance can be attributed to the strong adhesion force between graphene and MnO2, which provides compact structures for the enhancement of the mechanical stability. Thus, these combined results provide some technical considerations and scientific insights for the detection of heavy metal ions using composite electrodes.

Natural uranium-contaminated soil in Korea Atomic Energy Research Institute (KAERI) was generated by decommissioning of the natural uranium conversion facility in 2010. Some of the contaminated soil was expected to be clearance level, however the disposal cost burden is increasing because it is not classified in advance. In this study, pre-classification method is presented according to the ratio of naturally occurring radioactive material (NORM) and contaminated uranium in the soil. To verify the validity of the method, the verification of the uranium radioactivity concentration estimation method through γ-ray analysis results corrected by self-absorption using MCNP6.2, and the validity of the pre-classification method according to the net peak area ratio were evaluated. Estimating concentration for 238U and 235U with γ-ray analysis using HPGe (GC3018) and MCNP6.2 was verified by 􀟙-spectrometry. The analysis results of different methods were within the deviation range. Clearance screening factors (CSFs) were derived through MCNP6.2, and net peak area ratio were calculated at 295.21 keV, 351.92 keV(214Pb), 609.31 keV, 1120.28 keV, 1764.49 keV(214Bi) of to the 92.59 keV. CSFs for contaminated soil and natural soil were compared with U/Pb ratio. CSFs and radioactivity concentrations were measured, and the deviation from the 60 minute measurement results was compared in natural soil. Pre-classification is possible using by CSFs measured for more than 5 minutes to the average concentration of 214Pb or 214Bi in contaminated soil. In this study, the pre-classification method of clearance determination in contaminated soil was evaluated, and it was relatively accurate in a shorter measurement time than the method using the concentrations. This method is expected to be used as a simple pre-classification method through additional research.

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.

Kori Nuclear Power Plant Unit 1, which began operating in 1978, is Korea’s oldest commercial nuclear reactor. The reactor was permanently shut down in June 2017, and now the decommissioning process has begun. The decommissioning process will generate a significant amount of waste that requires appropriate management to minimize the impact on the environment and human health. And the waste routing, i.e. the activities and logistics for managing the material generated, is a key point in a decommissioning project. It determines the routes from the material inventory to the envisaged material end states. In this study, we review on several factors for the selection of the waste routes in a decommissioning project. In terms of sustainability, the ‘waste hierarchy’ should be applied to routing materials from nuclear facilities. According to the waste hierarchy, the preferred end state is reuse or recycling of the waste as material or, more preferably, the avoidance of waste generation. In addition, treatments (such as decontamination and thermal treatment) that can reduce the volumes requiring disposal as radioactive waste should be considered. Another important parameter is the need to secure availability and capacity of waste routes. Short-term bottlenecks or any delay in the removal of the waste from the site often has an impact on other site activities. If possible, at least two alternative waste routes should be identified for the main categories of waste and kept available throughout the decommissioning project. All routes should be direct to the material end state if possible, but it is more important that waste is removed from the site so that other site operations are not impeded.

Prevention of radiation hazards to workers and the environment in the event of decommissioning nuclear power plants is a top priority. To this end, it is essential to continuously perform radiation characterization before and during decommissioning. In operating nuclear power plants, various detectors are used depending on the purpose of measurement. Portable detectors used in power plants have excellent portability, but there is a limit to the use of a single measuring device alone to quantify radioactive contamination, nuclide analysis, and ensure representation of measurement results. In foreign countries, gamma-ray visualization detectors are being actively used for operating and decommissioning nuclear power plants. KHNP is also conducting research on the development of gamma-ray visualization detectors for multipurpose field measurement at decommissioning nuclear power plants. It aims to develop detectors capable of visualizing radioactive contamination, analyzing nuclides, estimating radioactivity, and estimating dose rates. To this end, we are developing related software according to the development process by purchasing sensors from H3D, which account for more than 75% of the US gamma-ray visualization detector market. In addition, field tests are planned in the order of Wolsong Unit 1 and Kori Unit 1 with Research reactor in Gongneung-dong in accordance with the progress of development. The detector will be optimized by analyzing the test results according to various gamma radiation field environments. The development detector will be used for various measurement purposes for Kori unit 1 and Wolsong

For decontamination and quantification of trace amount of tritium in water, an efficient separation technology capable of enriching tritium in water is required. Electrolysis is a key technology for tritium enichment as it has a high H/T and D/T separation factors. To separate tritium, it is important to develop a proton exchange membrane (PEM) electrolyzer having high hydrogen isotope separation factor as well as high electrolyzer cell efficiency. However, there has not been sufficient research on the separation factor and cell efficiency according to the composition and manufacturing method of the membrane electrode assembly (MEA) Therefore, it is necessary to study the optimal composition and manufacturing method of the MEA in PEM electrolyzer. In this study, the H/D separation factor and water electrolysis cell efficiency of PEM electrolyzer were analyzed by changing the anode and cathode materials and electrode deposition method of the MEA. After the water electrolysis experiment using deionized water, the D/H ratio in water and hydrogen gas was measured using a cavity ring down spectrometer and a mass spectrometer, respectively, and the separation factor was calculated. To calculate the cell efficiency of water electrolysis, a polarization curves were obtained by measuring the voltage changes while increasing the current density. As a result of the study, the water electrolyzer cell efficiency of the MEA fabricated with different anode/cathode configurations and electrode formation methods was higher than that of commercial MEA. On the other hand, the difference in H/D separation factor was not significant depending on the MEA fabrication methods. Therefore, using a cell with high cell efficiency when the separation factor is the same will help construct a more efficient water electrolysis system by lowering the voltage required for water electrolysis.

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.

Air conditioning facilities in nuclear power plants use pre-filters, HEPA filters, activated carbon filters, and bag filters to remove radionuclides and other harmful substances in the atmosphere. Spent filters generate more than 100 drums per year per a nuclear power plant and are stored in temporary radioactive waste storage. Plasma torch melting technology is a method that can dramatically reduce volume by burning and melting combustible, non-flammable, and mixed wastes using plasma jet heat sources of 1,600°C or higher and arc Joule heat using electric energy, which is clean energy. KHNP CRI & KPS are developing and improving waste treatment technology using MW-class plasma torch melting facilities to stably treat and reduce the volume of radioactive waste. This study aims to develop an operation process to reduce the volume of bag filter waste generated from the air conditioning system of nuclear power plants using plasma torch melting technology, and to stably treat and dispose of it. It is expected to secure stability and reduce treatment costs of regularly generated filter waste treatment, and contribute to the export of radioactive waste treatment technology by upgrading plasma torch melting technology in the future.

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.

A lot of solid wastes are generated when nuclear power plant is dismantled, and a lot of treatment costs and optimal waste treatment technologies are required to treat the generated solid wastes. Currently, there is no optimized reduction and solidification technology for each characteristics of radioactive dismantling waste, so the customized treatment technology for each waste is required to respond actively to this issue. This paper shows the evaluation results of molding and sintering characteristics using preliminary sample to derive operational characteristics and improvements for powder mixing device, molding device, and sintering device manufactured for solidification of dispersible radioactive waste. Zeolite was used as a preliminary sample for performing basic operation characteristics evaluation of each unit device. First of all, the basic operation characteristics of the powder mixing device was evaluated by analyzing the sample distribution, mixing degree, and tap density. It was confirmed that the preliminary sample was well mixed in all areas of the cylinder where the mixing was performed. In the tap density analysis, the increase effect of the volume reduction of the sample was confirmed according to the increase of the RPM speed (up to 2000 RPM). Since the particle size of zeolite sample is very small (nanometer size), the particular consistency of the change of average particle size with RPM speed couldn’t be confirmed, but the uniform of particle size distribution was confirmed with RPM speed size. The basic operation characteristics of the molding device was evaluated for each mold size (ID30, ID50, ID100) according to the moisture content (0-20%) and the molding pressure condition (25-200 MPa) for the preliminary sample. In the characteristics evaluation of the sintered body, the strength of the sintered body was much higher than that of the molded body. However, it was confirmed that as moisture evaporated during the sintering process according to the moisture content contained in the molded body, the swelling occurred in the sintered body due to vapor pressure, and this caused cracks in the longitudinal or transverse direction inside and outside the sintered body. Therefore, optimal moisture content conditions for sintering should be derived. In conclusion, if the operation characteristics and improvements of powder mixing, molding and sintering devices derived from this study are reflected and improved, it is judged that it is possible to derive the optimal process for solidification of dispersive radioactive wastes.