부추속은 수선화과에 속하며 약 1000여종에 이른다. 좀부 추는 자생 부추속에 속하며 식용이 가능하고 크기가 작아 분 화식물로 적합하다. 이에 따라 본 연구는 좀부추를 분화 소재 로서 개발하기 위하여 식물의 측지성장을 촉진하고 꽃의 품질 을 향상시키기 위해 수행되었다. 이를 위하여 좀부추의 구근 을 생장조절제(GA3, BA, 에테폰, TDZ)에 다른 시간(1, 5시 간)동안 침지처리했을 때 식물의 생장과 꽃에 어떤 영향을 주 는지 조사하였다. GA3 처리구는 엽수와 자구의 수가 증가해 증식을 촉진시키는 효과가 있었다. 게다가 GA3는 화수의 증 가와 조기개화를 유도했다. 그러나 엽폭과 화경의 폭은 감소 하였다. 에테폰은 좀부추의 초장과 초폭에서 영향을 주지 않 았으나 1h에서 자구의 수가 증가하였다. 에테폰은 화서 발생 을 억제하여 화경의 수가 감소하였다. BA는 식물의 생장과 개 화 특성에 유의적인 영향을 보이지 않았다. TDZ처리한 식물 은 왜화되고 생육상태가 불량하였으며, 식물을 고사시켰다.

본 논문에서는 웹툰 캐릭터 영상에 대해서 심층학습에 기반한 3D 안면 재구성 기술을 제안한다. 본 연구에서 제안하는 방법은 기본 사항 모듈과 상세 사항 모듈로 구성된다. 입력 받은 웹툰 캐릭터 영상에 대해서 기본 사항 모듈의 요소인 Albedo 모듈을 적용해서 안면에 들어오는 빛의 양을 계산하여 Albedo 맵을 생성한다. 그 리고 기본 사항 모듈의 다른 구성 요소인 FLAME 모듈에서는 입력 영상에 대한 기본적인 3D 안면 형태를 생 성한다. 이와 동시에 상세사항 모듈을 적용해서 실제 사람과 다르게 이목구비가 변형된 웹툰 캐릭터 영상의 표정이나 얼굴 깊이와 같은 특징을 살리는 세부사항을 추출한다. 계산한 세부사항들을 토대로 세부사항 맵을 생성하여 앞서 FLAME 모듈에서 생성된 3D 안면 형태와 결합하여 세부사항 안면 형태를 생성한다. 그 후 Albedo 모듈에서 생성된 Albedo 맵까지 적용하면 최종적으로 웹툰 캐릭터 영상에 대한 3D 안면 재구성이 완 료된다. 본 연구에서는 웹툰 캐릭터뿐만 아니라 안면이 스타일화된 애니메이 션 캐릭터에 대해서도 결과를 생성하고, 이를 기존 연구와 비교하여 그 우수성을 입증한다.

Roridin E, a constituent of Podostroma cornu-damae, has excellent medicinal properties and exhibits powerful anticancer effects. However, being a poisonous mushroom, P. cornu-damae is difficult to find in its natural environment. Therefore, only a limited amount of roridin E is isolated from P. cornu-damae. We isolated roridin E from mycelial culture medium of P. cornu-damae and examined its effects on five breast cancer cell lines. The purity of the isolated roridin E was 96%. It exhibited anticancer effects on all five breast cancer cell lines, with an IC50 of at least 0.0006 mg/L. To the best of our knowledge, this is the first report on the investigation of effects of roridin E isolated from the mycelium of P. cornu-damae mushrooms on five breast cancer cell lines. However, safety issues should be investigated and addressed in future studies before roridin E is used in clinical settings.

본 연구는 신생아 검사 중 포수클로랄(chloral hydrate)을 투여 후 진행되는 신생아 진정 검사 대비 진정 대체 방식 중 하나인 피드 및 랩(feed and wrap) 방식의 유용성을 평가한 연구이다. 본 연구에선 진정으로 진행한 신생아의 두뇌 T2 축면 영상과 피드 및 랩 방식으로 진행한 같은 영상 각 30개의 운동 허상(motion artifact)과 백질과 회백질의 구분 정도를 두 명의 영상의학과 전문의가 정성적으로 평가하였고, 운동 허상을 측정하기 위해서 위상부호화(phase encoding) 방향의 배경 영역(background area)의 평균 신호 강도(mean signal intensity)를 구하여서 정량적 방식으로 평가하였다. 또한 총검사 시간을 정리한 뒤 정량적 방식으로 평가하였고 투약 기록의 여부와 간호일지를 토대로 피드 및 랩 방식의 총 39건의 검사 건수 대비 성공률을 측정하였다. 운동 허상의 정량적 평가와 영상 품질의 정성적 평가 모두에서 두 집단은 유의미한 차이가 없었으나, 검사 시간의 정량적 평가에선 p값이 0.001로 유의한 차이가 있었다. 피드 및 랩 방식의 총검사 건수 대비 성공률은 100%였다. 결론적으로 본 논문에선 피드 및 랩 방식과 진정 방식의 영상 품질이 유의한 차이가 없고 성공률이 높기에 유용하다고 판단하였으나, 검사 시간이 더 지연되는 한계가 있다는 사실을 확인하였다.

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.

Radioactive waste generated during nuclear power plant decommissioning is classified as radioactive waste before the concentration is identified, but more than 90% of the amount generated is at a level that can be by clearance. However, due to a problem in the analysis procedure, the analysis is not carried out at the place of on-site but is transported to an external institution to identify concentration, which implies a problem of human error because 100% manual. As a way to solve this problem, research is underway to develop a mobile radioactive waste nuclide analysis facility. The mobile radionuclide analysis facility consists of a preparation room, a sample storage room, a measurement room, a pretreatment room, and a waste storage room, and is connected to an external ventilation facility. In addition, since the automation module is built-in from the sample pre-threatening step to the separation step, safety can be improved and rapid analysis can be performed by being located in the decommissioning site. As an initial study for the introduction of a mobile nuclide analysis facility, Visiplan, a preliminary external exposure evaluation code, was used to derive the analysis workload by a single process and evaluate the exposure dose of workers. Based on this, as a follow-up study, the amount of analysis work according to the continuous process and the exposure dose of workers were evaluated. As a result of the evaluation, the Regulatory dose limit was satisfied, and in future studies, internal and external exposure doses were evaluated in consideration of the route of movement, and it is intended to be used as basic data in the field introduction process.

According to IAEA PRIS, there is no record of dismantling commercial heavy water reactors among 57 heavy water reactors around the world. In Canada, which has the largest number of heavy water reactors, three of the 22 commercial heavy water reactors with more than 500 MW are permanently suspended, Gentilly unit 2 (2012), Pickering unit 2 (2007), and Pickering unit 3 (2008), all of which chose a delayed decommissioning strategy. On the other hand, Wolsong unit 1, which will be the world’s first heavy water reactor to be dismantled commercially, will be immediately carried out as a decommissioning strategy. KHNP has established various cooperation systems with advanced companies and international organizations related to overseas NPP decommission and is actively exchanging technologies. Among them, the most important focus is on research cooperation related to COG (CANDU owners Group). The first case is a joint study on Conceptual Calandria Segmentation. Four areas of process, waste management, ALARA, and cost for decommissioning reactors to be submitted to Canadian regulators for approval of Pickering and Gentilly-2’s preliminary decommissioning plan have been evaluated, and research on Wolsong unit 1 is currently underway. The second case is Decommissioning and long-term waste management R&D. Although the technical maturity is low, it studies the common interests of member companies in the decommissioning of heavy water reactor power generation companies and long-term waste management. Robotics for dismantling high-radiation structures, C- 14, H-3 measurement and removal methods, and concrete decontamination technology, which are characterized by heavy water, are being actively studied. KHNP is strengthening international cooperation with COG to prepare for the successful decommissioning of Wolsong unit 1. Based on previous studies by Pickering and Gentilly-2, an evaluation of the decommissioning of Wolsong unit 1 reactor is being conducted. In addition, it is preparing for decommissioning through experience analysis of the pressure tube replacement project.

KHNP is carrying out international technical cooperation and joint research projects to decommission Wolsong unit 1 reactor. Construction data of the reactor structures, experience data on the pressure tube replacement projects, and the operation history were reviewed, and the amount of dismantled waste was calculated and waste was classified through activation analysis. By reviewing COG (CANDU owners Group) technical cooperation and experience in refurbishment projects, KHNP’s unique Wolsong unit 1 reactor decommissioning process was established, and basic design of a number of decommissioning equipment was carried out. Based on this, a study is being conducted to estimate the worker dose of dismantling workers. In order to evaluate the dose of external exposure of dismantling workers, detailed preparation and dismantling processes and radiation field evaluation of activated structures are required. The preparation process can be divided into dismantlement of existing facilities that interfere with the reactor dismantling work and construction of various facilities for the dismantlement process. Through process details, the work time, manpower, and location required for each process will be calculated. Radiation field evaluation takes into account changes in the shape of structures by process and calculates millions of areas by process, so integrated scripts are developed and utilized to integrate input text data. If the radiation field evaluation confirms that the radiation risk of workers is high, mutual feedback will be exchanged so that the process can be improved, such as the installation of temporary shields. The results of this study will be used as basic data for the final decommissioning plan for Wolsong unit 1. By reasonably estimating the dose of workers through computer analysis, safety will be the top priority when decommissioning.

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

Wolsong unit 1, the first PHWR (Pressurized Heavy Water Reactor) in Korea, was permanent shut down in 2019. 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, activation assessment and waste classification of the End shield, which is a major activation component, were conducted. MCNP and ORIGEN-S computer codes were used for the activation assessment of the End shield. Radioactive waste levels were classified according to the cooling period of 0 to 20 years in consideration of the actual start of decommissioning. The End shield consists of Lattice tube, Shielding ball, Sleeve insert, Calandria tube shielding sleeve, and Embedment Ring. Among the components composed for each fuel channel, the neutron flux was calculated for the components whose level was not predicted by preliminary activation assessment, by dividing them into three channel regions: central channel, inter channel, and outer channel. In the case of the shielding ball, the neutron flux was calculated in the area up to 10 cm close to the core and other parts to check the decrease in neutron flux with the distance from the core. The neutron flux calculations showed that the highest neutron flux was calculated at the Sleeve insert, the component closest to the fuel channel. It was found that the neutron flux decreased by about 1/10 to 1/20 as the distance from the core increased by 20 cm. The outer channel was found to have about 30% of the neutron flux of the center channel. It was found that no change in radioactive waste level due to decay occurred during the 0 to 20 years cooling period. In this study, activation assessment and waste classification of End Shield in Wolsong unit 1 was conducted. The results of this study can be used as a basis for the preparation of the FDP for the Wolsong unit 1.

Radioactive contaminants, such as 137Cs, are a significant concern for long-term storage of nuclear waste. Migration and retention of these contaminants in various environmental media can pose a risk to the surrounding environment. The distribution coefficient (Kd) is a critical parameter for assessing the behavior of these contaminants and can introduce significant errors in predicting migration and remediation options. Accurate prediction of Kd values is essential to assess the behavior of radioactive contaminants and to ensure environmental safety. In this study, we present machine learning models based on the Japan Atomic Energy Agency Sorption Database (JAEA-SDB) to predict Kd values for Cs in soils. We used three different machine learning models, namely the random forest (RF), artificial neural network (ANN), and convolutional neural network (CNN), to predict Kd values. The models were trained on 14 input variables from the JAEA-SDB, including factors such as Cs concentration, solid phase properties, and solution conditions which are preprocessed by normalization and log transformation. We evaluated the performance of our models using the coefficient of determination (R2) value. The RF, ANN, and CNN models achieved R2 values of over 0.97, 0.86, and 0.88, respectively. Additionally, we analyzed the variable importance of RF using out-of-bag (OOB) and CNN with an attention module. Our results showed that the initial radionuclide concentration and properties of solid phase were important variables for Kd prediction. Our machine learning models provide accurate predictions of Kd values for different soil conditions. The Kd values predicted by our models can be used to assess the behavior of radioactive contaminants in various environmental media. This can help in predicting the potential migration and retention of contaminants in soils and the selection of appropriate site remediation options. Our study provides a reliable and efficient method for predicting Kd values that can be used in environmental risk assessment and waste management.

LILW disposal repository in Gyeongju, South Korea is considered with a concrete mixture that uses Ordinary Portland Cement (OPC) partially substituted with supplementary cementitious materials (SCMs). The degradation of cementitious materials that result from chemical and physical attacks is a major concern in the safety of radioactive waste disposal. We present a reactive transport model utilized as one of the geochemical simulation approaches for the timescales of concern that range from hundreds to thousands of years. The purpose of this study is to investigate the sensitivity of parameters in concrete disposal systems and to evaluate the influence of various assumptions on the chemical degradation of the systems using a reactive transport model. A reactive transport model in the concrete disposal vault was developed to evaluate the behavior of engineered barriers composed of cementitious materials. The sensitivity analysis was performed using reactive transport models through the coupling between COMSOL and PHREEQC. The databases selected for the analysis are the Thermochimie database presented by ANDRA. Among many variables considered, two variables that can highly affect chemical degradation were selected for detailed sensitivity analysis for dealing with uncertainties. This is important because the chemical degradation mechanism is generally sensitive to precipitation and diffusion coefficient. The first factor is precipitation, which might be the most important factor in chemical degradation because it acts as a calcium leaching of cementitious materials in a disposal system in a highly alkaline environment, increasing the porosity of the system. To predict the change in annual precipitation, the measurement of the precipitation observatory station in the nearest area of Gyeongju for the past 80 years was collected. The second factor is the diffusion coefficient, which plays an essential role in the durability of the concrete disposal system, promoting the decalcification of cementitious minerals, accelerating system degradation, and increasing the porosity of its system, thereby facilitating the migration of radionuclides. The diffusion coefficient values used in studies similar to this work were calculated and evaluated using the box-and-whisker method. The results of the sensitivity analyses for the reactive transport model in the concrete disposal system will be presented. The sensitivity cases show that the results obtained are much more sensitive to changes in transport parameters.

Domestic NPPs had produced the paraffin-solidifying concentrate waste (PSCW) for nearly 20 years. At that time radioactive waste management policy of KHNP was to reduce the volume and to store safely in site. The PSCW has been identified not to meet the leaching index after introducing the treatment system. PSCW has to be treated to meet current waste acceptance criteria (WAC) for permanent disposal. PSCW consists of dried concentrate 75% and paraffin 25% of volume. When PSCW is separated into a dried concentrate and a paraffin by solubility, total volume separated is increased twice. Final disposal volume of dried concentrate can reach to several times when solidifying by cement even considering exemption. Application of polymer solidification technology is difficult because dried concentrate is hard to make form to pellet. When PSCW is packaged in High Integrity Container (HIC), volume of PSCW is equal to the volume before package. The packaging process of HIC is simple and is no necessary of large equipment. It is important to recognize that HIC was developed to replace solidification of waste. HIC has as design goal a minimum lifetime of 300 years under disposal environment. The HIC is designed to maintain its structural integrity over this period, to consider the corrosive and chemical effects of both the waste contents and the disposal environment, to have sufficient mechanical strength to withstand loads on the container and to be capable of meeting the requirements for a Type A transport Package. The Final waste form is required for facilitating handling and providing protection of personnel in relation to solidification, explosive decomposition, toxic gases, hazardous material, etc. Structural stability of final waste form is required also. Structural stability of the waste can be provided by the waste itself, solidifying or placing in HIC. Final waste form ensure that the waste does not structurally degrade and affect overall stability of the disposal site. The HIC package contained PSCW was reviewed from several points of view such as physicochemical, radiological and structural safety according to domestic WAC. The result of reviewing shows that it has not found any violation of WCP established for silo type disposal facility in Gyeongju city.

The Ag0-containing sorbents synthesized by Na, Al, and Si alkoxides have higher maximum iodine capture capacity and textural properties than zeolite-based Ag0-containing sorbents. However, these sorbents were prepared in the form of granules via a step for cutting cylindrical alcogels. Since asmade sorbents decreased packing density, they must be additionally crushed and then classified into an appropriate size for increasing packing density. The bead formation in the step of sol-gelation could bring about the simplification of sorbent preparation process and an improvement of packing density. In the Na, Al, and Si alkoxides as starting materials, sol solution was hydrophilic and lower density than vegetable oil, which transformed sol droplets to sol-gel beads. Thus, in these precursors, sol droplets, which must be sprayed by single nozzle placed at bottom side of oil column, can rise up through oil column. Acetic acid (HOAc) was used as the catalyst for the hydrolysis of Na alkoxide (TEOS) and gelation of the Na+AlSi-OH alcosol. For obtaining sol-gel beads, experiments were performed by the flowrate change of sol solution and HOAc at different nozzle sizes using soybean oil column of 1 m in length. At a sol/HOAc flowrate ratio of 3.85, some Na+AlSi-OH alcogel beads were obtained. After the Ag/Na ion-exchange, Ag content in Ag+AlSi-OH hydrogel was low due to reaction between Na+ and HOAc during sol-gelation and aging step. The Ag+AlSi-OH hydrogel with high Ag content could be prepared by Na addition. After the solvent exchange and drying at ambient pressure, the bead sorbents had higher Ag0 content and larger pore size than granular sorbents. However, further experiments are needed to increase yield rate in bead sorbent.