Background: Cadmium (Cd) is toxic heavy metal that accumulates in organisms after passing through their respiratory and digestive tracts. Although several studies have reported the toxic effects of Cd exposure on human health, its role in embryonic development during preimplantation stage remains unclear. We investigated the effects of Cd on porcine embryonic development and elucidated the mechanism. Methods: We cultured parthenogenetic embryos in media treated with 0, 20, 40, or 60 μM Cd for 6 days and evaluated the rates of cleavage and blastocyst formation. To investigate the mechanism of Cd toxicity, we examined intracellular reactive oxygen species (ROS) and glutathione (GSH) levels. Moreover, we examined mitochondrial content, membrane potential, and ROS. Results: Cleavage and blastocyst formation rates began to decrease significantly in the 40 μM Cd group compared with the control. During post-blastulation, development was significantly delayed in the Cd group. Cd exposure significantly decreased cell number and increased apoptosis rate compared with the control. Embryos exposed to Cd had significantly higher ROS and lower GSH levels, as well as lower expression of antioxidant enzymes, compared with the control. Moreover, embryos exposed to Cd exhibited a significant decrease in mitochondrial content, mitochondrial membrane potential, and expression of mitochondrial genes and an increase in mitochondrial ROS compared to the control. Conclusions: We demonstrated that Cd exposure impairs porcine embryonic development by inducing oxidative stress and mitochondrial dysfunction. Our findings provide insights into the toxicity of Cd exposure on mammalian embryonic development and highlight the importance of preventing Cd pollution.

During the dismantling of nuclear facilities, a large quantity of radioactive concrete is generated and chelating agents are required for the decontamination process. However, disposing of environmentally persistent chelated wastes without eliminating the chelating agents might increase the rate of radionuclide migration. This paper reports a rapid and straightforward ion chromatography method for the quantification of citric acid (CA), a commonly used chelating agent. The findings demonstrate acceptable recovery yields, linearities, and reproducibilities of the simulated samples, confirming the validity of the proposed method. The selectivity of the proposed method was confirmed by effectively separating CA from gluconic acid, a common constituent in concretes. The limits of detection and quantification of the method were 0.679 and 2.059 mg·L−1, respectively, while the recovery yield, indicative of the consistency between theoretical and experimental concentrations, was 85%. The method was also employed for the quantification of CA in a real concrete sample. These results highlight the potential of this approach for CA detection in radioactive concrete waste, as well as in other types of nuclear wastes.

As the decommissioning of domestic nuclear power plants (Gori Unit 1 and Wolseong Unit 1) becomes more visible, many research projects are being conducted to safely and economically decommissioning of domestic nuclear power plants (NPPs). After permanent shutdown, decommissioning of NNPs proceeds through decontamination, cutting of main equipment, waste disposal and site restoration stages. And various technologies are applied at each stage. In particular, remote cutting of neutron induced structures (RV, RVI, etc.) is a technology used in developed countries in the cutting stage, and remote cutting has been evaluated as a core technology for minimizing workers’ radiation exposure. Generally, remote cutting technologies are divided into mechanical/thermal/electrical cutting. Among various thermal cutting technologies, plasma arc cutting (PAC) is more economical and easily to remote control than other cutting technologies, and is also effective in cutting STS304 plates. PAC is a thermal cutting technology that melts the base material at the cutting area with a plasma arc heat source and removes melted material by blowing it out with cutting gas. The cutting quality depends on the stand-off distance and power (current), material thickness, cutting speed, etc., while double arcing will occur if the cutting conditions are not suitable. A monitoring system that can confirm double arcing during remote cutting is necessary because double arcing can reduce cutting quality, increase secondary waste (increase kerf and aerosol), and cause non-cutting. In this study, we used an ultrahigh-speed camera equipped with a band-pass filter to capture clear arc shapes, and measured voltage waveforms with a data acquisition system. We studied a monitoring method that can confirm the occurrence of double arcing by synchronizing the obtained arc shape and voltage waveform, and the effects of double arcing on the STS304 plates. The results of this study are expected to be helpful in the development of the remote cutting process using plasma arc cutting when decommissioning of domestic NPPs.

Nuclear facilities present the important task related to the migration and retention of radioactive contaminants such as cesium (Cs), strontium (Sr), and cobalt (Co) for unexpected events in various environmental conditions. The distribution coefficient (Kd) is important factor for understanding these contaminants mobility, influenced by environmental variables. This study focusses the prediction of Kd values for radionuclides within solid phase groups through the application of machine-learning models trained on experimental data and open source data from Japan atomic energy agency. Three machine-learning models, such as the convolutional neural network, artificial neural network, and random forest, were trained for prediction model of the distribution coefficient (Kd). Fourteen input variables drawn from the database and experimental data, including parameters such as initial concentration, solid-phase characteristics, and solution conditions, served as the basis for model training. To enhance model performance, these variables underwent preprocessing steps involving normalization and log transformation. The performances of the models were evaluated using the coefficient of determination. These results showed that the environmental media, initial radionuclide concentration, solid phase properties, and solution conditions were significant variables for Kd prediction. These models accurately predict Kd values for different environmental conditions and can assess the environmental risk by analyzing the behavior of radionuclides in solid phase groups. The results of this study can improve safety analyses and longterm risk assessments related to waste disposal and prevent potential hazards and sources of contamination in the surrounding environment.

Over the years, in the field of safety assessment of geological disposal system, system-level models have been widely employed, primarily due to considerations of computational efficiency and convenience. However, system-level models have their limitations when it comes to phenomenologically simulating the complex processes occurring within disposal systems, particularly when attempting to account for the coupled processes in the near-field. Therefore, this study investigates a machine learning-based methodology for incorporating phenomenological insights into system-level safety assessment models without compromising computational efficiency. The machine learning application targeted the calculation of waste degradation rates and the estimation of radionuclide flux around the deposition holes. To develop machine learning models for both degradation rates and radionuclide flux, key influencing factors or input parameters need to be identified. Subsequently, process models capable of computing degradation rates and radionuclide flux will be established. To facilitate the generation of machine learning data encompassing a wide range of input parameter combinations, Latin-hypercube sampling will be applied. Based on the predefined scenarios and input parameters, the machine learning models will generate time-series data for the degradation rates and radionuclide flux. The time-series data can subsequently be applied to the system-level safety assessment model as a time table format. The methodology presented in this study is expected to contribute to the enhancement of system-level safety assessment models when applied.

In the case of dry storage facilities, slipping of the cask or tip-over are dangerous phenomena. For this reason, in dry storage facilities, measures against slipping and tip-over or related safety evaluations are important. Accidental conditions that can cause cask slippage and tip-over in dry storage facilities include natural phenomena such as floods, tornadoes, tsunamis, typhoons, earthquakes, and artificial phenomena such as airplane crashes. However, among natural phenomena, earthquakes are the most important natural phenomenon that causes tip-over. Also, many people had the stereotype that Korea is an earthquake-safe zone before 2016. However, earthquakes become a major disaster in Korea due to the 2016 Gyeongju earthquake and the 2017 Pohang earthquake, followed by the Goesan earthquake in October 2022. In this paper, seismic analysis was performed based on dry storage facilities including multiple casks. Design variables for the construction of an analysis model for dry storage facilities were investigated, and seismic analysis was performed. To evaluate tip-over accident during earthquake, seismic load was used from 0.2 g PGA to 0.8 g PGA and these earthquakes were followed Design Response Spectrum (DRS) in RG 1.60. The friction coefficient of concrete pad was used from 0.2 to 1.0. As a result of the analysis, tip-over accident could not find in the analysis from 0.2 g to 0.6 g. However, tip-over was appeared at friction coefficients of 0.8 and 1.0 at 0.8 g PGA. Tip-over angular velocity of cask was derived by seismic analysis and was compared with formula and tip-over analysis results. As a result, a generalized dry storage facility analysis model was proposed, and dry storage facility safety evaluation was conducted through seismic analysis. Also, tip-over angular velocity was derived using seismic analysis for tip-over analysis.

매년 국내로 비래해 오는 해충인 벼멸구는 그 기원이 중국 또는 중국 남부일 것으로 예상해왔으나, 이에 대한 유전학적 근거는 Mun et al. (1999)에 의해 제시된 세 가지 COI haplotype 비교가 유일하다. Mun et al. (1999)은 국내에 서 확인된 두 가지 haplotype 유형이 인도차이나반도 이남의 균일한 한 가지 haplotype 집단 유형과 중국에서 확인 된 또 다른 haplotype 집단 유형임을 근거로 국내 벼멸구의 기원을 중국으로 특정한 바 있다. 본 연구는 국내 및 동남아시아 5개국(부탄, 미얀마, 캄보디아, 라오스 및 태국)으로부터 직간접적으로 확보한 개체들을 대상으로 GBS (genotyping by sequencing) 및 NGS 기법을 통해 PCA를 포함한 다양한 집단유전학적 분석을 수행하였다. 그 결과 인도차이나반도의 벼멸구 집단은 크게 북부와 남부로 나뉘며, 국내 개체들은 북부에 비해 남부(캄보디 아, 태국)에 더 가깝다는 사실을 확인하였다. 따라서 벼멸구의 국내 비래는 중국으로부터의 기원 이전에 장마전 선이 형성될 무렵부터 인도차이나반도 남쪽의 고온다습한 서풍이 남남서풍으로 바뀌면서 중국 내륙을 거쳐 국내로 비래하는 경로를 따르는 것으로 보인다. 하지만 태안의 개체 중에는 인도차이나반도 집단들의 외군으로 확인되는 개체가 있었고, 이는 인도차이나반도 외의 샘플링되지 않은 다른 지역에서도 벼멸구가 국내로 비래할 수 있다는 가능성을 제시하였다. 따라서 국내로 유입되는 벼멸구의 유전적 기원을 확인하기 위해서는 인도차이 나반도 남쪽 지역에서 시작한 동아시아 여름 몬순의 바람이 한국으로 도착하는 경로에 위치한 다른 지역에서의 추가적인 샘플링 및 지속적인 관심과 추적이 필요할 것이다.

Recently, there has been an increase in human casualties and property damage caused by large-scale disasters such as massive wildfires, concentrated heavy rainfall, chemical accidents, and infectious diseases. To enhance the citizen protection capabilities of civil defense personnel, it is necessary to ensure their capacity to perform on-site citizen protection duties and guarantee their safety. Particularly, the current civil defense uniforms are vulnerable to fire and waterproofing, and the safety is threatened when civil defense personnel perform field activities such as fire and flood damage. Therefore, there is an urgent need for research and development of civil defense uniforms to ensure functionality and safety. In this study, we aim to understand the duties of domestic civil defense units and the characteristics of on-site situations, analyze overseas civil defense uniform replication cases, and derive safety criteria for civil defense uniforms. Developing civil defense uniforms based on safety and functional criteria is expected to enhance civil defense personnel's capabilities and increase public welfare by deriving optimal performance for each mission.

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.

As Korea has relatively small land area and large population density compared to other countries considering the DGD concept such as Finland and Sweden, improvements of disposal efficiency in the viewpoint of the disposal area might be needed for the current disposal system to alleviate the difficulties of site selection for the HLW repository. In this research, we conduct a numerical investigation of the disposal efficiency enhancement for a high-level radioactive waste (HLW) repository through three design factors: decay heat optimization, increased thermal limit of buffer, and double-layer concept. In the optimized decay heat model, seven SNFs with the maximum and minimum decay heat depending on actual burn-up and cooling time are iteratively combined in a canister. Thermal limit of buffer is assumed as 100°C and 130°C for reference and high-efficiency repository concepts, respectively. By implementing an optimized decay heat model and a single-layer concept with a thermal limit of buffer set at 100°C, the disposal efficiency increases to 2.3 times of the improved Korean Reference disposal System (KRS+). Additionally, incorporating either an increased thermal limit of buffer to 130°C or a double-layer concept leads to a further 50% improvement in disposal efficiency. By integrating all three design factors, the disposal efficiency can be enhanced up to five times that of the KRS+ repository. Our analysis of rock mass stability reveals that increasing the thermal limit of buffer can generate rock spalling failure in a wider area. However, when accounting for the effect of confining stress by swelling of buffer and backfill using the Mohr-Coulomb failure criteria, the rock mass failure only occurred at the corner between the disposal tunnel and deposition hole when the thermal limit of buffer was increased and a single-layer concept was applied. The results given in this study can provide various options for designing the high-efficiency repository in accordance with the target disposal area and quality of the rock mass in the potential repository site.

Concrete structures of spent nuclear fuel interim storage facility should maintain their ability to shield and structural integrity during normal, off-normal and accident conditions. The concrete structures may deteriorate if the interim storage facility operates for more than several decades. Even if deterioration occurs, the concrete structures must maintain their own functions such as radiation shielding protection and structural integrity. Therefore, it is necessary to establish an analysis methodology that can evaluate whether the deteriorated concrete structure maintains its integrity under not only normal or off-normal condition but also accident condition. In accident conditions such as tip over and aircraft collision, both static material properties and dynamic properties are needed to evaluate the structural integrity of the concrete structures. Especially, it has been known to be difficult to estimate the resulted damage precisely where an aircraft collides with the degraded concrete structures at a high strain rate. In this study, damage evaluation of concrete overpack due to aircraft collisions was conducted. First, in order to verify the impact analysis methodology, the aircraft impact analysis of plane concrete overpack was performed and compared with the test results previously conducted by our research team. Then, the impact analysis for the overpack of KORAD21C was performed. In the future, the radiation shielding analysis will be performed under the conditions to evaluate whether or not the radiation shielding ability is maintained.

The volatilization of alkali ions in (K,Na)NbO3 (KNN) ceramics was inhibited by doping them with alkaline earth metal ions. In addition, the grain growth behavior changed significantly as the sintering duration (ts) increased. At 1,100 °C, the volatilization of alkali ions in KNN ceramics was more suppressed when doped with alkaline earth metal ions with smaller ionic size. A Ca2+-doped KNN specimen with the least alkali ion volatilization exhibited a microstructure in which grain growth was completely suppressed, even under long-term sintering for ts = 30 h. The grain growth in Sr2+-doped and Ba2+-doped KNN specimens was suppressed until ts = 10 h. However, at ts = 30 h, a heterogeneous microstructure with abnormal grains and small-sized matrix grains was observed. The size and number of abnormal grains and size distribution of matrix grains were considerably different between the Sr2+-doped and Ba2+-doped specimens. This microstructural diversity in KNN ceramics could be explained in terms of the crystal growth driving force required for two-dimensional nucleation, which was directly related to the number of vacancies in the material.

Dose-rate monitoring instruments are indispensable to protect workers from the potential risk of radiation exposure, and are commonly calibrated in terms of the ambient dose equivalent (H*(10)), an operational quantity that is widely used for area monitoring. Plastic scintillation detectors are ideal equipment for dosimetry because of their advantages of low cost and tissue equivalence. However, these detectors are rarely used owing to the characteristics caused by low-atomic-number elements, such as low interaction coefficients and poor gamma-ray spectroscopy. In this study, we calculated the G(E) function to utilize a plastic scintillation detector in spectroscopic dosimetry applications. Numerous spectra with arbitrary energies of gamma rays and their H*(10) were calculated using Monte Carlo simulations and were used to obtain the G(E) function. We acquired three different types of G(E) functions using the least-square and first-order methods. The performances of the G(E) functions were compared with one another, including the conventional total counting method. The performance was evaluated using 133Ba, 137Cs, 152Eu, and 60Co radioisotopes in terms of the mean absolute percentage error between the predicted and true H*(10) values. In addition, we confirmed that the dose-rate prediction errors were within acceptable uncertainty ranges and that the energy responses to 137Cs of the G(E) function satisfied the criteria recommended by the International Commission.

As an alternative technology for the efficient disposal of spent nuclear fuel, various process flows can be selected based on the recovered and separated radioactive nuclide group. This is to examine the efficiency of the disposal area of spent nuclear fuel when various disposal technologies and several treatment processes are applied to spent nuclear fuel, compared to the deep geological disposal of burying the entire spent fuel in the ground. Above all, the biggest advantage of the optional treatment processes is that it can be applied to various disposal methods (deep borehole disposal, deep geological disposal) because it can process spent fuel in various sizes and separate into some groups according to the properties of radionuclides. These optional processes are not new technology and currently available as of today, and the level is classified based on the stepwise separation of high heat emission nuclides and long half-life nuclides. This is to increase the efficiency of the disposal of spent nuclear fuel by separating and managing high-risk radionuclides separately. Relatively various optional processes are possible depending on the level, and characteristic analysis is performed on wastes treated with alternative technologies. The mass balance for each option process is completed, and the amount of waste is also calculated accordingly. These are used as basic data for waste disposal area and economic evaluation. Besides it is easy to process spent fuel of various sizes suitable for deep geological disposal or deep borehole disposal technology when an optional treatment technology is applied to spent fuel. However, since this selective process is based on the process structure constructed in a broad framework, it is considered that additional follow-up studies are needed not only on detailed technology but also on the flow and amount of waste.

Considering the domestic condition with small land area and high population density, it is necessary to develop technology that can reduce the disposal area than the deep geological disposal method. For this, KAERI is developing a nuclide management process that can reduce the environmental burden of spent fuel, and establishing an evaluation model that can evaluate the performance of various process options. It is expected that an optimal option of the nuclide management process can be derived from disposal perspective by applying the evaluation model. The mass flow between processing steps of the radionuclide management process is the basic quantity required to quantify the evaluation criteria. Therefore, we built a generalized block model on GoldSim, which can simulate mass flow of various radionuclide management process options. In addition to the mass flow, this model was established to derive the amount of wastes generated by each processing step, the composition of nuclides, and radiological properties (decay heat, radioactivity, etc.). The mass flow and waste property derived from the models are closely related to the factors that determine the area of disposal concepts. Based on this, a disposal area calculation model was established as a model to evaluate the effectiveness of the radionuclide management process on environmental burden reduction. For verification, three process options, which can manage radionuclides having high decay heat (Cs, Sr) or large volume (U), were selected and evaluated as reference processes. And two disposal options, deep geological disposal and deep borehole disposal concepts were considered to be linked with the processes. As a result, it was confirmed that the disposal area could be reduced in the process separating radionuclides having high decay heat. In the future, other evaluation models for economic viability and safety will be added in the GoldSim model.

Some Spent Fuel Pools (SFPs) will be full of Spent Nuclear Fuels (SNFs) within several years. Because of this reason, building interim storage facilities or permanent disposal facilities should be considered. These storage facilities are divided into wet storage facilities and dry storage facilities. Wet storage facility is a method of storing SNF in SFP to cool decay heat and shielding radiation, and dry storage facility is a method of storing SNF in a cask and placing on the ground or storage building. However, wet storage facilities have disadvantages in that operating costs are higher than that of dry storage facilities, and additional capacity expansion is difficult. Dry storage facilities have relatively low operating costs and are relatively easy to increase capacity when additional SNFs need to be stored. For this reason, since the 1990s, the number of cases of applying dry storage facilities has been increasing even abroad. Dry storage facilities are divided into indoor storage facilities and outdoor storage facilities, and outdoor storage facilities are mostly used to take advantage of dry storage facilities. In the case of outdoor storage facilities, the cask in which SNFs are stored is placed on a designed concrete pad. During this storage, the boring heat generated by SNFs cools into natural convection and the cask shields the radiation that SNFs generates. However, if an accident such as an earthquake occurs and the cask overturns during storage, there may be a risk of radiation leakage. Such a tip-over accident may be caused by the cask slipping due to the vibration of an earthquake, or by not supporting the cask properly due to a problem in the concrete pad. Therefore, in the case of outdoor dry storage facilities, it is necessary to evaluate the seismic safety of concrete pads. In this paper, various soil conditions were applied in the seismic analysis. Soil conditions were classified according to the shear wave velocity, and the shear wave velocity was classified according to the ground classification criteria according to the general seismic design (KDS 17 10 00). The concrete pad was designed with a size that 8 casks can be arranged at regular intervals, and 11# reinforcing bars were used for the design of the internal reinforcement of the concrete pad according to literature research. The cask was designed as a rigid body to shorten the analysis time. The soil to which the elastic model was applied was designed under the concrete pad, and infinite elements were applied to the sides and bottom of the soil. The effect on the concrete pad and cask by applying a seismic wave conforming to RG 1.60 to the bottom of the soil was analyzed with a finite element model.

There are highly toxic radio-isotopes and high heat emitting isotopes in spent nuclear fuels which could be a burden in a deep geological repository. Some preliminary study in order to see if there are some advantages in terms of waste burden, in case that the spent fuel is appropriately processed and then disposed of in a final repository, has been carried out at KAERI. This study is focused on the proliferation resistance for various processing alternatives for them. The evaluation criteria and their indicators for proliferation resistance analysis are selected and then evaluated quantitatively or quantitatively for the alternatives. The processing alternatives are grouped into three categories according to the level of decrease of burden for final disposal and named them as Level I, Level II and Level III technolgy alternatives. Level I alternative is to maximize the long-term safety in the final repository from the removal of I- 129, semi-volatile radioisotope, which is the greatest impact on the long-term safety of the repository. Level II alternative is to remove the strontium-90, high heat emitter, in addition to the removal in Level I. The Level III is to additionally remove uranium from main stream of the level II to reduce the volume of the high level wastes to be disposed. The intrinsic radiation and chemical barriers against the nuclear proliferation are selected and analyised for the alternatives. It is resulted from the proliferation resistance analysis that all three options showed excellent resistance to nuclear proliferation for the two barriers. However, Level III technology including electrochemical refining process is relatively a little weaker than others. Overall, it could be an effective means to reduce the burden of disposal if the spent fuels are appropriately conditioned for final disposal. Further detailed studies are, however, needed to finalize its feasibility.