With South Korea increasingly focusing on nuclear energy, the management of spent nuclear fuel has attracted considerable attention in South Korea. This study established a novel procedure for selecting safety-relevant radionuclides for long-term safety assessments of a deep geological repository in South Korea. Statistical evaluations were performed to identify the design basis reference spent nuclear fuels and evaluate the source term for up to one million years. Safety-relevant radionuclides were determined based on the half-life criteria, the projected activities for the design basis reference spent nuclear fuel, and the annual limit of ingestion set by the Nuclear Safety and Security Commission Notification No. 2019-10 without considering their chemical and hydrogeological properties. The proposed process was used to select 56 radionuclides, comprising 27 fission and activation products and 29 actinide nuclides. This study explains first the determination of the design basis reference spent nuclear fuels, followed by a comprehensive discussion on the selection criteria and methodology for safety-relevant radionuclides.

Within the air purification system of a nuclear power plant, specific radioactive isotopes are extracted from gases through adsorption onto activated carbon. To properly dispose of used activated carbon, it is essential to determine the concentration of radioactive nuclides within it. This study discusses the application of the pyrolysis method for analyzing the concentrations of 3H and 14C in spent activated carbon. The pyrolysis was conducted using Raddec’s Pyrolyser, with adjustments made to parameters such as temperature profiles, airflow rates, sample quantities, and trapping solution volumes. The evaluation method for the pyrolysis of activated carbon to analyze 3H and 14C involved adding 3H and 14C sources to the activated carbon before use and subsequently assessing the recovery rates of the added sources in comparison to the analysis results.

In the decommissioning process of nuclear power plants, Ni-59, Ni-63 and Fe-55 present in radioactive waste are crucial radionuclides used as fundamental indicators in determining waste treatment methods. However, due to their low-energy emissions, the chemical separation of these two radionuclides is essential compared to others. Therefore, this study aims to evaluate the suitability of various pre-treatment methods for decommissioning waste materials by conducting characteristic assessments at each chemical separation stage. The goal is to find the most optimized pre-treatment method for the analysis of Ni-59, Ni-63 and Fe-55 in decommissioning waste. The comparative evaluation results confirm that the chemical separation procedures for Fe and Ni are very stable in terms of stepwise recovery rates and the removal of interfering radionuclides. However, decommissioning waste materials, which mainly consist of concrete, metals, etc., possess unique properties, and a significant portion may be low-radioactivity waste suitable for on-site disposal. Considering that the chemical behavior and reaction characteristics may vary at each chemical separation stage depending on the matrix properties of the materials, it is considered necessary to apply cascading chemical separation or develop and apply individual chemical separation methods. This should be done by verifying and validating their effectiveness on actual decommissioning waste materials.

This study presents distribution of naturally occurring radioactive materials in groundwater in Jeju island. Radon (222Rn) and potassium (40K) concentrations were performed by using Liquid Scintillation Counter and Ion Chromatograph respectively. In addition, the activities of uranium and thorium nuclides were analyzed by Inductively Coupled Plasma Mass Spectroscopy. Groundwater samples were collected from 9 sites of water intake facilities for wide area supply in Jeju island from September 2022 to September 2023. The 40K concentrations of groundwater ranged between 0.050 and 0.400 Bq·L-1. The radon concentrations in groundwater were in the range of 0 to 60 Bq L-1, and there was no groundwater exceeding the range of 148 Bq L-1 proposed by the US EPA. The distribution of uranium and thorium in groundwater varied from 0 to 500 ng L-1 and 0 to 2.4 ng L-1, respectively. The concentrations of uranium did not exceed 30 μg L-1, thresholds indicated by the US EPA. By analyzing the concentrations of 40K, 222Rn, 238U and 232Th, the annual effective dose of residents can be assessed. The evaluated residents’ effective dose from natural radionuclides due to intake of drinking water is less than the recommended value of 100 μSv y-1. Consequently, this study indicates that the cancer risks of the residents in Jeju island from naturally occurring radioactive materials ingested with water is insignificant.

The inorganic scintillator used in gamma spectroscopy must have good efficiency in converting the kinetic energy of charged particles into light as well as high light output and high light detection efficiency. Accordingly, various studies have been conducted to enhance the net-efficiency. One way to improve the light yield has been studied by coating scintillators with various nanoparticles, so that the scintillation light can undergo resonance on surface between scintillators and nanoparticles resulting in higher light yield. In this study, an inorganic scintillator coated with CsPbBr3 perovskite nanocrystals using dip coating technique was proposed to improve scintillation light yield. The experiment was carried out by measuring scintillation light output, as the result of interaction between inorganic scintillator coated with CsPbBr3 perovskite nanocrystals and gamma-ray emitted from Cs-137 gamma source. The experimental results show that the channel corresponding to 662 keV full energy peak in the Cs-137 spectrum shifted to the right by 14.37%. Further study will be conducted to investigate the detailed relationships between the scintillation light yield and the characteristics of coated perovskite nanoparticles, such as diameter of nanoparticles, coated area ratio and width of coated region.

The mobility of radionuclides in the subsurface environment is governed by a interaction of radioactivity characteristics and geochemical conditions with adsorption reactions playing a critical role. This study investigates the characteristics and mechanisms of radionuclides adsorption on site media in viewpoint of nuclear safety, particularly focusing on the potential effect of seawater infiltration in coastal site near nuclear power plant. Seawater intrusion alters the chemistry in groundwater, including parameters such as pH, redox potential, and ionic strength, thereby affecting the behavior of radionuclides. To assess the safety of site near nuclear power plant and the environmental implications of nuclide leakage, this research conducted various experiments to evaluate the behavior of radionuclides in the subsurface environment. High distribution coefficients (50-2,500 ml/g) were observed at 10 mg/L Co, with montmorillonite > hydrobiotite > illite > kaolinite. It decreased with competing cations (Ca2+) and was found to decrease significantly by 90% with a decrease in pH to 4. It is believed that the adsorption capacity of cationic radionuclides decreases significantly as the clay mineral surface becomes less negatively charged. For Cs, the distribution coefficient (180-560 ml/g) was higher for montmorillonite > hydrobiotite > illite > kaolinite. Compared to Co, it was found to be less influenced by pH and more influenced by competing cations. For Sr, the distribution coefficient (100-380 ml/g) was higher in the order of hydrobiotite > montmorillonite > illite > kaolinite. Compared to Cs, it was found to be less affected by pH and also less affected by the effect of competing cations compared to Cs. Seawater samples from Gampo and Uljin site near Nuclear Power Plant in Korea were analyzed to determine their chemical composition, which was subsequently used in adsorption experiments. Additionally, the seawater-infiltrated groundwater was synthesized in laboratory according to previous literature. The study focused on the adsorption and behavior of three key radionuclides such as cesium, strontium, and cobalt onto four low permeability media (clay minerals) such as kaolinite, illite, hydrobiotite, and montmorillonite known for their high adsorption capacity at a site of nuclear power plant. At concentrations of 5 and 10 mg/L, the adsorption coefficients followed the order of cobalt > cesium > strontium for each radionuclide. Notably, the distribution coefficient (Kd) values exhibited higher values in seawater-infiltrated groundwater environments compared to seawater with relatively high ionic strength. Cobalt exhibited a substantial adsorption coefficient, with a marked decrease in Kd values in seawater conditions due to elevated ionic strength. In contrast, cesium displayed less dependency on seawater compared to other radionuclides, suggesting distinct adsorption mechanisms, possibly involving fractured edge sites (FES) in clay. Strontium exhibited a significant reduction in adsorption in seawater compared to groundwater in all Kd sorption experiments. The adsorption data of cobalt, cesium, and strontium on clay minerals in contact with seawater and seawater-infiltrated solutions offer valuable insights for assessing radioactive contamination of groundwater beneath coastal site near nuclear power plant sites. This research provides a foundation for enhancing the safety assessment protocols of nuclear power plant sites, considering the potential effects of seawater infiltration on radionuclide behavior in the subsurface environment.

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.

To effectively assess the inventory of radionuclides generated from nuclear power plants using a consistent evaluation method across diverse groups, it is imperative to analyze the similarity in radioactive distribution between these groups. Various methodologies exist for evaluating this similarity, and the application of statistical approaches allows us to establish similarity at a specific confidence level while accounting for the dataset size (degrees of freedom). Initially, if the variance characteristics of the two groups are similar, a t-test for equal variances can be employed. However, if the variance characteristics differ, methods for unequal variances should be applied. This study delineates the approach for assessing the similarity in radioactive distribution based on the analytical characteristics of the two groups. Furthermore, it delves into the results obtained through two case studies to offer insights into the assessment process.

A disposal system for spent nuclear fuel mainly divides into two parts; Engineered barriers include spent nuclear fuel, canister, buffer and backfill and natural barriers mean a host rock surrounding engineered barriers. If radionuclides released from a repository, they can migrate to the ecosystem. Sorption plays an important role in retarding the migration of released radionuclides. Hence, the safety assessment for the disposal of a spent nuclear fuel should consider the migration and retardation of radionuclides in geosphere. Distribution coefficient is one of input parameters for the safety assessment. In this work, distribution coefficients for crystalline rock as a natural barrier were collected and evaluated for the purpose of safety assessment for the deep geological disposal of a spent nuclear fuel. The radionuclides considered in this work are as follows; alkali and alkaline earth metals (Cs, Sr, Ba), lanthanides (Sm), actinides (Ac, Am, Cm, Np, Pa, Pu, Th U), transition elements (Nb, Ni, Pd, Tc, Zr), and others (C, Cl, I, Rn, Se, Sn). The sorption of radionuclides is influenced by various geochemical conditions such as pH/carbonates, redox potential, ionic strength, radionuclide concentration, kinds and amounts of minerals, and microbes. For the evaluation of distribution coefficients, the data from Sweden (SKB), Finland (Posiva), Switzerland (Nagra), and Japan (JAEA) were collected, analyzed, and the recommended distribution coefficients have been suggested.

Spent nuclear fuel management is a high-priority issue in South Korea, and addressing it is crucial for the country’s long-term energy sustainability. The KORAD (Korea Radioactive Waste Agency) is leading a comprehensive, long-term project to develop a safe and effective deep geological repository for spent nuclear fuel disposal. Within this framework, we have three primary objectives in this work. First, we conducted statistical analysis to assess the inventory of spent nuclear fuel in South Korea as of 2021. We also projected future generation rates of spent nuclear fuels to identify what we refer to as reference spent nuclear fuels. These reference spent nuclear fuels will be used as the design basis spent fuels for evaluating the safety of the repository. Specifically, we identified four types of design basis reference spent nuclear fuels: high and low burnup from PLUS7 (with a 16×16 array) and ACE7 (with a 17×17 array) assemblies. Second, we analyzed radioactive nuclides’ inventory, activities, and decay heats, extending up to a million years after reactor discharge for these reference spent nuclear fuels. This analysis was performed using SCALE/TRITON to generate the burnup libraries and SCALE/ORIGEN for source term evaluation. Third, to assess the safety resulted from potential radioactive nuclides’ release from the disposal canister in future work, we selected safety-related radionuclides based on the ALI (Annual Limit of Intake) specified in Annex 3 of the 2019-10 notification by the NSSC (Nuclear Safety and Security Commission). Conservative assumptions were made regarding annual water intake by humans, canister design lifetime, and aquifer flow rates. A safety margin of 10-3 of the ALI was applied. We selected 56 radionuclides that exceed the intake limits and have half-lives longer than one year as the safety-related radionuclides. However, it is crucial to note that our selection criteria focused on ALI and half-lives. It did not include other essential factors such as solubility limits, distribution coefficients, and leakage processes. So, some of these nuclides can be removed in a specific analysis area depending on their properties.

Given the situation in the Republic of Korea that all nuclear power plants are located at the seaside, the interim storage facility is also likely to be located at seaside and the maritime transportation of Spent Nuclear Fuel is considered inevitable. The Republic of Korea does not have an independently developed maritime transportation risk assessment code, and no research has been conducted to evaluate the release rate of radionuclides from a submerged transportation cask in the sea. Therefore, there is a need to develop a technology that can assess the impact of immersion accidents and establish a regulatory framework for maritime transportation accidents. The release rate of radionuclides should be calculated from the flow rate through a flow path in the breached containment boundary. According to the cask design criteria, it is anticipated that even under severe accident conditions, the flow path size will be very small. Previous studies have evaluated fluid flow passing through micro-scale channel by integrating internal and external flows within and around a transport cask. As part of the evaluation, a comprehensive “Full-Field Model” incorporating external flow fields and a localized “Local-Field Model” with micro-scale flow paths were constructed. Sub-modeling techniques were employed to couple the flow field calculated by the two models. The aforementioned approach is utilized to conduct the evaluation of fluid flow passing through micro-scale flow paths. This study aims to evaluate fluid flow passing through micro-scale flow paths using the aforementioned CFD (Computational Fluid Dynamics) method and aims to code the findings. The Gaussian Process Regression technique, a machine learning model, is utilized for developing a mathematical metamodel. The selected input parameters for coding are organized and their respective impacts are analyzed. The range of these selected parameters is tailored to suit domestic environments, and computational experiments are planned through Design of Experiments. The flow path size is included as an input parameter in the coded model. In cases where the flow path size becomes extremely small, making it impractical to use CFD techniques for calculations, Poiseuille’s law is employed to calculate the release rate. In this study, a model is developed to evaluate the release rate of radionuclides using CFD and mathematical equations covering the whole possible range of flow path size in a lost cask in the deep sea. The model will be used in the development of a maritime transportation risk assessment code suitable for the situation and environment in Korea.

The organic complexing agents such as ethylenediaminetetraacetic acid (EDTA), nitrilotriacetic acid (NTA), and isosaccharinic acid (ISA) can enhance the radionuclides’ solubility and have the potential to induce the acceleration of radionuclides’ mobility to a far-field from the radioactive waste repository. Hence, it is essential to evaluate the effect of organic complexing agents on radionuclide solubility through experimental analysis under similar conditions to those at the radioactive waste disposal site. In this study, five radionuclides (cesium, cobalt, strontium, iodine, and uranium) and three organic complexing agents (EDTA, NTA, and ISA) were selected as model substances. To simulate environmental conditions, the groundwater was collected near the repository and applied for solubility experiments. The solubility experiments were carried out under various ranges of pHs (7, 9, 11, and 13), temperatures (10°C, 20°C, and 40°C), and concentrations of organic complexing agents (0, 10-5, 10-4, 10-3, and 10-2 M). Experimental results showed that the presence of organic complexing agents significantly increased the solubility of the radionuclides. Cobalt and strontium had high solubility enhancement factors, even at low concentrations of organic complexing agents. We also developed a support vector machine (SVM) model using some of the experimental data and validated it using the rest of the solubility data. The root mean square error (RMSE) in the training and validation sets was 0.012 and 0.016, respectively. The SVM model allowed us to estimate the solubility value under untested conditions (e.g., pH 12, temperature 30°C, ISA 5×10-4 M). Therefore, our experimental solubility data and the SVM model can be used to predict radionuclide solubility and solubility enhancement by organic complexing agents under various conditions.

As a result of various generation, transmutation, and decay schemes, a wide variety of radionuclides exist in the reactor prior to accident occurrence. Considering all of the radionuclides as the accident source term in an offsite consequence analysis will inevitably take up excessive computer resources and time. Calculation time can be reduced with minimal impact on the accuracy of the results by considering only the nuclides that have a significant effect on the calculation among the potential radioactive sources that may be released into the environment. In earlier studies related to offsite consequence analysis, it is shown that the principal criteria for the radionuclide screening applied are as follows; radionuclide inventory in the reactor, radioactive half-life, radionuclide release fraction to the environment, relative dose contribution of nuclides within a specific group, and radiobiological importance. As a result, it is confirmed that 54, 60, and 69 nuclides are applied to the risk assessment performed in WASH-1400, NUREG-1150, and SOARCA (State-of-the-Art Reactor Consequence Analyses) project in the United States, respectively. In addition, in this study, the technical consultations with domestic and foreign experts were carried out to confirm details on criteria and process for screening out radionuclides in offsite consequence analysis. In this paper, based on the literature survey and technical consulting, we derived the screening process of selecting a list of radionuclides to be considered in the offsite consequence analysis. The first step is to eliminate radionuclides with little core inventory (less than specific threshold) or very short half-lives. However, important decay products of radionuclides that have short half-lives should not be excluded by this process. The next step is to further eliminate radionuclides by considering contribution to offsite impact, which is defined as a product of radioactivity released to the environment (i.e. ‘inventory in the reactor’ times ‘release fraction to offsite’) and comprehensive dose (or risk) coefficient taking into account all exposure pathways to be included. The final step is to delete isotopes that contribute less than certain threshold to any important dose metric through additional computer runs for each important source term. Even though it is presumed that this process is applicable to existing light water reactors and the set of accidents that would be considered in PSA, some of the assumptions or specific recommendations may need to be reconsidered for other reactor types or set of accident categories.