To address the pressing societal concern in Korea, characterized by the imminent saturation of spent nuclear fuel storage, this study was undertaken to validate the fundamental reprocessing process capable of substantially mitigating the accumulation of spent nuclear fuel. Reprocessing is divided into dry processing (pyro-processing) and wet reprocessing (PUREX). Within this context, the primary focus of this research is to elucidate the foundational principles of PUREX (Plutonium Uranium Redox Extraction). Specifically, the central objective is to elucidate the interaction between uranium (U) and plutonium (Pu) utilizing an organic phase consisting of tributyl phosphate (TBP) and dodecane. The objective was to comprehensively understand the role of HNO3 in the PUREX (Plutonium Uranium Redox Extraction) process by subjecting organic phases mixed with TBPdodecane to various HNO3 concentrations (0.1 M, 1.0 M, 5.0 M). Subsequently, the introduction of Strontium (Sr-85) and Europium (Eu-152) stock solutions was carried out to simulate the presence of fission products typically contented in the spent nuclear fuel. When the operation proceeds, the complex structure takes the following form. 􀜷􀜱􀬶 􀬶􀬾(􀜽􀝍) + 2􀜰􀜱􀬷 􀬿(􀜽􀝍) + 2􀜶􀜤􀜲(􀝋􀝎􀝃) ↔ 􀜷􀜱􀬶(􀜰􀜱􀬷)􀬶 ∙ 2􀜶􀜤􀜲(􀝋􀝎􀝃) Subsequently, separate samples were gathered from both the organic and aqueous phases for the quantification of gamma-rays and alpha particles. Alpha particle measurements were conducted utilizing the Liquid Scintillation Counter (LSC) system, while gamma-ray measurements were carried out using the High-Purity Germanium Detector (HPGe). The distribution ratio for U, Eu (Eu-152), and Sr (Sr-84) was ascertained by quantifying their activity through LSC and HPGe. Through the experiments conducted within this program, we have gained a comprehensive understanding of the selective solvent extraction of actinides. Specifically, uranium has been effectively separated from the aqueous phase into the organic phase using a combination of tributyl phosphate (TBP) and dodecane. Subsequently, samples containing U(VI), Eu(III), and Sr(II) underwent thorough analysis utilizing LSC and HPGe detectors. Our radiation measurements have firmly established that the concentration of nitric acid enhances the selective separation of uranium within the process.

Once discharged, spent nuclear fuel undergoes an initial cooling process within deactivation pools situated at the reactor site. This cooling step is crucial for reducing the fuel’s temperature. Once the heat has sufficiently diminished, two viable options emerge: reprocessing or interim storage. A method known as PUREX, for aqueous nuclear reprocessing, involves a chemical procedure aimed at separating uranium and plutonium from the spent nuclear fuel. This separation not only minimizes waste volume but also facilitates the reuse of the extracted materials as fuel for nuclear reactors. The transformation of uranium oxides through dissolution in nitric acid followed by drying results in uranium taking the form of UO2(NO3)2 + 6H2O, which can then be converted into various solid-state configurations through different heat treatments. This study specifically focuses on investigating the phase transitions of artificially synthesized UO2(NO3)2 + 6H2O subjected to heat treatment at various temperatures (450, 500, 550, 600°C) using X-ray Diffraction (XRD) analysis. Heat treatments were also conducted on UO2 to analyze its phase transformations. Additionally, the study utilized XRD analysis on an unidentified oxidized uranium oxide, UO2+X, and employed lattice parameters and Bragg’s law to ascertain the oxidation state of the unknown sample. To synthesize UO2(NO3)2 + 6H2O, U3O8 powder is first dissolved in a 20% HNO3 solution. The solid UO2(NO3)2 + 6H2O is obtained after drying on a hotplate and is subsequently subjected to heat treatment at temperatures of 450, 500, 550, and 600°C. As the heat treatment temperature increases, the color of the samples transitions from orange to dark green, indicating the formation of different phases at different temperatures. XRD analysis confirms that uranyl nitrate, when heattreated at 500 and 550°C, oxidizes to UO3, while the sample subjected to 600°C heat treatment transforms into U3O8 due to the higher temperature. All samples exhibit sharp crystal peaks in their XRD spectra, except for the one heat-treated at 450°C. In the second experiment, the XRD spectra of the heat-treated UO2 consistently indicate the presence of U3O8 rather than UO3, regardless of the temperature. Under an oxidizing atmosphere within a temperature range of 300 to 700°C, UO2 can be oxidized to form U3O8. In the final experiment, the oxidation state of the unknown UO2+X was determined using Bragg’s law and lattice parameters, revealing that it was a material in which UO2 had been oxidized, resulting in an oxidation state of UO2.24.

Spent fuels (SFs) are stored in a storage pool after discharge from nuclear power plants. They can be transferred to for the further processes such as dry storage sites, processing plants, or disposal sites. One of important measures of SF is the burnup. Since the radioactivity of SF is strongly dependent on its burnup, the burnup of SF should be well estimated for the safe management, storage, and final disposal. Published papers about the methodology for the burnup estimation from the known activities of important radioactive sources are somewhat rare. In this study, we analyzed the dependency of the burnup on the important radiation source activities using ORIGEN-ARP, and suggested simple correlations that relate the burnup and the important source activities directly. A burnup estimation equation is suggested for PWR fuels relating burnup with total neutron source intensity (TNSI), initial enrichment, and cooling time. And three burnup estimation equations for major gamma sources, 137Cs, 134Cs, and 154Eu are also suggested.

Recently, the spent fuel pools withdrawn from nuclear power plants in Korea have been saturated. Therefore, specific regulations on the management of spent fuel pools, such as transportation and intermediate storage are needed. The burnup history is directly related to the management of spent nuclear fuel. This is because the decision to handle nuclear fuel may vary depending on the initial concentration of nuclear fuel, the degree to which nuclear fuel is irradiated and radioisotope nuclides are decayed, and the cooling state in the spent nuclear fuel storage tank. The purpose of this study is to determine the burnup of fuel based on the value obtained by scanning the surface of spent nuclear fuel through a neutron detector. Conversely, a database of neutron signals that scan bundles of spent nuclear fuel with an instrument with an already identified combustion history needs to be completed. First of all, the correlation between burnup history and nuclides was identified in previous studies. By setting the burnup history as the input value in the ORIGEN-ARP code, it was possible to identify the radioactive isotopes remaining in the bundle of nuclear fuel. Neutrons can finally be measured based on the amount of nuclide inventory that constitutes spent nuclear fuel. Through MCNP, the neutron detector was simulated and signals were measured to confirm how it correlates with the previously acquired burnup history database. In addition, the M (sub-critical multiplication) value, which is essential for neutron measurement, was checked to confirm the degree to which additional neutrons were generated in spent nuclear fuel in a subcritical state. The target nuclear fuel assembly was CE16×16, WH14×14, and WH17×17, which confirmed the correlation (1) between burnup, enrichment, and cooling time with the previous research topic, TNSI (Total neutron source intensity). 􀜤􀜷􁈺􀜩􀜹􀝀/􀜯􀜶􀜷􁈻 = 0.83􁈺􀜵􀯇􁈻􀬴.􀬶􀬷􀬼 ∙ 􁈺􀜫􀜧􁈻􀬴.􀬸􀬺􀬶􀬻 ∙ 􀝁􀬴.􀬴􀬴􀬼􀬷∙􀯧 􁈺1􁈻 A neutron signal will be obtained from the case according to each burnup history constituting this database. In particular, PAR=SF, a function that calculates the production amount of the fission product, was used. To confirm the computational logic of SF, it was confirmed whether a reasonable calculation was made by calculating with a nuclide spectrum.

Radioactive contamination of soil on the site of a nuclear facility has a characteristic that radioactive nuclides are adsorbed into the pores between soil particles, making it quite difficult to decontaminate. For this reason, research on the development of various decontamination processes is being actively conducted. In this study, among various decontamination studies, a soil decontamination process using supercritical carbon dioxide was presented. The decontamination process uses supercritical carbon dioxide as the main solvent, which has a higher penetration power than other materials. Therefore, the process consists of the process of desorbing and extracting the target radionuclides between particles of soil. However, since nuclides exist as ions in the soil, polar chelating ligand material was introduced as an additive to nonpolar supercritical carbon dioxide for smooth chemical reactions in the soil. Thereafter, from the viewpoint of improving process continuity and efficiency, an alcohol material was introduced as an auxiliary solvent for liquefaction of chelating ligand in a solid state. Through prior research on the selection of a solvent for liquefaction of chelating ligand, ethanol and 2-propanol were finally selected based on whether the chelating ligand was dissolved. However, if the auxiliary solvent in which the chelating ligand is dissolved is to be combined with radionuclides in the soil, it must first be well dissolved in supercritical carbon dioxide, the main solvent. Therefore, in this study, the solubility of ethanol and 2-propanol in supercritical carbon dioxide was measured and the suitability was evaluated. The temperature conditions were carried out at 40°C, the same as the previously designed decontamination process, and the measurement was conducted by adjusting the pressure and volume through a syringe pump and a variable volume device. In addition, solubility was measured based on the observation of the ‘cloud point’ in which the image becomes cloudy and then bright. As a result of the experiment, several solubility points were measured at a pressure of 150 bar or less. If the flow rate ratio of supercritical carbon dioxide and auxiliary solvent derived from the results is applied to the soil decontamination process, it is expected that the process efficiency will increase in the future.

Korea faces decommissioning the nation’s first commercial nuclear power plant, the Kori-1 and Wolseong-1 reactors. In addition, other nuclear power plants that will continue to operate will also face decommissioning over time, so it is essential to develop independent nuclear facility decommissioning and site remediation technologies. Among these various technologies, soil decontamination is an essential not only in the site remediation after the decommissioning of the highly radioactive nuclear facility, but also in the case of site contamination caused by an accident during operation of the nuclear facility. But the soil, which is a porous material, is difficult to decontaminate because radionuclides are adsorbed into the pores. Therefore, with the current decontamination technology, it is difficult to achieve the two goals of high decontamination efficiency and secondary waste reduction at the same time. In this study, a soil decontamination process with supercritical carbon dioxide as the main solvent was presented, which has better permeability than other solvents and is easy to maintain critical conditions and change physical properties. Through prior research, a polar chelating ligand was introduced as an additive for smooth extraction reaction between radionuclides present as ions in soil and nonpolar supercritical carbon dioxide. In addition, for the purpose of continuity of the process, a candidate group of auxiliary solvents capable of liquefying the ligand was selected. In this research evaluated the decontamination efficiency by adding the selected auxiliary solvent candidates to the supercritical carbon dioxide decontamination process, and ethanol with the best characteristics was selected as the final auxiliary solvent. In addition, based on the decontamination effect under a single condition of the auxiliary solvent found in the Blank Test process, the possibility of a pre-treatment leaching process using alcohol was tested in addition to the decontamination process using supercritical carbon dioxide. Finally, in addition to the existing Cs and Sr, the possibility of decontamination process was tested by adding U nuclides as a source of contamination. As a result of this research, it is expected that by minimizing secondary waste after the process, waste treatment cost could be reduced and the environmental aspect could be contributed, and a virtuous cycle structure could be established through reuse of the separated carbon dioxide solvent. In addition, adding its own extraction capacity of ethanol used for liquefaction of solid-phase ligands is expected to maximize decontamination efficiency in the process of increasing the size of the process in the future.

The management before disposal of spent nuclear fuel is an essential process for safe management. It is important to determine the amount of nuclide inventory in order to ensure the integrity of spent nuclear fuel, as radiation generated from the nuclides is generated along with residual heat in the spent nuclear fuel. Based on the data on the characteristics of spent nuclear fuel generated in Korea, the correlation equation between burnup and enrichment was derived by referring to overseas cases (Sweden). Source term analysis was performed using the SCALE ORIGEN ARP code by securing the burnup history of nuclear fuel. Calculation was performed by inputting the combustion history of the fuel WH14×14 and WH17×17 as a reference for CE16×16 spent fuel. Through this study, the relationship was identified using the burnup, enrichment, and cooling time factors that influence the characteristics of spent nuclear fuel. In addition, the total source and spectrum data from neutrons and gamma sources were used to find out the characteristics of fuel.

After the Treaty on the Non-Proliferation of Nuclear Weapons (NPT) was signed, Korea is undergoing nuclear inspection by the International Atomic Energy Agency (IAEA) as a non-nuclear-armed state. By the inspection, nuclear material measurement and management have been carried out according to safety measures. Uranium dioxide, a major component of nuclear fuel, is a material that naturally oxidizes at room temperature, yielding a volume change. In this case, it will have an impact on the management of nuclear material measurement, and a model for predicting this will be required. At room temperature, an oxide film is grown by oxygen diffusion on the surface of uranium dioxide, and if the thickness of the oxide film is predicted based on this, the volume change of uranium dioxide can also be predicted. In relation to this, Ghargozloo’s ionic diffusion oxidation model exists. Therefore, in this paper, an modified oxidation model based on Ghargozloo’s oxygen diffusion in uranium dioxide is presented and the volume change of uranium dioxide due to oxidation is predicted.

As the number of nuclear power plants whose design life has expired worldwide increases, the attempts are continuing to complete the project of nuclear back-end cycle, the last task of the nuclear industry. Decontamination is essential in the process of dismantling nuclear facilities and restoration sites to remove all or some of the regulatory controls from an authorized facility. Among radioactive wastes, particularly contaminated soil is characterized by difficult physical decontamination because radionuclides are adsorbed between soil particles, that is, pores. Therefore, chemical decontamination is mainly used, which has the disadvantage of generating a lot of secondary waste. In order to overcome these disadvantages, an eco-friendly soil decontamination process is being developed that can drastically reduce the amount of secondary waste generated by using supercritical carbon dioxide. Supercritical carbon dioxide can easily control its physical properties and has both liquid and gas properties. However, since supercritical carbon dioxide is non-polar, additives are needed to extract polar metal ions, which are the goal of decontamination. Therefore, ligand with both CO2-philic and metal binding regions was selected. In previous studies, the decontamination efficiency of soil was evaluated by reacting contaminated soil with solid ligand and co-ligand at once. When solid ligands were used, the decontamination efficiency was lower than expected, which was expected because chemical substances were somewhat difficult to exchange in the closed process. In this study, in order to increase the efficiency of the decontamination process, the need for a process of liquefying ligand and continuously flowing it has been raised. Therefore, a co-solvent that dissolves well at the same time in SCCO2, ligand, and co-ligand was selected. In the selection process, a total of eight substances were selected by dividing into six polar substances and two non-polar substances through various criteria such as economic feasibility, eco-friendliness, and harmlessness. Thereafter, ethanol was finally selected through solubility evaluation for SCCO2 and additives. It is expected that a more effective decontamination process can be constructed when the additive is liquefied using a solvent selected from the results of this study.

Today, the domestic and international nuclear power industry is experiencing an acceleration in the scale of the nuclear facility decommissioning market. This phenomenon is also due to policy changes in some countries, but the main reason is the rapid increase in the proportion of old nuclear power plants in the world, mainly in countries that introduced nuclear power plants in the early stages. Decontamination is essential in the process of decommissioning nuclear facilities. Among various decontamination targets, radionuclides are adsorbed between pores in the soil, making physical decontamination quite difficult. Therefore, various chemical decontamination technologies are used for contaminated soil decontamination, and the current decontamination technologies have a problem of generating a large amount of secondary wastes. In this study, soil decontamination technology using supercritical carbon dioxide is proposed and aimed to make it into a process. This technology applies cleaning technology using supercritical fluids to decontamination of radioactive waste, it has important technical characteristics that do not fundamentally generate secondary wastes during radioactive waste treatment. Supercritical carbon dioxide is harmless and is a very useful fluid with advantages such as high dissolution, high diffusion coefficient, and low surface tension. However, since carbon dioxide, a non-polar material, shows limitations in removing polar and ionic metal wastes, a chelating ligand was introduced as an additive. In this study, a ligand material that can be dissolved in supercritical carbon dioxide and has high binding ability with polar metal ions was selected. In addition, in order to increase the decontamination efficiency, an experiment was conducted by adding an auxiliary ligand material and ultrasonic waves as additives. In this study, the possibility of liquefaction of chelating ligands and auxiliary ligands was tested for process continuity and efficiency, and the decontamination efficiency was compared by applying it to the actual soil classified according to the particle size. The decontamination efficiency was derived by measuring the concentration of target nuclides in the soil before and after decontamination through ICP-MS. As a result of the experiment, it was confirmed that the liquefaction of the additive had a positive effect on the decontamination efficiency, and a difference in the decontamination efficiency was confirmed according to the actual particle size of the soil. Through this study, it is expected that economic value can be created in addition to the social value of the technology by ensuring the continuity of the decontamination process using supercritical carbon dioxide.

Currently, the interim storage pools of spent fuels in South Korea are expected to become saturated from 2024. It is required to prepare an operation plan of a domestic dry storage facility during a long-term period, with the researches on safety evaluation methods. This study modified the FRAPCON code to predict the spent fuel integrity evaluation such as the axial cladding temperature, the hoop stress and hydrogen distribution in dry storage. The cladding temperature in dry storage was calculated using the COBRA-SFS code with the burnup information which was calculated using the FRAPCON code. The hoop stress was calculated using the ideal gas equation with spent fuel information such as rod internal pressure. Numerical analysis method was used to calculate the degree of hydrogen diffusion according to the hydrogen concentration and temperature distribution during a dry storage period. Before 50 years of dry storage, the cladding temperature and hoop stress decreased rapidly. However, after 50 years, they decreased gradually and the cladding temperature was below 400 K. The initial temperature distribution and hydrogen concentration showed a parabolic line, but hydrogen was transferred by the hydrogen concentration and temperature gradient over time.