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.

To secure approval for a decommissioning plan in Korea, it is essential to evaluate contamination dispersion through groundwater during the decommissioning process. To achieve this, licensees must assess the groundwater characteristics of the facility’s site and subsequently develop a groundwater flow model. It is worth noting that Combustible Radioactive Waste Treatment Facility (CRWTF) is characterized by their simplicity and absence of liquid radioactive waste generation. Given these facility characteristics, the groundwater flow model for CRWTF utilizes data from neighboring facilities, with the feasibility of using reference data substantiated through comparative analysis involving groundwater characteristic testing and on-site modeling. To enable a comparison between the actual site’s groundwater characteristics and the referenced modeling, two types of hydraulic constant characterization tests were conducted. First, hydraulic conductivity was determined through long-term pumping and recovery tests. The ‘Theis’ and ‘Cooper-Jacob’ equations, along with the ‘Theis recovery’ equation, were applied to calculate hydraulic conductivity, and the final result adopted the average of the calculated values. Secondly, a groundwater flow test was conducted to confirm the alignment between the main flow direction of the referenced model and the groundwater flow in the CRWTF, utilizing the particle tracking technique. The evaluation of hydraulic conductivity from the hydraulic constant test revealed that the measured value at the actual site was approximately 1.84 times higher than the modeled value. This variance is considered valid, taking into consideration the modeling’s calibration range and the fact that measurements were taken during a period characterized by wet conditions. Furthermore, a close correspondence was observed between the groundwater flow direction in the reference model (ranging from 90° to 170°) and the facility’s actual flow direction (ranging from 78° to 95°). The results of reference data for the CRWTF, based on the nearby facility’s model, were validated through the hydraulic properties test. Consequently, the modeling data can be employed for the demolition plan of CRWTF. It is also anticipated that these comparative analysis methods will be instrumental in shaping the groundwater investigation plans for facilities with characteristics similar to CRWTF.

Various types of radioactive liquid and solid wastes are generated during the operation and decommissioning of nuclear power plants. To remove radionuclides Co-60, Cs-137 etc. from a liquid waste, the ion-exchange process based on organic resins has been commonly used for the operation of nuclear facilities. Due to the considerations for the final disposal of process endproduct, other treatment methods such as adsorption, precipitation using some inorganic materials have been suggested to prepare for large amounts of waste during decommissioning. This study evaluated sintering characteristics for radioactive precipitates generated during the liquid waste treatment process. The volume reduction efficiency and compressive strength of sintered pellets were the major parameters for the evaluation. Major components of a simulated precipitate were some coagulated (oxy) hydroxides containing light elements, such as Si, Al, Mg, Ca, and zeolite particles. Green pellets compressed to around 100 MPa were heated at a range of 750~850°C to synthesize sintered pellets. It was observed that the volume reduction percentages were higher than 50% in the appropriate sintering conditions. The volume reduction was caused by the reduction of void space between particles, which is an evidence of partial glassification and ceramization of the precipitates. This result can also be attributed to conversion reactions of zeolite particles into other minerals. The compressive strength ranged from 6 to 19 MPa. These results also showed a significant correlation with the volume reduction of sintered body. Although our lab-scale experiments showed many benefits of sintering for the precipitates, optimized conditions are needed for large-scale practical applications. Evaluation of sintering characteristics as a function of pellet size and further testing will be conducted in the future.

Various dry active wastes (DAWs) have been accumulated in nuclear power plants since the DAWs are mostly combustible. KAERI has developed a thermochemical treatment process for the used decontamination paper as an operational waste to substitute for incineration process and to decontaminate radionuclides from the DAWs. The thermochemical process is composed of thermal decomposition in a closed vessel, chlorination of carbonated DAWs, separation of soluble chlorides captured in water by hydroxide precipitation, and immobilization of the precipitate. This study examined the third and fourth steps in the process to immobilize Co-60 by fabricating a stable wasteform. Precipitation behaviors were investigated in the chloride solution by adding 10 M KOH. It was shown that the precipitates were composed of Mg(OH)2 and Al(OH)3. Then, the glass-ceramic wasteform for the precipitates were produced by adding additive mixtures in which silica and boron oxide were blended with various ratios. The wasteform was evaluated in terms of volume reduction ratio, bulk density, compressive strength, and leachability.

Plasma melting technology uses electrical arc phenomena such as lightning to create hightemperature sparks of about 1,600 degrees or more to meet waste disposal requirements through treatment and reduction without distinguishing radioactive waste generated during nuclear power plant operation and dismantling according to physical characteristics. Decommissioning radioactive waste scabbed concrete occurs when polishing and cutting the contaminated structure surface to a certain depth. In this study, Scabbed concrete containing paint was confirmed for volume reduction and disposal safety using plasma treatment technology, and it is planned to be verified through continuous empirical tests.

The thermal treatment of radioactive waste attracts great attention. The thermal treatment offers lots of advantages, such as significant volume reduction, hazard reduction, increase of disposal safety, etc. There are various thermal technologies to waste. The developed technologies are calcination, incineration, melting, molten salt oxidation, plasma, pyrolysis, synroc, vitrification, etc. The off-gas treatment system is widely applied in the technologies to increase the safety and operation efficiency. The thermal treatment generates various by-product and pollutants during the process. The dust or fly ash are generated as a particulate from almost every radioactive waste. The treatment of PVC related components generates hydrogen chloride, which usually brings corrosion of facility. The treatment of rubber and spent resin generates sulfur oxide, SOx. The treatment of nitrile rubber generates nitrogen oxide, NOx. The incomplete combustion of radioactive waste usually generates carbon oxide, COx. The process temperature also affects the generation of off gas, such as NOx and/or COx. Various off gas treatment components are organized for the proper treatment of the previously mentioned materials. In this study systematical review on off gas treatment will be reported. Also, worldwide experiences and developed facility will be reported.

The treatment process for Spent Filter(SF) of Kori-1 was developed that includes the following : 1) Taking out by robot system 2) Screening by ISOCS 3) Collection of representative samples using a sampling machine 4) Compression 5) Immobilization 6) Packaging and nuclide analysis and 7) Delivery/disposal. Although the robot system, ISOCS, sampling machine and immobilization facility are essentially required for building the above processing but decision to build the compression system and nuclide analysis system must be made after reviewing the need and cost benefit for their construction. In addition, for effcient SF treatment, it is necessary to determine the nuclide concentration range of the SF to which immobilization will be applied. In this study, a cost benefit analysis was performed on existing and alternative methods for processes related to compression treatment, nuclide analysis and immobilization methods, which are greatly affected by economics and efficiency according to the design. First, although the disposal cost is reduced with reducing the number of packaging drums by compressed and packaged but the expected benefits not be equal to or greater than the cost invested in building a compression system. As a result, non-compressed treatment of SF is expected to be economical because the construction cost of compression system is more expensive than the benefits of reducing disposal costs by compression. Second, a cost benefit analysis of direct and indirect nuclide analysis methods was performed. For indirect analysis, scaling factors should be developed and the drum scanner suitable for the analysis for DAW should be improved. As a result, direct analysis applied grouping options is expected to be more economical than indirect analysis requiring the cost for developing scaling factors and improving the scanner. Third, it is timeconsuming and inefficient to distinguish and collect filters that are subject to be immobilized according to the waste acceptance criteria among the disorderly stored SFs in the filter rooms. If the benefits of immobilization of the SFs selectively are not greater than the benefits of immobilization of all SFs, it can be economical to immobilize all SFs regardless of the nuclide concentration of them. As a result, it is more economical to immobilize all SFs with various nuclide concentrations than to selectively immobilize them. The conclusion of this study is that it is not only cost-effective but also disposal-effective to design the treatment process of SF to adopt non-compressed processing, direct analysis and immobilization of all SFs.

The type of radioactive waste that may occur in the process of nuclear power plant dismantling can be classified into solid, liquid, gas, and mixed waste. The amount of these wastes must be defined in the Final Decommissioning Plan for approval of the licensing. Also, in the case of Metal radioactive waste, it is necessary to calculate the generation amount in order to treat radioactive waste at a Radioactive Waste Treatment Facility (RWTF). Since a large quantity of metal radioactive waste is generated during the decommissioning of a nuclear power plant, the application of a metal melter for reduction is considered. The metal waste is heated to a temperature above the melting point and separated into liquid and gas forms. Nuclides existing on the surface of metal waste vaporize in a melting furnace to become dust or collect in sludge. Nonvolatile nuclides such as Co, Fe and Mn remain in ingot, but other nuclides can be captured and reduced with dust and sludge. And the types of melting furnaces to be applied can be broadly classified into Atmospheric Induction Melter (AIM) and Vacuum Induction Melter (VIM). Therefore, this review intends to compare the two types of metal furnaces to be included in RWTF.

When aluminum is in an alkaline state, the aluminum oxide film surrounding aluminum is dissolved and moisture penetrates the exposed aluminum surface, causing corrosion of aluminum. At this time, hydrogen gas is generated and there is a risk of explosion due to the generated hydrogen gas. Aluminum radioactive waste is difficult to permanently dispose of because it does not meet the standards for the acquisition of low- and intermediate-level radioactive waste cave disposal facilities currently managed and operated by the Korea Nuclear Environment Corporation. However, because of this risk, it is necessary to study how to safely treat and dispose aluminum waste. In this study, overseas cases were investigated and analyzed to ensure the safety of aluminum waste disposal, and the current status of aluminum radioactive waste generated during decommissioning of the Korea Research Reactor 1&2 and a treatment plan to secure disposal suitability were presented. The process of removing a little remaining oxygen in molten steel during the reduction of iron oxide in the iron refining process is called deoxidation, and a representative material used for deoxidation is aluminum. In the case of metal melting decontamination, which is one of the decontamination processes of radioactive metal waste, a method of treating aluminum waste by using aluminum as a deoxidizer is proposed.

During the operation of nuclear power plant (NPP), the concentrates and spent resin are generated. They show relatively high radioactivity compared to other radioactive waste, such as dry active waste, charcoals, and concrete wastes. The waste acceptance criteria (WAC) of disposal facility defines the structure and property of treated waste. The concentrates and spent resin should be solidified or packaged in high integrity container (HIC) to satisfy the WAC in Korea. The Kori NPP has stored history waste. The large concrete package with solidified concentrates and spent resin. The WAC requires identification of 18 properties for the radioactive waste. Since some of the properties are not clearly identified, the large concrete packages could not satisfy the WAC in this moment. The generation of the large concrete package (rectangular type and cylindrical type), pretreatment of the package, treatment of inner drum, process development for clearance waste, etc. will be discussed in this paper. In addition, the conceptual design of whole treatment process will be discussed.

Economical radioactive soil treatment technology is essential to safely and efficiently treat of high-concentration radioactive areas and contaminated sites during operation of nuclear power plants at home and abroad. This study is to determine the performance of BERAD (Beautiful Environmental construction’s RAdioactive soil Decontamination system) before applying magnetic nanoparticles and adsorbents developed by the KAERI (Korea Atomic Energy Research Institute) which will be used in the national funded project to a large-capacity radioactive soil decontamination system. BERAD uses Soil Washing Process by US EPA (402-R-007-004 (2007)) and can decontaminate 0.5 tons of radioactive soil per hour through water washing and/or chemical washing with particle size separation. When contaminated soil is input to BERAD, the soil is selected and washed, and after going through a rinse stage and particle size separation stage, it discharges decontaminated soil separated by sludge of less than 0.075 mm. In this experiment, the concentrations of four general isotopes (A, B, C, and D which are important radioisotopes when soil is contaminated by them.) were analyzed by using ICP-MS to compare before and after decontamination by BERAD. Since BERAD is the commercial-scale pilot system that decontaminates relatively large amount of soil, so it is difficult to test using radioactive isotopes. So important general elements such as A, B, C, and D in soil were analyzed. In the study, BERAD decontaminated soil by using water washing. And the particle size of soil was divided into a total of six particle size sections with five sieves: 4 mm, 2 mm, 0.850 mm, 0.212 mm, and 0.075 mm. Concentrations of A, B, C, and D in the soil particles larger than 4 mm are almost the lowest regardless of before and after decontamination by BERAD. For soil particles less than 4 mm, the concentrations of C and D decreased constantly after BERAD decontamination. On the other hand, the decontamination efficiency of A and B decreased as the soil particle became smaller, but the concentrations of A and B increased for the soil particle below 0.075 mm. As a result, decontamination efficiency of one cycle using BERAD for all nuclides in soil particles between 4 mm and 0.075 mm is about 45% to 65 %.

In light of recent significant seismic events in Korea and worldwide, there is an urgent need to reevaluate the adequacy of seismic assessments conducted during facility construction. This study reexamines the ongoing viability of the Safety Shutdown Earthquake (SSE) criteria assessment for the Combustible Radioactive Waste Treatment Facility (CRWTF) site at the Korea Atomic Energy Research Institute (KAERI), originally established in 1994. To validate the SSE assessment, we delineated 13 seismic structure zones within the Korean Peninsula and employed two distinct methodologies. Initially, we updated earthquake occurrence data from 1994 to the present year (2023) to assess changes in the site’s horizontal maximum earthquake acceleration (g). Subsequently, we conducted a comparative analysis using the same dataset, contrasting the outcomes derived from the existing distance attenuation equation with those from the most recent attenuation equations to evaluate the reliability of the applied attenuation model. The Safety Shutdown Earthquake (SSE) criterion of 0.2 g remains unexceeded, even when considering recent earthquake events since the original evaluation in 1994. Furthermore, when applying various assessment equations developed subsequently, the maximum value obtained from the previously utilized ‘Donvan and Bornstein’ attenuation equation is 0.1496 g, closely resembling the outcome derived from the recently employed ‘Lee’ reduction equation of 0.1451 g. The SSE criteria for CRWTF remain valid in the current context, even in light of recent seismic occurrences such as the 2016 Gyeongju earthquake. Additionally, the attenuation equation employed in the evaluation consistently yields conservative results when compared to methodologies used in recent assessments. Consequently, the existing SSE criteria remain valid at present. This study is expected to serve as a valuable reference for confirming the SSE criterion assessment of similarly constructed facilities within KAERI.

Dry active wastes (DAWs) are combustible waste generated during the operation and decommissioning of nuclear facilities, and are known to be generated in the amount of approximately 10,000 to 40,000 drums (based on 200 L) per unit. It consists of various types of protective clothing, paper, and plastic bags, and is stored in radioactive waste storage facilities. Therefore, reducing the volume of DAWs is an important issue in order to reduce storage costs and utilize the limited space of waste storage facilities. Heat treatment such as incineration can dramatically reduce the volume of waste, but as the waste is thermally decomposed, CO2, a global warming gas, is generated and there is a risk of emissions of harmful gases including radionuclides. Therefore, a heat treatment process that minimizes the generation of CO2 and harmful gases is necessary. One of the alternatives to incineration is to carbonize DAWs, dispose of carbonized materials below the release standard as non-radioactive waste, and selectively separate and stabilize inorganic components, including radionuclides, from carbonized DAWs. In this study, 13 types of DAWs generated from nuclear power plants were selected and their thermal decomposition characteristics were investigated to design a heat treatment process that replaces incineration. As a result of TGA analysis, the temperature at which thermal decomposition of each waste begins is 260-300°C for cotton, 320-330°C for paper, 315-420°C for synthetic fiber, 350°C for latex gloves. The mass of most samples decreased to less than 1 % of the initial weight after heat treatment, and dust suit and latex gloves had residues of 13.83% and 13.71% of the initial mass, respectively. The metal components of the residue produced after heat treatment of the sample were analyzed by EDS. According to the EDS results, cotton contains Ca and Al, paper contains Ca, Al and Si, synthetic fiber contains Ca, Cu and Ti, latex gloves contain Ca and Mg. Additionally, ICP analysis was performed to quantify the inorganic components. These results are expected to be applicable to the processing of DAW generated at nuclear facilities in the future.

Concentrated effluent and spent ion exchange resins (IERs) from nuclear power plants (NPPs) were generated prior to the establishment of a disposal facility site and waste acceptance criteria have been temporarily stored at the NPPs because their suitability for disposal has not been confirmed. In particular, at the Kori Unit 1, which was the first to start the commercial operation in South Korea, the initially generated concentrated effluent and IERs are repackaged in large size of concrete containers and stored without provided regulation standard. The concentrated effluent is package as cementitious form in 200 L drums and repackaged in concrete containers, case of the IERs were solidified or dehydrated and repackaged in round concrete container. In this study, we review and propose a disposal plan for concentrated effluent and IERs repackaging drums that have not been confirmed to be suitable for disposal from the first operating nuclear power plant, Kori Unit 1, 2. First, the concentrated effluent was stored in four 200 L drums respectively, and then, it was again stored in concrete container and which was poured on top using grouted concrete. Therefore, the process was required by cutting concrete container for extracting the internal drums at first. Internal radioactive waste should be crushed to the suitable waste criteria and solidified, finally disposal in to the polymer concrete high integrity container (PC-HIC). IER was repackaged and disposal in square type of 200 L concrete drums respectively covered the cap. So, extracting the internal drums should be extracted after removing the cap of external concrete container. Cement solidification drums can be crushed and re-solidified or disposed in the PC-HIC. Stored IER after dehydrated can be disposal in PC-HIC. In conclusion, the container was used as a package that repackaging the concentrated effluent and IER was separated into two different types of waste depending on the level of contamination of radioactivity, the polluted area is disposed of as radioactivity contamination or the unspoiled area will be treated as self-disposal waste.

Among the nuclear power plant facility improvement projects, out of a total 10 replacement reactor vessel closure head (RRVCH), five have been replaced, starting with Gori Unit 1, and five, including Hanul Unit 1, Hanbit Units 5 and 6, and Hanul Units 3 and 4 will be replaced in the future. This paper presents the method of treating Latch Housing among radioactive waste generated during the replacement of Hanul Unit 2 (February 2023). Latch Housing controls the control rod by receiving magnetic force from the CRDM’s Coil Stack. Located in the Old Reactor Vessel Head (ORVH) Hot Spot, the range of measured radiation dose rate was 0.3 to 0.8 mSv h-1 (up to 4.5 mSv h-1). The amount of radioactive waste generated was 35.8 Baler-Drum (based on 200L), and the order of treatment was to cut into the Omega Seal of the CRDM, the CRDM and Latch Housing Transfer, the boundary of the CRDM and Latch Housing, the Rod Travel Housing, the Motor Housing and the Latch Assembly, and then transfer and Drumming. In the United States, out of 93 operating reactors, 31 reactor vessel heads have been replaced and 19 reactor vessel heads are scheduled to be replaced. In Korea, 25 reactors are in operation, and two reactors have been permanently shut down. Among them, the nine old reactors for more than 30 years (as of September 2021) are expected to achieve ALARA and reduce radwaste management costs through the management method applied to replace the reactor vessel head.

Domestic nuclear power plants can affect the environment if multiple devices are operated on one site and even a trace amount of pollutants that may affect the environment after power generation are simultaneously discharged. Therefore, not only radioactive substances but also ionic substances such as boron should be discharged as minimally as possible. We adopted pilot CDI and SD-ELIX sytem to separating and concenrating of boron containing nulcear power plant discharge water. The boron concentration of the initial inflow water tended to decrease over time. The water quality of concentrated water also reached its peak until the initial 60 minutes, but tended to decrease in line with the decrease in the inflow water concentration. The boron removal rate was in the range of 85 to 99% with respect to the initial boron concentration of 15 to 25 mg/L. On the other hand, performance degradation due to the use of electrochemical modules is also observed, and regeneration through low ion-containing water cleaning effective. We shortened processing time by considering the optimal flow rate conditions and conductivity conditions and converting electrochemical modules into series or parallel.

Radioactive iodine-129, a byproduct of nuclear fission in nuclear power plants, presents significant environmental and health risks due to its high solubility in water and volatility. Iodine-129, with its half-life of 1.57×1017 years, necessitates safe management and disposal. Therefore, safely capturing and managing I-129 during spent nuclear fuel reprocessing is of paramount importance. To address these challenges, various glass waste forms containing silver iodide have been developed, such as borosilicate, silver phosphate, silver vanadate, and silver tellurite glasses. These glasses effectively immobilize iodine, but the high cost of silver raises affordability concerns. This study introduces CuI·Cu2O·TeO2 glass waste forms for iodine immobilization, a novel approach. The cost-effectiveness of copper, in contrast to silver, makes it an attractive alternative. The CuI·Cu2O·TeO2 glass waste forms were synthesized with varying CuI content (x) in (1-x)(0.3Cu2O·0.7TeO2) glass matrices. Xray diffraction (XRD) confirmed amorphous structures, and X-ray fluorescence (XRF) quantified composition. X-ray photoelectron spectroscopy (XPS) and Raman spectroscopy provided insights into structural properties. Durability assessments using a 7-day product consistency test (PCT-A) and inductively coupled plasma-mass spectrometry (ICP-MS) revealed compliance with U.S. glass regulations, making CuI·Cu2O·TeO2 glasses a promising choice for iodine immobilization in radioactive waste.

The solid-state chemistry of uranium is essential to the nuclear fuel cycle. Uranyl nitrate is a key compound that is produced at various stages of the nuclear fuel cycle, both in front-end and backend cycles. It is typically formed by dissolving spent nuclear fuel in nitric acid or through a wet conversion process for the preparation of UF6. Additionally, uranium oxides are a primary consideration in the nuclear fuel cycle because they are the most commonly used nuclear fuel in commercial nuclear reactors. Therefore, it is crucial to understand the oxidation and thermal behavior of uranium oxides and uranyl nitrates. Under the ‘2023 Nuclear Global Researcher Training Program for the Back-end Nuclear Fuel Cycle,’ supported by KONICOF, several experiments were conducted at IMRAM (Institute of Multidisciplinary Research for Advanced Materials) at Tohoku University. First, the recovery ratio of uranium was analyzed during the synthesis of uranyl nitrate by dissolving the actual radioisotope, U3O8, in a nitric acid solution. Second, thermogravimetric-differential thermal analysis (TG-DTA) of uranyl nitrate (UO2(NO3)2) and hyper-stoichiometric uranium dioxide (UO2+X) was performed. The enthalpy change was discussed to confirm the mechanism of thermal decomposition of uranyl nitrate under heating conditions and to determine the chemical hydrate form of uranyl nitrate. In the case of UO2+X, the value of ‘x’ was determined through the calculation of weight change data, and the initial form was verified using the phase diagram for the U-O system. Finally, the formation of a few UO2+X compounds was observed with heat treatment of uranyl nitrate and uranium dioxide at different temperature intervals (450°C-600°C). As a result of these studies, a deeper understanding of the thermal and chemical behavior of uranium compounds was achieved. This knowledge is vital for improving the efficiency and safety of nuclear fuel cycle processes and contributes to advancements in nuclear science and technology.

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.