The rise of nuclear power plants to meet escalating global energy needs has made environmental pollution including the contamination of uranium due to improper disposal of radioactive wastewater during uranium milling and mining processes. Adsorption, a water purification method known for its fast kinetics, high selectivity, and ease of use, has emerged as a popular choice for the treatment of radioactive wastewater. In response to the critical need for the purification of radioactive wastewater contaminated with U(VI), this review provides a comprehensive summary of the various types of materials, synthetic methods, and adsorption mechanisms used for the purification process. The materials are categorized into four main groups: organic, inorganic, composite/nanomaterials, and framework materials. To enhance the adsorption capacity for U(VI), researchers have explored physical and chemical modifications as well as the development of organic-inorganic hybrids. The improved adsorption performance resulting from these modifications is mostly attributed to electrostatic interaction, surface complexation, and ion exchange mechanisms. However, despite the present understanding of the processes involved, further research is still needed to fully determine the optimal approach for purifying contaminated radioactive wastewater.

The mobility of uranium (U) in the environment of a deep geological repository is controlled by various geochemical conditions and parameters. In particular, oxidation state of uranium is considered as a major factor to control the mobility of uranium in most of geological environments. In this study, therefore, we investigated the mobility of uranium in a deep geological repository by a natural analogue approach using a uranium deposit in the Ogcheon Metamorphic Belt (OMB). Uranium contents of rock samples from the study site ranged from 1.3 to 71 ppm (average 17.4 ppm). Uranium minerals found in the study site were mostly uraninite (UIVO2+x) and uranothorite ((UIV, Th)SiO4). The concentrations of U in the groundwater samples were very low (0.025~0.690 ppb) even though redox conditions are weakly oxidizing. Calculation results for U speciation in groundwater samples showed that major dissolved uranium species in the groundwater samples are mainly as calcium uranyl (UO2 2+) carbonate complexes such as Ca2UO2(CO3)3(aq) and CaUO2(CO3)3 2-. However, the activity ratios between 234U and 238U (AR(234U/238U)) showed U behavior in reducing conditions although the groundwater conditions were not reducing conditions and major dissolved U species were U(VI) species. Results from electron microscopic analyses for rock samples showed that major uranium minerals were U(IV) minerals such as uraninite and uranothorite. We could not identify other uranyl minerals and altered minerals from uraninite. This means that the geochemical condition of the study site has been maintained a reducing condition although the groundwater condition was a weakly oxidizing condition. Thus, the dissolution of uranium is strongly limited by the low solubility of uraninite. It is not obvious how the reducing condition of the study site has been maintained. Reducing agents such as pyrite, organic materials, and reducing bacteria might contribute to maintaining the reducing condition although further studies will be necessary. Results from this study imply that uranium mobility will be greatly limited by low dissolution of uraninite into groundwater if the reducing condition is well reserved. This limited mobility of uranium will be also contributed by low possibility of uraninite alteration into uranyl minerals which have a higher solubility than uraninite.

The stabilization technology for the damaged spent fuel is being developed to process the damaged fuel into sound pellet suitable for dry re-fabrication. It requires several treatments including oxidative decladding followed by reduction treatment for oxidized powder closely related to the quality of oxidized powders for pellet fabrication. For the development of operating condition for the reduction treatment, in this study, we evaluated the effect of air-cylinder based vertical shaking previously applied to oxidative decladding on powder reduction. For U3O8 of 50-100 g, the reduction test were applied with and without vertical shaking at 700°C under reduction atmosphere (Ar + 4%H2) and the concentration of hydrogen in effluent was measured to evaluate the reduction reaction. It was found that the vertical shaking system has allowed the reaction time of 50 g and 100 g U3O8 reduced by 33% compared to the test in static mode. Based on XRD analysis, the better crystallinity of the products was also achieved.

Given the limited terrestrial reserves of uranium (about 4.6 million tons), exploring alternative resources is essential to ensure the long-term supply and sustainability of nuclear energy. Uranium extraction from seawater (UES) is a potential solution to this issue since the amount of uranium dissolved in seawater (about 4.5 billion tons) is approximately 1000 times that of terrestrial reserves. However, the ultra-low concentration of uranium in seawater (about 3.3 ppb) makes it a challenging task to make UES economically feasible. This paper provides an overview of the current status of UES technology, which has evolved over the past seven decades. Starting from inorganic adsorbents such as hydrous titanium oxide in the 1960s, amidoxime-based fibrous adsorbents gained the most attention until the early 2010s due to their ease of deployment in actual seawater conditions and high affinity for uranium. Nowadays, research on organic adsorbents with microstructures is prevailing due to their ability to easily control surface area and compositions. In addition, this study identifies the key issues that need to be addressed to make UES technology economically viable.

The characterization of nuclear materials is crucial for global nuclear safeguards efforts, as these materials can potentially be used for illicit purposes. In this study, we evaluated the applicability and performance of the In-Situ Object Counting System (ISOCS) equipment for the characterization and quantification of uranium, including uranium pellets and radioactive wastes. Our methodology involved using ISOCS to measure samples with different enrichments and total amounts of uranium, and to analyze the results in order to evaluate the ISOCS’s effectiveness in accurately characterizing the various uranium samples. To this end, we compared the ISOCS results with those of the Multi-Group Analysis for Uranium (MGAU) system, which is currently used in the field of international safeguards. The results of this study showed that the ISOCS was sensitive enough to analyze small amounts of uranium pellet, with %differences ranging from -0.7% to 19%. However, when analyzing shielded nuclear materials like in concrete waste, the uncertainty was relatively high, with %differences ranging from 11% to 67%. On the other hand, the MGAU system was unable to analyze uranium for the same spectrum, indicating the superiority of the ISOCS in terms of usability. The ISOCS instrument was also found to be effective in analyzing uranium in various types of samples without the need of standard sources. Overall, the findings of this study have important implications for the development of more effective safeguards strategies for the characterization of nuclear materials. The ISOCS instrument could be a reliable tool for analyzing nuclear materials, contributing to global safeguards efforts to reduce the risk of nuclear proliferation.

Nuclear spent fuel (SNF) disposal in deep geological repositories is considered as one of sound options for the long-term and safe sequestration of radiotoxic SNF and the sustainable use of nuclear energy. The chemical behaviors of various radionuclides originated from SNF should be well understood to evaluate the migrational behaviors of radionuclides and their reactions and interactions with various geochemical components. Formation of secondary minerals, colloids, other insoluble precipitates is of interest since the concentrations of radionuclides in groundwaters can be limited by the solubility of those solid phases. Particularly when evaluating their solubility, the use of well-defined solid materials in terms of chemical composition and molecular structure is crucial to obtain reliable measurement results. In this study, a synthetic calcium uranyl silicate (Ca-U(VI)-silicate, or uranophane) was prepared and characterized by using various analytical methods including powder X-ray diffraction (pXRD), scanning electron microscopy/energy dispersive X-ray spectrometry (SEM/EDX), and vibrational (FTIR and Raman) spectroscopies. Uranyl silicate minerals are significant to the disposal of nuclear wastes. Our simulation demonstrates that uranophane (Ca[UO2SiO3OH]2·5H2O), one having a U:Si ratio of 1:1, can be a mineral species limiting U(VI) solubility under groundwater conditions in Korea. For the preparation of Ca-U(VI)-silicate, we applied a two-step hydrothermal synthetic procedure reported in literature with modification. Briefly, we conclude that the obtained mineral phase is the ‘α-uranophane’; our characterization results show that the structural and spectroscopic properties of the synthetic Ca-U(VI)-silicate agree well with those of α-uranophane. For instance, the pXRD patterns obtained from the solid show nearly identical diffraction peak positions with those from the reference XRD pattern. From IR and Raman spectroscopy it is noticed that the stretching modes of UO2 2+ and SiO4 4- ions result in strong absorption bands in a region of 700 ~ 1,100 cm-1. Elemental compositions of the synthetic solids were also estimated by using EDX analysis, which results in a Ca:U:Si ratio close to 1:2:2 on average. However, we found that it is difficult to obtain good crystallinity of uranophane, which can be observable by using SEM and its image analysis. We believe that this work serves as a model study to provide synthetic routes of radionuclide-related mineral phases and applicable solid phase characterization methods. In the presentation, the potential use of the U(VI)-silicate solid phase for the upcoming groundwater solubility measurements will be discussed. Keywords: Hexavalent Uranium, Silicate

The Korea Atomic Energy Research Institute operates the Nuclear Cycle Experimental Research Facility which has radiation controlled area in the laboratory with the aim of realizing pyroprocessing technology. In this Facility, depleted Uranium feed material and a depleted Uranium mixed with some surrogate material are used for performing experiments. Therefore the facility is using uranium, users should be careful of radiation. This paper will explain the radiation protection of the Nuclear Cycle Experimental Research facility and will also explain how much alpha radiation comes out from the facility. The RMS (Radiation Monitoring System) detector is made by CANBERRA and the model name is ICAM. ICAM RMS is the detector which can detect Alpha Radiation by absorbing the air in the facility. The RMS detector is installed in the HVAC room on the third floor to check the air contamination through the chimney. The RMS is connected to the air ventilation line for detecting Alpha radiation in the whole facility. Experiment was performed for two weeks to check the radiation level and the air ventilation fan continued to operate 24 hours a day. the results are below the required value which is 0.1 Bq/m3, indicating that the facility is safe in terms of radiation safety management.

Concrete waste generated in the result of dismantling a concrete structure in a radiation control area and refractory brick waste generated from uranium pellet sintering furnace are surface-contaminated by uranium particle of which the enrichment is below 5%. These wastes are hard to decontaminate so it was necessary to develop the process for its disposal. So, we developed the Process Control Plan (PCP) for disposal of radioactive concrete waste describing a whole sequence of disposal and inspecting procedures based on the KNF Radioactive Waste Quality Assurance Plan (KN-WQAP) established in 2021. Based on the PCP, we crushed the concrete waste by jaw-crusher. Then we sieved the crushed concrete waste and removed the particle of which size is below 0.3 mm, using sieve-vibrator where the 0.3 mm mesh-sized sieve is installed inside. Before conducting the crush-sieving method based on the PCP, we conducted Process Control Assessment (PCA) based on the KN-WQAP. The purpose of the PCA is to check whether the output of the process satisfies the Acceptance Criteria of Korea Radioactive Waste Agency (KORAD) so that we could confirm the validity of the PCP. The evaluation item of the PCA is a particulate size verification test. The test is passed only if the component ratio of a particle size below 0.2 mm is less than 15% and the particle size below 0.01 mm is less than 1%. The very first 3 drums passed the test, so we began applying the PCP to whole target drums. In the process of conducting the crush-sieving method in earnest, qualified inspectors based on KNWQAP participated conducting sampling, measuring and checking whether a foreign material was included. They tested samples and packaged drums regarding 5 spheres of general, radiological, physical, chemical and biological characteristic. KNF disposed concrete and refractory brick waste by the crush-sieving method so that KNF could take over 100 drums to KORAD in 2021. But, it is needed to be improved that a dust size below 0.3 mm is generated as a secondary waste which needs to be solidified for the final disposal and the work environment is not good enough because of the dust.

A sequential column experiment was conducted for uranium removal of excessively high or highly U-contaminated soils, simultaneously. Two pilot-scale acryl columns with a 24 cm ID and 48 cm length were uniformly packed with each U-contaminated soil (both < 2 mm, 119, and 22.4 Bq/g as initial U-238 activities). A column packed with soil contained very high U constant located first then sequentially located second columns with relatively lower U-contaminated soil. Thus the effluents which passed very high U-contaminated soil and having extremely high dissolved U concentration was directly inflowed the second columns. Both columns initially and respectively flushed with demi water (or condensing water of air conditioner generated from radiation controlled area) to saturate and displace the air from the pore space. Elution was carried out with alkaline and acidic solutions, respectively, and sequentially. The uranium removal efficiencies were found and a comparison was made with the pilot soil flushing experiments. During this study, a new approach to reducing acidic flushing waste which is considered the biggest defect of soil washing/flushing was established, and optimal factors were calculated to demonstrate industrial-scale uranium decontamination of soil with high uranium content.

The decommissioning of nuclear-related facilities at the end of their design life generates various types of radioactive waste. Therefore, the research on appropriate disposal methods according to the form of radioactive waste is needed. This study is about the solidification of uranium contaminated soils that may occur on the site of nuclear facilities. A large amount of radioactively contaminated soil waste was generated during the decommissioning of the uranium conversion plant in KAERI, and research on the proper disposal of this waste has been actively conducted. Numerous minerals in the soil can become glass-ceramic through the phase change of minerals during the sintering process. This method is effective in reducing the volume of waste and the glassceramic waste form has excellent mechanical strength and leaching resistance. In this study, the optimum temperature and time conditions were established for the production of glass-ceramic sintered body of soil. The compressive strength and leachability of the sintered body made by applying the optimal conditions to simulated waste was confirmed. The basic physicochemical properties of simulated soil waste were identified by measuring the pH, moisture content, density, and organic matter content. The elemental compositions in the soil was confirmed by XRF. Soils were classified by particle size, and each sample was compressed with a pressure of 150 MPa or more to prepare a green body. Based on the TG-DSC analysis, an appropriate heating temperature was set (>1,000°C), and the green body was maintained in a muffle furnace for 2~6 hours. The optimal sintering conditions were selected by measuring the compressive strength and volume reduction efficiency of the sintered body for each condition. The difference between the green body and sintered body was observed by XRD and SEM. In the experiments for evaluation of additives, the selected chemical substances were mixed with the soil sample in a rotator. Based on the results of TG-DSC, sintered body was made at 850°C, and the compressive strength and volume reduction were compared. Based on the results, the most effective additive was determined, and the appropriate ratio of the additive was found by adjusting the range of 1~5 wt%. This study was confirmed that the sintered soil waste showed sufficient stability to meet the disposal criteria and effective volume reduction for final disposal.

This study aimed to remove uranium (U(VI)) ions from sulfate-based acidic soil-washing effluent using the ion-exchange method. For effective ion exchange of U(VI) ions under acidic conditions, one chelate resin (Purolite S950) stable under low pH conditions and two anion-exchange resins (Ambersep 400 SO4 and 920U SO4) used in sulfuric acid leaching systems were selected. The exchange performance of the three selected ion-exchange resins for U(VI) ions was evaluated under various experimental conditions, including ion-exchange resin dosages, pH conditions, reaction times, and reaction temperatures. U(VI) ion exchange was consistent with the Langmuir model and followed pseudo-second-order kinetics. Thermodynamic experiments revealed that the U(VI) ion exchange by the ion-exchange resins is an endothermic and spontaneous process. On the other hand, U(VI) ions was effectively desorbed from the ion-exchange resins using 0.5 M H2SO4 or Na2CO3 solution. Overall, on the basis of the results of the present study, we propose that Purolite S950, Ambersep 400 SO4, and Ambersep 920U SO4 are ion-exchange resins that can be practically applied to effectively remove U(VI) ions from sulfate-based acidic soil-washing effluents.

The analysis of uranium migration is crucial for the accurate safety assessment of high-level radioactive waste (HLW) repository. Previous studies showed that the migration of the uranium can be affected by various physical and chemical processes, such as groundwater flow, heat transfer, sorption/ desorption and, precipitation/dissolution. Therefore, a coupled Thermal-Hydrological-Chemical (THC) model is required to accurately simulate the uranium migration near the HLW repository. In this study, COMSOL-PHREEQC coupled model was used to simulate the uranium migration. In the model, groundwater flow, heat transfer, and non-reactive solute transport were calculated by COMSOL, and geo-chemical reaction was calculated by PHREEQC. Sorption was primarily considered as geo-chemical reaction in the model, using the concept of two-site protolysis nonelctrostatic surface complexation and cation exchange (2 SP NE SC/CE). A modified operator splitting method was used to couple the results of COMSOL and PHREEQC. Three benchmarks were done to assess the accuracy of the model: 1) 1D transport and cation exchange model, 2) cesium transport in the column experiment done by Steefel et al. (2002), and 3) the batch sorption experiment done by Fernandes et al. (2012), and Bradbury and Baeyens (2009). Three benchmark results showed reliable matching with results from the previous studies. After the validation, uranium 1D transport simulation on arbitrary porewater condition was conducted. From the results, the evolution of the uranium front with sequentially saturating sites was observed. Due to the limitation of operator splitting method, time step effect was observed, which caused the uranium to sorbed at further sites then it should. For further study, 3 main tasks were proposed. First, precipitation/ dissolution will be added to the reaction part. Second, multiphase flow will be considered instead of single phase Darcy flow. Last, the effect of redox potential will be considered.

Spent nuclear fuels are temporarily stored in nuclear power plant site. When a problem such as cracking of spent nuclear fuel assembly or cladding occurs or uranium that has not been separated during the reprocessing remains, it is necessary to treat it. The borosilicate glasses have been considered to vitrify whole spent nuclear fuel assembly. However, a large amount of Pb addition was necessary to oxidize metals in assembly to make them suitable for oxide glass vitrifcation. Furthermore, these borosilicate glasses need to be melted at high temperatures (> 1,400°C) when UO2 content is more than 20wt%. Iron phosphate glasses can be melted at a relatively low temperature (< 1,300°C) even with a similar UO2 addition. A composition of iron phosphate glass for immobilization of uranium oxide has been developed. The glasses have glass transition temperatures of ~555°C that are high enough to maintain its phase stability in geological repositories. The waste loading of UO2 in the glass is ~33.73wt%. Normalized elemental releases from the product consistency test were well below the US regulation of 2 g/m2. Nuclear criticality safety and heat generation in deep geological repositories were calculated using MCNP and computational fluid dynamics simulation, respectively. The glass had effective neutron multiplication factor (keff) of 0.755, which is smaller than the nuclear- criticality safety regulation of 0.95. Surface temperature of the disposal canister is expected to lower than the limit temperature (< 100°C). Most of the U in the glass is in the 4+state, which is more chemically durable than the 6+state. As a result of long-term dissolution experiment, chemically-durable uranium pyrophosphate (UP2O7) crystals were formed.

In south Korea, most of uranium deposits are distributed in the Ogcheon belt, which is one of two late Precambrian to Paleozoic fold belts (the Imjingang and Ogcheon belts). A study site of the Ogcheon metamorphic belt (OMB) in Hoenam-myun, Boeun-gun was selected for the natural analogue study by preliminary site investigation for several candidate study sites. Three boreholes were drilled in the site and some rock cores and groundwater samples were taken from the boreholes. Various analytical studies for the samples are now being performed. Thus, in this study, various basic characteristics of the study site such as occurrence, geological, mineralogical, and chemical properties were investigated for a future study. Base rocks containing uranium in the OMB are usually black slate and coaly slate. Coaly slate usually shows a higher content of uranium and larger grain size of uranium than black slate. Uranium minerals found in the OMB are uraninite, uranothorite, brannerite, ekanite, coffinite, francevillite, uranophane, autunite, and torbernite depending on the base rock types. Uranothorite is abundant in black slate whereas uraninite is mostly abundant in coaly slate. Chemical compositions of the solid and groundwater samples from the study site were also analyzed by using ICP-MS/OES (Inductively Coupled Plasma Mass Spectrometry) and XRF (X-ray Fluorescence). This will contribute to determine uranium minerals in the solid samples and uranium speciation in the groundwater. The results of this study will contribute to performing future natural analogue studies in domestic uranium deposits and provide basic information and knowledge for understanding long-term geochemical behaviors of radionuclides in a high-level radioactive repository.

Spent nuclear fuel (SNF) is the main source of high-level radioactive wastes (HLWs), which contains approximately 96% of uranium (U). For the safe disposal of the HLWs, the SNF is packed in canisters of cast iron and copper, and then is emplaced within 500 m of host rock surrounded by compacted bentonite clay buffer for at least 100,000 years. However, in case of the failure of the multi-barrier disposal system, U might be migrated through the rock fractures and groundwater, eventually, it could reach to the biosphere. Since the dissolved U interacts with indigenous bacteria under natural and engineered environments over the long storage periods of geologic disposal, it is important to understand the characteristics of U-microbe interactions under the geochemical conditions. In particular, a few of bacteria, i.e., sulfate-reducing bacteria (SRB), are able to reduce soluble U(VI) into insoluble U(IV) under anaerobic conditions by using their metabolisms, resulting in the immobilization of U. In this study, the aqueous U(VI) removal performance and change in bacterial community in response to the indigenous bacteria were investigated to understand the interactions of U-microbe under anaerobic conditions. Three different indigenous bacteria obtained from different depths of granitic groundwater (S1: 44–60 m, S2: 92–116 m, and S3: 234–244 m) were used for the reduction of U(VI)aq. After the anaerobic reaction of 24 weeks, the changes in bacterial community structure in response to the seeding indigenous bacteria were observed by high-throughput 16S rDNA gene sequencing analysis. The highest uranium removal efficiency of 57.8% was obtained in S3 sample, and followed by S2 (43.1%) and S1 (37.7%). Interestingly, SRB capable of reducing U(VI) greatly increased from 4.8% to 44.1% in S3 sample. Among the SRB identified, Thermodesulfovibrio yellowstonii played a key role on the removal of U(VI)aq. Transmission electron microscopy (TEM) analysis showed that the dspacing of precipitates observed in this study was identical with that of uraninite (UO2). This study presents the potential of U(VI)aq removal by indigenous bacteria under deep geological environment.

Uranium (U) is hazardous material can cause chemical and radiological toxicity, e.g., kidney toxicity and health effects associated with radiation. In Korea, where shallow weathered granitic aquifers are ubiquitous, several previous studies have reported high level of radioactivity in shallow groundwater. This eventually led to the closure of 60 out of 4,140 groundwater production wells in South Korea. Here, we examined aquifers currently dedicated for drinking water supply and investigated the 11,225 dataset of 103 environmental parameters. This dataset includes 80 physical parameters associated with the hydraulic system and 23 chemical parameters associated with waterrock interactions. Among hydraulic parameters, coarse loamy texture in subsoil displayed a notable relation with U concentration level, implied it is controlling the leaching of U from host rocks. Fluorine (F), is one of major products from water-rock interaction in granitic aquifer, exhibited high correlation with U concentration distribution. Positive relation of F concentration with uranium level suggested the dissolved U originated from groundwater interacted with granites. Conclusively, we found that infiltration capacity of soil layers and (2) aqueous speciation in groundwater formulated by interaction of groundwater with local solids, played important role for U concentration in granitic aquifer.

Under the circumstance of energy transition policy of the previous government in which nuclear energy portion will be gradually reduced, some R&D study looking for alternatives other than Pyro- SFR recycling could be very valuable and timely suitable. New alternative study started to evaluate the possibility of it if there are some advantages in terms of waste burden in case that the spent fuel are appropriately treated and disposed of in a disposal site, instead of recycling of spent nuclear fuels (SNF). The alternative study separate the fission products (minor actinides and rare earths) from SNF in a molten salt medium. The molten salt coming from the alternative study is radioactive and heat generating because it contains the fission products chlorides. It is necessary to collect the fission products from the waste molten salt for minimization of the high-level waste volume and to generate a final waste form containing the fission products compatible to the disposal site. Based on the results of a review for various precipitation methods, phosphorylation (phosphate precipitation) of metal chlorides selected as a proper treatment method for recovering of the fission products in a molten salt. Phosphate precipitation has the potential for removing most of fission product elements from a molten salt arising from the treatment of spent nuclear fuel. The performance of phosphate precipitation method evaluated using a salt mixture with the actinide and rare earth chlorides. The molten salt containing uranium as surrogate of the actinides and three rare earths (Nd, Ce, La) chloride was used for testing a phosphate precipitation method at experimental condition (temperature 500°C, salt stirring 200~300 rpm, and 1~1.2 eq. of phosphorylation agent). A cyclic voltammetry (CV) method monitored in-situ phosphate precipitation progress for determining the precipitation rate and conversion ratio evaluated. The phosphorylation reaction increased greatly at a salt stirring 300 rpm.

The damaged spent fuel rods must be stabilized by encapsulation or dry re-fabrication technologies before geological disposal. For applying the dry re-fabrication technology, we manufactured a vertical type furnace to perform both fuel material recovery from damaged fuel rods by oxidative decladding and sinterability improvement of fuel powder by repetition of oxidative and reaction treatment. A horizontal type furnace provides only a diffusion-controlled reaction resulting in longer reaction time and decreasing amount of powder for oxidation and reduction, whereas a vertical type furnace with a submerged gas distributor gives rapid reaction due to flowing gas-solid contact by fluidization. For observation of fluidization behaviors of uranium oxides at room temperature, fluidized column was prepared with transparent cylindrical tube, pressure transmitter and gas flow meter. Number of size of orifice holes was determined by equations in Fluidization Engineering [D.Kunii, O. Levenspiel]. Before uranium oxide test, as surrogates, WO2 (10.8 g/cm3) and Ta2O5 (8.2 g/cm3) powder similar to density of UO2 (10.96 g/cm3) and U3O8 (8.3 g/cm3), respectively were used to achieve fluidization operation conditions in the region from minimum to expanded fluidization. Fluidization behaviors and pressure drop of powder bed was observed according to operation parameters such as gas velocity, number and size of orifice holes, and powder amount.

Interests in molten salt reactor (MSR) using a fast spectrum (FS) have been increased not only for having a high power density but for burning the high-level waste generated from nuclear power plants. For developing the FS-MSR technologies, chloride-based fuels are considered due to the advantage of higher solubility of actinides and lanthanides over fluoride-based salts. Despite significant progress in development of MSR technology, the manufacturing technology for production of the fuel is still insufficiently understood. One of the option to prepare the MSR fuel is to use products from pyroprocessing where oxide form of spent nuclear fuel is reduced into metal form and useful elements can be collected via electrochemical methods in molten salt system at high temperature. In order to chlorinate the products into chloride form, previous study used NH4Cl to chlorinate U metal into UCl3 in an airtight reactor. It was found that the U metal was completely chlorinated into chloride forms; however, impurities generated by the reaction of NH4Cl and reactor wall were found in the product. Therefore, in this work, the air tight reactor was re-deigned to avoid the reaction of reactor wall by insertion of Al2O3 crucible inside of the reactor. In addition, the reactor size was increased to produce UCl3 over 100 g. Using the newly designed reactor, U metal chlorination experiments using NH4Cl chlorinating agent were performed to confirm the optimal experimental conditions. The detailed results will be further discussed.

It has been studied on the disposal area reduction for the used nuclear fuel by the management of high decay-heat nuclides, long-lived nuclides, and highly mobile nuclides. It was investigated on the management of the nuclides in KAERI. Strontium-90 is a high heat-generating nuclide in spent nuclear fuel. It is needed to separate the salt from the salt solution for the recovery of strontium after the chlorination of the strontium oxide in molten salt. Vacuum distillation was used for the separation of strontium from the molten salt. Potassium carbonate was chosen as a reactive distillation reagent for SrCl2 – LiCl – KCl system by the thermodynamic calculation. Reactive distillation experiments were carried out. The residual was mainly SrCO3 in the XRD analysis. It could be concluded that K2CO3 could be one of the suitable reagents for the reactive distillation. The salt in the long–lived nuclide powders should be removed to prepare the block for disposal. Experiments were carried out using W powders (surrogate) and U3O8 powders to develop a process for the removal of the residual salt from UOx powders. The salts were successfully removed from the W and U3O8 powders by distillation.