UO2 kernels, a key component of fuel elements for high temperature gas cooled reactors, have usually been prepared by sol-gel methods. Sol-gel processes have a number of advantages, such as simple processes and facilities, and higher sphericity and density. In this study, to produce 900 μm-sized UO2 particles using an external gelation process, contact length extension of the NH3 gas of the broth droplets pass and the improvement of the gelation device capable of spraying 14 MNH4OH solution are used to form 3,000 μm-sized liquid droplets. To produce high-sphericity and high-density UO2 particles, HMTA, which promotes the gelation reaction in the uranium broth solution, is added to diffuse ammonium ions from the outside of the gelation solution during the aging process and generate ammonium ions from the inside of the ADU gel particles. Sufficient gelation inside of ADU gel particles is achieved, and the density of the UO2 spheres that undergo the subsequent treatment is 10.78 g/cm3; the sphericity is analyzed and found to be 0.948, indicating good experimental results.

One of the promising candidates for accident-tolerant fuel (ATF), a ceramic microcell fuel, which can be distinguished by an unusual cell-like microstructure (UO2 grain cell surrounded by a doped oxide cell wall), is being developed. This study deals with the microstructural observation of the constituent phases and the wetting behaviors of the cell wall materials in three kinds of ceramic microcell UO2 pellets: Si-Ti-O (STO), Si-Cr-O (SCO), and Al-Si-Ti-O (ASTO). The chemical and physical states of the cell wall materials are estimated by HSC Chemistry and confirmed by experiment to be mixtures of Si-O and Ti-O for the STO; Si-O and Cr-O for SCO; and Si-O, Ti-O, and Al-Si-O for the ASTO. From their morphology at triple junctions, UO2 grains appear to be wet by the Si-O or Al-Si-O rather than other oxides, providing a benefit on the capture-ability of the ceramic microcell cell wall. The wetting behavior can be explained by the relationships between the interface energy and the contact angle.

In this study, we investigated the unit process parameters in spherical kernel preparation. Nearly perfect spherical microspheres were obtained from the 0.6M of U-concentration in the broth solution, and the microstructure of the kernel appeared the good results in the calcining, reducing, and sintering processes. For good sphericity, high density, suitable microstructure, and no-crack final microspheres, the temperature control range in calcination process was , and the microstructure, the pore structure, and the density of kernel could be controlled in this temperature range. Also, the concentration changes of the ageing solution in aging step were not effective factor in the gelation of the liquid droplets, but the temperature change of the ageing solution was very sensitive for the final ADU gel particles

The effect of a CAM (Continuous Attrition Mill) treatment on the sinterability of ex-ADU powder was investigated. As the cycles of a CAM increased, the apparent density, specific surface area and O/U of the milled powder increased, but there particle sizes decreased. However the sintered density of the pellet decreased as the cycles of the CAM increased. It is considered that the decrease of the sintered density is due to the formation of , which was produced by a CAM mechanism

농축 우라늄 UF6로 건식법 중의 하나인 DCP법으로 핵연료 UO2+x 분말을 제조하였다. Rotary kiln 내로 수증기를 주입할 때 일어나는 온도 변화에 따른 UO2+x 분말 특성을 우라늄 분석기, 수분 측정기, SEM 등으로 측정하였다. 그 결과 불소의 함유량은 8ppm을 나타냈고, 수분 함량의 경우 최적화되었음을 알 수 있었다.

UO2분말에 PVA-Al(III) 착물과 AlOOH를 각각 0.03~1.0wt%를 첨가하여 소결체를 제조한 후 소결체 특성을 비교하였다. PVA-Al(III) 착물과 AlOOH는 1000˚C의 소소분위기에서 열분해하는 경우 생성상은 θ-Al2O3이었다. 순수 UO2분말에 비해 AlOOH가 첨가된 혼합 분말의 곁보기 밀도는 더 높았고, PVA-Al(III) 착물이 첨가된 혼합 분말은 더 낮은 겉보기 밀도를 보였다. AlOOH가 첨가된 소결체의 경우 약 800˚C 부근에서 치밀화가 시작되었지만, PVA-Al(III) 착물이 첨가된 소결체의 경우에는 약 900˚C에서 치밀화가 시작되었다. 기공 크기 분포는 AlOOH가 첨가된 UO2소결체의 경우에 monomodal 형태로, 그리고 PVA-Al(III) 착물의 첨가된 소결체의 경우에는 bimodal 형태로 나타났다. 결정립 크기는 1wt% 첨가시 AlOOH가 첨가된 UO2소결체의 경우에 약 13μm이었지만, PVA-Al(III) 착물의 첨가된 소결체의 경우에는 약 36mum까지 성장하는 현저한 효과를 확인하였다.

UO2-6wt%Gd2O3가연성 독물질 소결체에 미량첨가한 Al 화합물(Al(OH)3, ADS(aluminium disterate), Al(OH)3+ADS)이 소결성 및 미세조직에 미치는 영향을 고찰하고자 하였다. 이를 위하여 Al이 첨가된 UO2-6wt%Gd2O3압분체를 1700˚C, 수소 분위기에서 4시간동안 소결한 후 특성시험을 수행하였다 Al을 첨가한 UO2-6wt%Gd2O3의 소결밀도는 94% T.D.이상이였고, ADS를 이용한 Al 첨가가 개기공도 감소에 상대적으로 크게 기여하였다. 또한 Al을 첨가하면 10μm 이상의 큰 기공과 1μm 이하의 작은 기공은 많이 줄어들었고 첨가된 Al 화합물의 종류와는 무관하게 평균 기공크기는 2-3μm였다. 그리고 Al을 첨가하지 않은 소결체의 결정립은 이중 결정립 형태를 갖는 반면에 Al을 첨가하면 결정립은 균일하였다. 특히, ADS를 첨가한 소결체의 평균 결정립 크기는 4.6μm로 가장크게 증가하였다.

윤활제인 zinc stearate의 첨가량 (0~4wt%)을 변화시켜 UO2분말의 겉보기밀도, 성형시의 분말입자간 마찰과 입자/다이벽 마찰간의 상관관계를 조사하였다. 소량의 윤활제 첨가시에는 UO2분말입자간 박막의 윤활제 도포층이 형성되어 겉보기밀도가 증가한 반면 다량의 윤활제를 첨가한 경우에는 UO2 분말입자에 두꺼운 윤활제 도포층이 형성되고 미혼합된 윤활제가 존재하여 겉보기밀도는 감소하는 경향을 보였다. 윤활제를 첨가혼합한 상태에서 다이벽 윤활도포 유무에 따라 구한 UO2 성형체의 성형압력/성형밀도 자료로부터 분말입자간 마찰, 입자/다이벽 마찰 그리고 성형시 lubrication/inhibition등의 상대적 중요성을 조사하였다. 입자/다이벽 마찰에 의한 압력손실은 입자간 마찰에 의한 압력손실보다 크게 나타났다. 입자/다이벽 마찰에 의한 압력손실은 다이벽 윤활제 도포에 의해 최소화될 수 있지만 상대적으로 바람직하지 않은 성형시의 inhibition이 야기되는 것으로 나타났다.

순수 UO2에 첨가량 변화 및 ball-milling 시간에 따른 (U, Ce)O2 분말의 특성과 각 조건별로 제조된 분말을 압분 및 소결하여 (U, Ce)O2 분말 특성에 따른 소결성을 비교 조사하였다. 실험 결과로 부터 ball-milling시간이 길어짐에 따라 입자들은 미세화되고, CeO2 함량이 증가할수록 압분, 소결밀도는 저하 하였으며, CeO2는 소결성을 저하시키는 산화물임을 확인하였다. 10wt%,CeO2 가 첨가된 (U, Ce)O2 분말의 경우, ball-milling 4시간 수행한 분말의 소결체가 기공의 수도 적고, 구형에 가까왔으며, 소결밀도가 가장 높았다. 이는 4시간 ball-milling한 (U, Ce)O2분말이 비표면적이 크로 그의 packing ratio가 적절하였기 때문이다.

Pyroprocessing technology has emerged as a viable alternative for the treatment of metal/oxide used fuel within the nuclear fuel cycle. This innovative approach involves an oxide reduction process wherein spent fuel in oxide form is placed within a cathode basket immersed in a molten LiCl-Li2O salt operating at 923 K. The chemical reduction of these oxide materials into their metallic counterparts occurs through a reaction with Li metal, which is electrochemically deposited onto the cathode. However, during process, the generation of Li2O within the fuel basket is inevitable, and due to the limited reduction efficiency, a significant portion of rare earth oxides (REOx) remains in their oxide state. The presence of these impurities, specifically Li2O and REOx, necessitates their transfer into the electrorefining system, leading to several challenges. Both Li2O and REOx exhibit reactivity with UCl3, the primary electrolyte within the electrorefining system, causing a continuous reduction in UCl3 concentration throughout the process. Furthermore, the formation of fine UO2 powder within the salt system, resulting from chemical reactions, poses a potential long-term operational and safety concern within the electrorefining process.Various techniques have been developed to address the issue of UO2 fine particle removal from the salt, utilizing both chemical and mechanical methods. However, it is crucial that these methods do not interfere with the core pyroprocessing procedure. This study aims to investigate the impact of Li2O and REOx introduced from the electrolytic reduction process on the electrorefining system. Additionally, we propose a method to effectively eliminate the generated UO2 fine powder, thereby enhancing the long-term operational stability of the electrorefining process. The efficiency of this proposed solution in removing oxidized powder has been confirmed through laboratory-scale testing, and we will provide a comprehensive discussion of the detailed results.

To improve the safety of nuclear fuel, research on the advanced nuclear fuel (UO2) by adding various trace elements is being conducted. For example, the addition of metals such as Mo, Cr can improve the thermal conductivity of nuclear fuel, minimizing the diffusion of fission products. Trace metal oxide additives (SiO2, Cr2O3, Al2O3, etc.) can suppress the release of fission gases. In general, complete dissolution of the fuel sample is required for chemical analysis to determine its elemental compositions. Among widely used metal oxide additives, aluminum oxide is difficult to dissolve in nitric acid due to its excellent thermal and chemical stability. In this study, we investigated on different chemical dissolution methods by applying a microwave digestion system under various acid solutions. We confirmed the validity of the digestion method by carrying out trace element analysis using an Inductively-Coupled Plasma Atomic Emission Spectrometer (ICP-AES).

The stabilization techniques are highly required for damaged nuclear fuel to strengthen safety in terms of transportation, storage, and disposal. This technique includes recovering fuel materials from spent fuel, fabrication of stabilized pellets, and fabrication of fuel rods. Thus, it is important to identify the leaching behavior of the stabilized pellets to verify their stability in humid environments which are similar to storage conditions. In this study, we introduce various leaching experiment methods to evaluate the leaching behavior of the stabilized pellets, and determine the most suitable leaching test methods for the pellets. Also, we establish the leaching test conditions with various factors that can affect the dissolution and leaching behavior of the stabilized pellets. Accordingly, we prepare the simulated high- (55 GWd/tU) and low- (35 GWd/tU) burnup nuclear fuel (SIMFUEL) and pure UO2 pellets sintered at 1,550°C and 1,700°C, respectively. Each pellet is placed in a vessel and filled with DI water and perform the leaching test at three different temperature to verify the leaching mechanism at different temperature range. Based on the standard leaching test method (ASTM C1308-21), the test solution is removed from the pellet after specific time intervals and replaced in the fresh water, and the vessel is placed back into the controlled-temperature ovens. The test solutions are analyzed by using ICP-MS.

Irradiated uranium dioxide in damaged used fuel could oxidize during transportation, interim storage or disposal, resulting that the fuel pellet fragments are reduced to a grain-sized powder that can easily escaped from the damaged rod. It has been reported that oxidized spent fuel (i.e. U4O9+x) that was in contact with water could increase the dissolution rate by making the grain boundaries more accessible to the water. Therefore, the damaged used fuel requires stabilization technology including nuclear material recovery, pellet manufacturing process, and stabilization fuel rod manufacturing that can secure safety in terms of permanent disposal. In this study, we prepared pure UO2 and SIMFUEL pellets that are a mixture of UO2 and surrogated metallic oxides for fission products equivalent to a burn-up of 35 GWd/tU and 55 GWd/tU as the stabilized spent fuel. The UO2 and fission products powders were milled and pressed into pellets at 250 MPa and sintered at 1,550°C and 1,700°C for 6 hours in an atmosphere of 4%H2-Ar. The prepared UO2 and SIMFUEL pellets were placed in PTFE Teflon vessels and filled with deionized water to identify the leaching behavior by a long-term leaching experiment under the similar condition to a repository for the safe disposal.

Accurate understanding of structural integrity and chemical reactivity of UO2 disposed in deep underground sites is of importance. Owing to the specific condition of the site location, UO2 may have substantially different properties from the conventional prediction. In this study, we demonstrate that the oxidation resistivity of UO2 is considerably modified by gadolinium (Gd), which is the element of neutron absorber and a byproduct of nuclear decay of radioactive U-235. Using density functional theory calculations, we investigate how the oxidation mechanism of UO2 changes with Gd incorporation in U lattice. Our study indicates that Gd remarkably enhances the thermodynamic stability of pristine UO2 against surface oxidation via three underlying mechanisms: (i) weakens the chemical bonding of adsorbed oxygen atom (O) with U, (ii) reduces active sites (U) for oxygen adsorption, and (iii) suppresses the subsurface diffusion of adsorbed O delaying the growth of the oxide layers on the UO2. Electronic and lattice structure analyses for Gd-doped UO2 indicate that amount of charge transfer from U to O is critically reduced and the lattice of the UO2 surface is contracted. Our results provide useful information for understanding long-term stability and improving the structural integrity of UO2 through the chemical doping process.