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.

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.

사고저항성 핵연료의 일환으로 UO2 입자가 세라믹 셀 벽으로 둘러싸인 미세구조를 갖는 세라믹 미소셀 UO2 소결체를 개발 중이다. 이는 핵분열생성물들을 UO2 펠렛 내에 포집하여 펠렛 외부로의 방출을 저감함으로써 봉내압 상승을 완화하고 응력부식균열 발생률을 낮춘다. 생성량이나 방사능 측면에서 위험한 핵분열생성물 중 하나로 여겨지는 세슘은 세라믹 미소셀 소결체 내에서 셀 물질과 화학반응 하여 포집될 수 있다. 따라서, 세슘 포집능은 해당 화학반응의 열역학적, 속도론적 특성에 의해 결정된다. 역으로, 미소셀 소결체의 조성설계 시 해당 반응에 대한 열역학적 예측이 필수적이다. 본 연구는 세라믹 현재 개발 중인 여러 미소셀 조성(Si-Ti-O, Si-Cr-O, Si-Al-O)에 대해 세슘의 포집능을 평가하는 열역학적 계산을 다룬다. 계 산에 앞서 먼저 HSC Chemistry를 이용해 세슘과 셀 물질의 물리/화학적 상태를 정의한 후, LWR 정상운전 모사환경에서 계 산된 세슘포텐셜(ΔGCs)과 산소포텐셜(ΔG02)에 근거하여 세슘포집 반응성을 평가하였다. 계산 결과에 근거하면, 세슘 포집 반응은 상기 모든 조성에서 자발적일 것으로 예상되며 이로써 조성설계의 근거를 제시함과 동시에 세슘의 포집능을 평가하는 효과적인 방법을 제공한다.