In all geodisposal scenarios it is key to understand the interaction of radionuclides with mineral particles during their formation/recrystallisation. Studying processes at the molecular scale provides insight into long-term radionuclide behaviour. Uranium is a significant radionuclide in higher activity wastes destined for geological disposal, and iron (oxyhydr) oxides (e.g. goethite, 􀟙-FeOOH). are ubiquitous in and around these systems, formed via processes including metal corrosion and microbially induced reactions. There are numerous reports of uranium-incorporation into iron (oxyhydr) oxides, therefore it has been suggested that they may be a barrier to uranium migration in geodisposal systems. However, long-term stability of these phases during environmental perturbations are unexplored. Specifically, U-incorporated iron (oxyhydr) oxide phases may interact with Fe(II) and sulphide from biological or geological origin. Firstly, electron transfer occurs between adsorbed Fe(II) and iron oxyhydroxides, with potential for changes in the speciation of incorporated uranium e.g. oxidation state changes and/or release. Secondly, on exposure to aqueous sulfide, iron (oxyhydr) oxides undergo reductive dissolution and recrystallisation to iron sulphides. Understanding the fate of incorporated uranium during these process in key to understanding its long term behaviour in subsurface systems. A series of experimental studies were undertaken where U(VI)-goethite was synthesized then reacted with either aqueous Fe(II) or S(-II), and the system monitored over time using geochemical analysis and X-ray absorption spectroscopy (XAS) techniques e.g. U LIII-edge and MIV-edge HERFD-XANES. Reaction with aqueous Fe(II) resulted in electron transfer between Fe(II) and U(VI)-goethite, with > 50% U(VI) reduced to U(V). XAS analysis revealed that U remained within the goethite structure, and electron transfer only occurred within the outermost atomic layers of goethite. which led to U reduction. Rapid reductive dissolution of U(VI)-goethite occurred on reaction with sulfide at pH7. A transient release of aqueous U was observed during the first day, likely due to uranyl(VI)-persulfide species. However, U was retained in the solid phase in the longer term. In contrast, the sulfidation of U adsorbed to ferrihydrite at pH 12.2 led to the immediate release of U (< 10% Utotal) associated with a colloidal erdite (NaFeS2·2H2O) phase. Moreover, in the bulk phase the surface of ferrihydrite was passivated by sulfide, and U was found to have been trapped within surface associated erdite-like fibres. Overall, these studies further understanding of the long-term behaviour of U-incorporated iron (oxyhydr)oxides supporting the overarching concept of iron (oxyhydr) oxides acting as a barrier to U migration.

Currently, the Korea Atomic Energy Research Institute is conducting research on the development of technology to reduce the disposal area for SF (Spent nuclear Fuel). If the main radionuclides contained in SF can be separated and recovered according to their characteristics (long half-life, high mobility and high heat load) and uranium oxide which is expected to be the final residue, can be made into solids, the burden of the permanent disposal area of the SF will be greatly reduced. The waste form that end up in the repository must be verified for ease of manufacture and stability of the block. And, in order to increase the loading efficiency, a large block manufacturing technology is needed. This study describes the background of introducing PSA (Particle Size Analyzer) which is one of the necessary equipment for manufacturing UO2 blocks using slip casting, the method of using the equipment and performance verification of the equipment using standard samples. The particle size affects the sintering quality by the way the particles rearrange themselves during sintering. Powders of small particles are generally less free flowing and more difficult to compress, they form thin pores between the particles and sinter to higher density. In contrast, larger particle has a lower sintered density. Therefore, accurate particle size measurement and the selection of a suitable particle size are important. For this purpose, a PSA was installed in nuclear cycle experiment research center. To verify the performance of the equipment, a standard sample of 1.025 μm was analyzed. We got an average particle size of 1.0293 μm and standard deviation of 0.0668 μm. This value was within the uncertainty(±0.018 μm) of the sample’s certificate. In the future, this equipment will measure the size of UO2 (depleted uranium) powder and to produce large scale uranium oxide blocks.

Confirmation of the thermal behavior of spent fuel is one of the important points in the management of high-level radioactive waste. This is because various fission products exist in spent nuclear fuel, and a management plan according to their behavior is required. Among the fission products, epsilon particles exist in the form of metal deposits and have a great influence on their physical and chemical properties. However, observing the thermal behavior of epsilon particles is important for understanding spent fuel behavior in thermally environment, but it is difficult to maintain a consistent thermal environment. In this work, we report the thermal behaviors study of uranium oxide with epsilon particle using in situ high temperature X-ray diffraction. We measured the variation of temperature on the size of crystalline, which is a cell parameter in the reaction process. And then, the change of lattice parameters is calculated by Rietveld refinement.

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.

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.

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.

파이로 공정에서는 사용후핵연료 관리 공정 개발의 일환으로 산화 우라늄을 고온 용융염 전해질계에서 전기화학적 방법으로 환원시키기 위한 전해환원 공정이 개발되고 있다. 이에 따른 전해환원 공정의 반응기 설계를 위해서는 전기화학적 이론에 기초한 모델이 요구되고 있다. 본 연구에서는 상 분리를 설명하는 phase-field 이론에 기초하여 우라늄 산화물의 전해환원 모사를 위한 1차원 모델이 개발되었다. 모델은 우라늄 산화물 내 산소 원소의 확산과 펠렛 표면에서 전기화학 반응 속도를 나타내는 매개변수를 사용하여 외부로부터 내부로 진행되는 전해환원을 잘 모사하고 있으며 계산 결과 전체 전류는 산소 원소의 내부 확산에 크게 의존하는 것으로 나타났다. 전해환원 반응에 대한 모델은 대용량 장치 설계에 최적화된 조건 도출에 활용될 것으로 예상되며 장치 설계가 완료되면 공정 연계 모사에 직접 사용될 것으로 기대된다.

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 effects of thermal treatment conditions on ADU (ammonium diuranate) prepared by SOL-GEL method, so-called GSP (Gel supported precipitation) process, were investigated for kernel preparation. In this study, ADU compound particles were calcined to particles in air and Ar atmospheres, and these particles were reduced and sintered in 4%-/Ar. During the thermal calcining treatment in air, ADU compound was slightly decomposed, and then converted to phases at . At , the phase appeared together with . After sintering of theses particles, the uranium oxide phases were reduced to a stoichiometric . As a result of the calcining treatment in Ar, more reduced-form of uranium oxide was observed than that treated in air atmosphere by XRD analysis. The final phases of these particles were estimated as a mixture of and .

650의 LiCl-LiO 용융염계에서 10 g UO/batch 규모의 장치를 이용해서 우라늄산화물의 전해환원 특성에 대한 평가를 수행하였다. 일체형 음극은 고체전극, 우라늄산화물과 우라늄산화물을 담아주는 다공성 용기(멤브레인)로 구성된다. 멤브레인 재료로는 325-mesh 스테인레스강막과 다공성 마그네시아 도가니를 사용하였다. 일체형 음극의 재질에 따른 LiCl-3 wt LiO계와 UO-LiCl-3 wt LiO계의 순환 전압측정법 결과로부터 전해환원 반웅 메커니즘을 규명하였다. 일체형 음극의 재질에 따른 우라늄산화물의 직접 및 간접 전해환원에 대한 실험을 수행하였다. 그 결과, 325-mesh스테인레스강막을 사용하여 직접 및 간접 전해환원으로 금속전환을 수행하였을 때 낮은 전류효율로 인해 우라늄산화물을 금속우라늄으로 환원시키지 못했으며, 마그네시아 다공성 도가니를 사용하여 간접 전해환원으로 금속전환을 수행하였을 때는 높은 전류효율로 인해 우라늄산화물을 금속우라늄으로 환원시킬 수 있었다

본 연구에서는 고온의 LiCl-LlO 용융염계에서 우라늄 산화물의 금속전환과 LiO의 전해반응이 동시에 진행되는 통합 반응 메카니즘을 기초로 한 전기화학적 금속전환기술을 제안하였다. 본 실험에서는 전기화학적 환원반응에 의해 생성된 Li 금속이온이 음극에 전착과 동시에 우라늄 산화물과 반응하여 금속전환율 99 % 이상의 우라늄 감속을 생성하는 통합 반응 메카니즘을 확인할 수 있었다. 또한 전기화학적 금속전환기술의 공정 적용성 평가 일환으로 우라늄 산화물의 금속전환성, 반응 메카니즘 규명, LiO의 closed recycle rate 및 물질전달 특성 등의 기초 데이터를 확보하였다 향후 전기화학적 금속전환기술은 LiCl-Li 용융염계의 금속전환공정의 반응조건 제한성 해소, 금속전환율 향상 및 공정의 단순화 등의 기술성과 경제성 향상 측면에서 획기적인 방안으로 고려될 수 있을 것으로 판단된다.

The nano-scale crystallite sizes of uranium oxide powders in simulated spent fuel were measured by the neutron diffraction line broadening method in order to analyze the sintering behavior of the dry process fuel. The mixed and fission product powders were dry-milled in an attritor for 30, 60, and 120 min. The diffraction patterns of the powders were obtained by using the high resolution powder diffractometer in the HANARO research reactor. Diffraction line broadening due to crystallite size was measured using various techniques such as the Stokes' deconvolution, profile fitting methods using Cauchy function, Gaussian function, and Voigt function, and the Warren-Averbach method. The non-uniform strain, stacking fault and twin probability were measured using the information from the diffraction pattern. The realistic crystallite size could be obtained after separation of the contribution from the non-uniform strain, stacking fault and twin.