The effects of diffusive transport limitations on the dissolution of UO(2) were investigated using an artificial groundwater prepared to simulate the conditions at the Old Rifle aquifer site in Colorado, USA. Controlled batch, continuously-stirred tank (CSTR), and plug flow reactors were used to study UO(2) dissolution in the absence and presence of diffusive limitations exerted by permeable sample cells. The net rate of uranium release following oxidative UO(2) dissolution obtained from diffusion-limited batch experiments was ten times lower than that obtained for UO(2) dissolution with no permeable sample cells. The release rate of uranium to bulk solution from UO(2) contained in permeable sample cells under advective flow conditions was more than 100 times lower than that obtained from CSTR experiments without diffusive limitations. A 1-dimensional transport model was developed that could successfully simulate diffusion-limited release of U following oxidative UO(2) dissolution with the dominant rate-limiting process being the transport of U(VI) out of the cells. Scanning electron microscopy, X-ray diffraction, and extended X-ray absorption fine structure spectroscopy (EXAFS) characterization of the UO(2) solids recovered from batch experiments suggest that oxidative dissolution was more evident in the absence of diffusive limitations. Ca-EXAFS spectra indicate the presence of Ca in the reacted UO(2) solids with a coordination environment similar to that of a Ca-O-Si mineral. The findings from this study advance our overall understanding of the coupling of geochemical and transport processes that can lead to differences in dissolution rates measured in the field and in laboratory experiments.
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http://dx.doi.org/10.1016/j.watres.2012.08.034 | DOI Listing |
Sci Rep
June 2024
Fukushima Daiichi Decontamination and Decommissioning Engineering Company, Tokyo Electric Power Company Holdings Inc., Fukushima, 979-1301, Japan.
Particles containing alpha (α) nuclides were identified from sediment in stagnant water in the Unit 3 reactor building of the Fukushima Daiichi Nuclear Power Station (FDiNPS). We analyzed different concentrations of α-nuclide samples collected at two sampling sites, the torus room and the main steam isolation valve (MSIV) room. The solids in the stagnant water samples were classified, and the uranium (U) and total alpha concentrations of each fraction were measured by dissolution followed by inductively coupled plasma mass spectrometry and α-spectrometry.
View Article and Find Full Text PDFHO produced from water radiolysis is expected to play a significant role in radiation induced oxidative dissolution of spent nuclear fuel under the anoxic conditions of a deep geological repository if the safety-barriers fail and ground water reaches the fuel. It was recently found that the coordination chemistry between U(vi), HCO and HO can significantly suppress HO induced dissolution of UO in 10 mM bicarbonate. This was attributed to the much lower reactivity of the U(vi)O-coordinated O as compared to free HO.
View Article and Find Full Text PDFInorg Chem
April 2024
Radiochemistry Division, Bhabha Atomic Research Centre, Mumbai 400085, India.
A quantitative, rapid, endothermic dissolution of UO in Cmim·PF (1-alkyl-3-methyl imidazolium hexafluorophosphate) has been achieved within 2 h at 65 °C by in situ generated fluoride ions by pre-equilibrating the ionic liquid with suitable concentrations of nitric acid. The efficiency of the dissolution followed the trend: UO > UO > UO. The fluoride generation was found to increase with the concentration of nitric acid being equilibrated, the water content of the ionic liquid, and also the time of equilibration.
View Article and Find Full Text PDFRSC Adv
January 2024
Shanghai Institute of Applied Physics, Chinese Academy of Sciences Shanghai 201800 China +86-21-39194027.
Oxides are one of the most important impurities in the fuel salt of molten salt reactors (MSRs), and excessive oxide impurities pose a risk to the safe operation of MSRs. This study focused on investigating the precipitation behavior between Th, U, and Be with O in the 2LiF-BeF (FLiBe) eutectic salt system. The results showed that the solubility of UO was 5.
View Article and Find Full Text PDFRSC Adv
September 2023
Nuclear Science and Engineering Center, Japan Atomic Energy Agency (JAEA) Tokai Ibaraki 319-1195 Japan
Upon nuclear waste canister failure and contact of spent nuclear fuel with groundwater, the UO matrix of spent fuel will interact with oxidants in the groundwater generated by water radiolysis. Bicarbonate (HCO) is often found in groundwater, and the HO induced oxidative dissolution of UO in bicarbonate solution has previously been studied under various conditions. Temperatures in the repository at the time of canister failure will differ depending on the location, yet the effect of temperature on oxidative dissolution is unknown.
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