The oxidative dissolution mechanism of uranium dioxide. I. The effect of temperature in hydrogen carbonate medium (original) (raw)
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Journal of Nuclear Materials, 2005
The objective of this work is to study the UO 2 oxidation by O 2 and dissolution in bicarbonate media and to extrapolate the results obtained to improve the knowledge of the oxidative dissolution of spent nuclear fuel. The results obtained show that in the studied range the oxygen consumption rate is independent on the bicarbonate concentration while the UO 2 dissolution rate does depend on. Besides, at 10 À4 mol dm À3 bicarbonate concentration, the oxygen consumption rate is almost two orders of magnitude higher than the UO 2 dissolution rate. These results suggest that at low bicarbonate concentration (<10 À2 mol dm À3 ) the alteration of the spent nuclear fuel cannot be directly derived from the measured uranium concentrations in solution. On the other hand, the study at low bicarbonate concentrations of the evolution of the UO 2 surface at nanometric scale by means of the SFM technique shows that the difference between oxidation and dissolution rates is not due to the precipitation of a secondary solid phase on UO 2 .
Physical and Chemical Aspects of Radiation Induced Oxidative Dissolution of UO2
2006
The general subject of this thesis is oxidative dissolution of UO 2. The dissolution of UO 2 is mainly investigated because of the importance of the UO 2 matrix of spent nuclear fuel as a barrier against radionuclide release in a future deep repository. U(IV) is extremely insoluble under the reducing conditions prevalent in a deep repository, whereas U(VI) is more soluble. Hence, oxidation of the UO 2-matrix will affect its solubility and thereby its function as a barrier. In this thesis the relative efficiency of one-and two electron oxidants in dissolving UO 2 is studied. The oxidative dissolution yield of UO 2 was found to differ between one-and two-electron oxidants. At low oxidant concentrations the dissolution yields for one-electron oxidants are significantly lower than for two-electron oxidants. However, the dissolution yield for one-electron oxidants increases with increasing oxidant concentration, which could be rationalized by the increased probability for two consecutive one-electron oxidations at the same site and the increased possibility for disproportionation. This licentiate thesis is based on the following publications:
Effect of external gamma irradiation on dissolution of the spent UO2 fuel matrix
Journal of Nuclear Materials, 2005
Leaching experiments were performed on UO 2 pellets doped with alpha-emitters (238/239 Pu) and on spent fuel, in the presence of an external gamma irradiation source (A 60 Co = 260 Ci, _ Dc ¼ 650 Gy h À1). The effects of a, b, c radiation, the fuel chemistry and the nature of the cover gas (aerated or Ar + 4%H 2) on water radiolysis and on oxidizing dissolution of the UO 2 matrix are quantified and discussed. For the doped UO 2 pellets, the nature of the cover gas clearly has a major role in the effect of gamma radiolysis. The uranium dissolution rate in an aerated medium is 83 mg m À2 d À1 compared with only 6 mg m À2 d À1 in Ar + 4%H 2. The rate drop is accompanied by a reduction of about four orders of magnitude in the hydrogen peroxide concentrations in the homogeneous solution. The uranium dissolution rates also underestimate the matrix alteration rate because of major precipitation phenomena at the UO 2 pellet surface. The presence of studtite in particular was demonstrated in aerated media; this is consistent with the measured H 2 O 2 concentrations (1.2 • 10 À4 mol L À1). For spent fuel, the presence of fission products (Cs and Sr), matrix alteration tracers, allowed us to determine the alteration rates under external gamma irradiation. The fission product release rates were higher by a factor of 5-10 than those of the actinides (80-90% of the actinides precipitated on the surface of the fragments) and also depended to a large extent on the nature of the cover gas. No significant effect of the fuel chemistry compared with UO 2 was observed on uranium dissolution and H 2 O 2 production in the presence of the 60 Co source in aerated conditions. Conversely, in Ar + 4%H 2 the fuel self-irradiation field cannot be disregarded since the H 2 O 2 concentrations drop by only three orders of magnitude compared with UO 2 .
Journal of Nuclear Materials, 2005
The effects of alpha dose-rate on UO 2 dissolution were investigated by performing dissolution experiments with 238 Pu-doped UO 2 materials containing nominal alpha-activity levels of $1-100 Ci/kg UO 2 (actual levels 0.4-80 Ci/kg UO 2 ), in 0.1 M NaClO 4 and in 0.1 M NaClO 4 + 0.1 M carbonate. Dissolution rates increased less than 10-fold for an almost 100-fold increase in doping level and fall within the range of predictions of the Mixed Potential Model (a detailed mechanistic model for used fuel dissolution). Dissolution rates were lower in carbonate-free solutions and enrichment of 238 Pu on the UO 2 surface was suggested in carbonate solutions. Effective G values, defined as the ratio of the total amount of U dissolved divided by the maximum possible amount of U dissolved by radiolytically produced H 2 O 2 , increased with decreasing doping levels. This suggests that the dissolution reaction at high dose rates is limited by the reaction rate between UO 2 and H 2 O 2 , but becomes increasingly limited by the rate of production of H 2 O 2 at lower dose rates.
The effect of ion irradiation on the dissolution of UO2 and UO2-based simulant fuel
Journal of Alloys and Compounds, 2018
The aim of this work was to study the separate effect of fission fragment damage on the dissolution of simulant UK advanced gas-cooled reactor nuclear fuel in water. Plain UO 2 and UO 2 samples, doped with inactive fission products to simulate 43 GWd/tU of burn-up, were fabricated. A set of these samples were then irradiated with 92 MeV 129 Xe 23þ ions to a fluence of 4.8 Â 10 15 ions/cm 2 to simulate the fission damage that occurs within nuclear fuels. The primary effect of the irradiation on the UO 2 samples, observed by scanning electron microscopy, was to induce a smoothening of the surface features and formation of hollow blisters, which was attributed to multiple overlap of ion tracks. Dissolution experiments were conducted in single-pass flow-through (SPFT) mode under anoxic conditions (<0.1 O 2 ppm in Ar) to study the effect of the induced irradiation damage on the dissolution of the UO 2 matrix with data collection capturing six minute intervals for several hours. These time-resolved data showed that the irradiated samples showed a higher initial release of uranium than unirradiated samples, but that the uranium concentrations converged towards~10 À9 mol/l after a few hours. Apart from the initial spike in uranium concentration, attributed to irradiation induced surficial micro-structural changes, no noticeable difference in uranium chemistry as measured by X-ray electron spectroscopy or 'effective solubility' was observed between the irradiated, doped and undoped samples in this work. Some secondary phase formation was observed on the surface of UO 2 samples after the dissolution experiment.
Journal of Nuclear Materials, 2012
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On the Stability of Uranium Carbide in Aqueous SolutionEffects of HCO3 and H2O2
Uranium carbide (UC) is a candidate fuel material for future Generation IV nuclear reactors. As part of a general safety assessment, it is important to understand how fuel materials behave in aqueous systems in the event of accidents or upon complete barrier failure in a geological repository for spent nuclear fuel. As irradiated nuclear fuel is radioactive, it is important to consider radiolysis of water as a process where strongly oxidizing species can be produced. These species may display high reactivity toward the fuel itself and thereby influence its integrity. The most important radiolytic oxidant under repository conditions has been shown to be H 2 O 2. In this work, we have studied the dissolution of uranium upon exposure of UC powder to aqueous solutions containing HCO 3 − and H 2 O 2 , separately and in combination. The experiments show that UC dissolves quite readily in aqueous solution containing 10 mM HCO 3 − and that the presence of H 2 O 2 increases the dissolution further. UC also dissolves in pure water after the addition of H 2 O 2 , but more slowly than in solutions containing both HCO 3 − and H 2 O 2. The experimental results are discussed in view of possible mechanisms.
The oxidative dissolution of unirradiated UO2 by hydrogen peroxide as a function of pH
Journal of Nuclear Materials, 2005
The dissolution of non-irradiated UO 2 was studied as a function of both pH and hydrogen peroxide concentration (simulating radiolytic generated product). At acidic pH and a relatively low hydrogen peroxide concentration (10 À5 mol dm À3 ), the UO 2 dissolution rate decreases linearly with pH while at alkaline pH the dissolution rate increases linearly with pH. At higher H 2 O 2 concentrations (10 À3 mol dm À3 ) the dissolution rates are lower than the ones at 10 À5 mol dm À3 H 2 O 2 , which has been attributed to the precipitation at these conditions of studtite (UO 4 AE 4H 2 O, which was identified by X-ray diffraction), together with the possibility of hydrogen peroxide decomposition. In the literature, spent fuel dissolution rates determined in the absence of carbonate fall in the H 2 O 2 concentration range 5 · 10 À7 -5 · 10 À5 mol dm À3 according to our results, which is in agreement with H 2 O 2 concentrations determined in spent fuel leaching experiments.