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ABSTRACT: The so-called Instant Release Fraction (IRF) is considered to govern the dose released from Spent Fuel repositories. Often, IRF calculations are based on estimations of fractions of inventory release based in fission gas release [1]. The IRF definition includes the inventory located within the Gap although a conservative approach also includes both the Grain Boundary (GB) and the pores of restructured HBS inventories.A correction factor to estimate the fraction of Grain Boundary accessible for leaching has been determined and applied to spent fuel static leaching experiments carried out in the ITU Hot Cell facilities [2]. Experimental work focuses especially on the different properties of both the external rim area (containing the High Burn-up Structure (HBS)) and the internal area, to which we will refer as Out and Core sample, respectively. Maximal release will correspond to an extrapolation to simulate that all grain boundaries or pores are open and in contact with solution.The correction factor has been determined from SEM studies taking into account the number of particles with HBS in Out sample, the porosity of HBS particles, and the amount of transgranular fractures during sample preparation.
MRS Proceedings. 12/2007; 1107.
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ABSTRACT: Due to the higher stability of the spent nuclear fuel (mainly composed of UO2) under reducing conditions, and in order to enhance the retention/retardation of some key radionuclides, the olivine rock from the Lovasjärvi intrusion has been proposed as a potential redox-active backfill-additive in deep high-level nuclear waste (HLNW) repositories. In this work, two different approaches have been undertaken in order to establish the redox buffer capacity of olivine rock: (1) The capacity of the rock to respond to changes in pH or pe has been demonstrated and the final (pH, pe) coordinates agree with the control exerted by the system Fe(II)/Fe(III). (2) The rate of consumption of oxygen has been determined at different pH values. These rates are higher than the ones reported in the literature for other solids, what would point to the possibility of using this rock as an additive to the backfill material in a HLNW.
Journal of Contaminant Hydrology 03/2006; 83(1-2):42-52. · 2.32 Impact Factor
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J. Quiñones,
E. Iglesias,
A. Martínez-Esparza,
J. Merino,
E. Cera,
J. Bruno,
J. de Pablo,
I. Casas,
J. Giménez, F. Clarens,
M. Rovira
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ABSTRACT: This paper focuses on how to extrapolate current knowledge of spent fuel matrix alteration processes from laboratory to repository conditions, i.e., the influence of changes in both the environmental conditions and the range of time scale considered. Therefore, a spent fuel matrix alteration model allowing the alteration rate evolution to be predicted as a function of both the host rock considered and evaluation time scale of interest is described.At present, the model assumes that alteration of the spent fuel will start when the groundwater reaches the solid surface and that only the radiolytic species of the groundwater (oxidants generated by α-radiation of spent fuel) will produce the surface oxidation process and subsequent matrix dissolution; O2, H2O2 and OH· are the species that react with UO2(s) for oxidation of the pellet surface. The dissolution process of the surface sites that are oxidized is modelled in two steps: first, a surface co-ordination of the oxidized layer with aqueous ligands and, second, detachment (dissolution) of the product species. Taking this mechanism into account, the model gives the evolution of the spent fuel matrix alteration rate over periods as long as 1,000,000 years.In this work the matrix alteration rate results obtained for two repository environments, granitic and argillaceous, will be presented. Furthermore, a sensitivity analysis study has been performed on the influence of the following variables: type of spent fuel considered, α-dose rate evolution, α-range in groundwater, carbonate and iron concentration in groundwater, H2 partial pressure, container time failure and specific surface area of the pellet.
MRS Proceedings. 12/2005; 932.
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ABSTRACT: Understanding the formation of alteration phases on the surface of spent nuclear fuel, such as those observed during leaching experiments, is necessary in order to predict the concentration of radionuclides in the near-field of a final repository. Hydrogen peroxide has been identified as one of the oxidants formed by the radiolysis of water in the presence of spent nuclear fuel; especially due to alpha activity. The presence of this species in solution can contribute to the formation of uranium peroxide secondary phases. In this work, we have studied the oxidative dissolution of synthetic UO2 disks in hydrogen peroxide solutions of two different concentrations (5 x 10(-4) and 5 x 10(-6) mol dm(-3)), both at pH 5.8 +/- 0.1. The solid surface evolution of the disks has been followed by means of ex-situ scanning force microscope (SFM) measurements, and uranium concentration in solution has been determined by inductively coupled plasma mass spectrometry. During the first stage of the experiment, SFM images indicate that only UO2 dissolution is occurring. After 142 h, a secondary phase is observed on the surface of the solid at 5 x 10(-4) mol dm(-3) hydrogen peroxide concentration. This secondary phase has been identified by X-ray diffraction as studtite (UO4 x 4H2O). From the analysis of SFM topographic profiles at different elapsed times, a precipitation rate for the studtite has been estimated to be in the range of (8-32) x 10(-10) mol m(-2) s(-1).
Environmental Science and Technology 01/2005; 38(24):6656-61. · 5.23 Impact Factor
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ABSTRACT: In this work, we studied the effect of β radiation (Sr90 source with an activity of 7 mCi) on the dissolution of non irradiated UO2, as a chemicalanalogue of the spent nuclear fuel, in different leaching solutions.The experiments were carried out using a specifically designed continuous flow-through reactor and in nitrogen atmosphere to avoid oxygen contamination. The solid used was a non irradiated uranium dioxide with a particle size of 100–320 μm, which specific surface are was determined by the BET method. The experiments were carried out in NaCl media at different pH values. Both pH and redox potential of the solutions were continuously monitored. In all the cases, blank experiments were performed in parallel.Dissolution rates obtained under the effect of β radiation were compared with dissolution rates determined in the presence of hydrogen peroxide (the main oxidizing species radiolitically formed by the β radiation according to the CHEMSIMUL code) and with electrochemically determined UO2 corrosion rates found in the literature.
MRS Proceedings. 12/2001; 757.
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ABSTRACT: The dissolution rate of unirradiated uranium dioxide was studied in batch experiments as a function of hydrogen peroxide concentration (from 10−5 to 10−3 mol dm−3). Unirradiated UO2(s) was used in order to differentiate surface chemical processes from radiolytic effects. Dissolution rates were determined from both uranium release and hydrogen peroxide consumption. Results showed that H2O2consumption rate was higher than UO2 dissolution rate. This observation may indicate that the overall UO2 oxidative dissolution process would be controlled by the dissolution of the oxidized solid surface. The calculated hydrogen peroxide reaction order was 1 in the H2O2 concentration range from 10−5 to 10−4 mol dm−3, while at higher concentrations no clear dependence was observed.
MRS Proceedings. 12/1999; 663.
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ABSTRACT: The dissolution of UO2 (a chemical analog of uraninite and of spent nuclear fuel) has been studied by using a flow-through reactor. The UO2 dissolution rates at total concentrations of 10−4, 10−5, and 10−6 mol dm−3 have been determined to be: 1.3 × 10−10 mol m−2 s−1, 6.7 × 10−11 mol m−2 s−1, and 2.0 × 10−11 mol m−2 s−1, respectively. The dissolution rates determined are found to be higher than the ones determined for similar carbonate concentrations. Moreover, the surface of the UO2 has been studied in static tests by means of the scanning force microscopy technique (SFM) in order to follow the formation of any secondary solid phase on its surface. The formation of chernikovite (H2(UO2)2(PO4)2 · 8H2O) has been observed at a 10−4 M total concentration, while no uranyl-phosphate secondary phases have been found at lower concentrations. In experiments performed in the presence of both carbonate and phosphate, no precipitation of secondary phases has been observed. It is postulated that this is due to the formation of the highly stable uranyl-carbonate complexes in solution.
Applied Geochemistry.
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ABSTRACT: Calibration and testing are inherent aspects of any modelling exercise and consequently they are key issues in developing a model for the oxidative dissolution of spent fuel. In the present work we present the outcome of the calibration process for the kinetic constants of a UO2 oxidative dissolution mechanism developed for using in a radiolytic model. Experimental data obtained in dynamic leaching experiments of unirradiated UO2 has been used for this purpose. The iterative calibration process has provided some insight into the detailed mechanism taking place in the alteration of UO2, particularly the role of OH radicals and their interaction with the carbonate system. The results show that, although more simulations are needed for testing in different experimental systems, the calibrated oxidative dissolution mechanism could be included in radiolytic models to gain confidence in the prediction of the long-term alteration rate of the spent fuel under repository conditions.
Journal of Nuclear Materials. 346(1):40-47.
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ABSTRACT: The objective of this work is to study the UO2 oxidation by O2 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 UO2 dissolution rate does depend on. Besides, at 10−4 mol dm−3 bicarbonate concentration, the oxygen consumption rate is almost two orders of magnitude higher than the UO2 dissolution rate. These results suggest that at low bicarbonate concentration (<10−2 mol dm−3) 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 UO2 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 UO2.
Journal of Nuclear Materials. 345:232-238.
