F. Clarens

Combined effect of H2O2 and HCO3- on UO2(s) dissolution rates under anoxic conditions

Abstract The influence of both hydrogen peroxide (H2O2) and bicarbonate (HCO3-) on the dissolution of UO2 has been studied in this work. Two different series of experiments have been carried out using a flow-through reactor. In the first series, the influence of H2O2 concentration (between 10-6 and 5×10-4 mol dm-3) on the dissolution rate of UO2 has been studied at a fixed bicarbonate concentration of 2×10-3 mol dm-3. An increase in the dissolution rate is observed as the concentration of hydrogen peroxide increases. In the second series, the influence of bicarbonate (between 10-4 and 10-2 mol dm-3) on the dissolution rate of UO2 has been studied in the presence of a fixed hydrogen peroxide concentration (10-4 mol dm-3). The main result was that UO2 dissolution rates increased with bicarbonate concentration. From the experimental data, an oxidative dissolution model has been developed that can reproduce spent nuclear fuel dissolution rates obtained under relatively low oxygen concentrations. Under these conditions, the influence of radiolysis products, rather than O2 concentration, is expected to determine the oxidative dissolution rates of the fuel.

Leaching of 53 MW/d kg U spent nuclear fuel in a flow-through reactor

Abstract The dissolution behaviour of powdered commercial spent fuel (UO2 with burn-up of 53 MW/d kg U) has been studied in a carbonate-containing solution ([HCO3-] =0.001 mol dm-3) by using a flow-through reactor specially designed for the use in a hot cell. This method allows studying spent fuel dissolution while avoiding the parallel process of secondary solid phase formation. The dissolution behaviour of U, Np, Pu, Sr and Cs was studied. The main trend of the results obtained in this work is that only neptunium releases congruently with uranium (FIAPNp/FIAPU=1.21±0.01) because both strontium and caesium have higher FIAP values (FIAPSr/FIAPU=2.3±0.8; FIAPCs/FIAPU=5±1) and plutonium lower (FIAPPu/FIAPU=0.07±0.02). The FIAP value for uranium at the steady-state is 4(±2)×10-4.

Surface Site Densities of Uranium Oxides: UO 2 , U 3 O 8

The estimation of the surface site density for UO 2 (main component of the SF matrix) is an important aspect to take into account in the development of radiolytical models for the dissolution of the spent nuclear fuel (SF). Also, other oxides can be formed on the SF surface due to the effect of radiolytically-formed oxidizing species. Due to the lack or reliable data in literature we have studied the surface site densities of two uranium oxides: UO 2 and U 3 O 8 . For this determination, the knowledge of both the reactive surface area and the surface charge of the solid are necessary. In this work the reactive surface area of the two solids was determined by the BET method while the surface charge was determined by potentiometric acid-base titrations of suspensions of the solids at different ionic strengths (0.001, 0.01, and 0.1 mol·dm −3 ) under N 2 atmosphere. The amount of adsorbed protons was calculated by subtracting blank titration from thesolid suspension titration. The single-site nonelectrostatic model (NEM) was used to describe titration data. Uranium speciation in solution was included in the model as well. The surface area values obtained were 0.15 ± 0.01 m 2 ·g −1 for UO 2 and 0.77 ± 0.02 m 2 ·g −1 for U 3 O 8 , while the surface acidic site densities were determined to be 165 ± 10 sites·nm −2 and 48 ± 3 sites·nm −2 for UO 2 and U 3 O 8 , respectively.

Wastewater treatment by MBR pilot plant: flat sheet and hollow fibre case studies

Abstract The design and implementation of membrane bioreactor (MBR) pilot plant was performed in order to treat municipal wastewater derived from the Suria municipality (Catalonia, Spain) wastewater treatment plant. Two submerged membrane configurations (flat sheet and hollow fibre) in MBR pilot plant were used for this purpose. The influent and effluents were monitored and controlled in order to ensure the achievements of the highest quality determined by Spanish legislation for water reuse. The Remosa company interest was focused in pilot plants applications for the small urban areas. Taking into account that the level of control and maintenance of this small plant can be lower than recommended, the start-up was performed under less favourable conditions without any sludge seed. After 8-months of continuous operation, the physico-chemical and microbial parameters of both MBR configurations achieved the water quality specifications defined for urban service, agricultural and recreational uses. The flat s...

RN Fractional Release of High Burn-Up Fuel: Effect of HBS and Estimation of Accessible Grain Boundary

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.

Sorption of Caesium on Commercial Magnetite with low Silica Content: Experimental and Modelling

Interaction of Cs on commercial magnetite has been studied under different experimental conditions at high solid/liquid ratios. At pH 8.2 the Freundlich isotherm was able to describe an important range of the experimental data and at high Cs concentration surface precipitation was found likely to occur. Special attention has been given to silica, present in the system as an impurity of magnetite, since it plays an important role on the sorption mechanism in which a ternary complex magnetite-Si-Cs is formed.

Sorption of Molybdenum(VI) on Synthetic Magnetite

In this study we experimentally investigated the interaction of Mo(VI) with magnetite, which is a corrosion product of steel. The work was conducted with commercial magnetite, and we studied the influence of pH, pe and solid/liquid ratio on Mo sorption. A Surface Complexation Model (SCM) has been applied tothe experimental data, allowing to explain the results using the Diffuse Layer Model (DLM) and by considering the formation of the monodentate complex: >FeOMo(OH) 5 . At pH 2, experimental data were satisfactorily fitted to a a Langmuir isotherm.

Effect of β-Radiation on the Non Irradiated UO 2 (s) Dissolution

In this work, we studied the effect of β radiation (Sr 90 source with an activity of 7 mCi) on the dissolution of non irradiated UO 2 , 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 UO 2 corrosion rates found in the literature.