Javier Giménez

Arsenic removal by goethite and jarosite in acidic conditions and its environmental implications

Schwertmannite (Fe(8)O(8)(OH)(5.5)(SO(4))(1.25)), jarosite (KFe(3)(SO(4))(2)(OH)(6)) and goethite (FeOOH) control natural attenuation of arsenic in acid mine drainage (AMD) impacted areas. Batch experiments were conducted to examine the sorption capacity of synthetic goethite and synthetic jarosite at highly acidic pH (1.5-2.5), at two ionic strengths (0.02-0.15 mol dm(-3), NaCl) and at sulphate concentrations in the range of 5 x 10(-3) to 2.8 x 10(-1) mol dm(-3). In the absence of competitive effects of other anions, K-jarosite presents better removal efficiency than goethite for As(V). The maximum sorption capacity is estimated to be 1.2 x 10(-4) and 7.0 x 10(-6)mol m(-2) for jarosite and goethite, respectively, under similar experimental conditions. The variation of arsenic sorbed on goethite as a function of the equilibrium arsenic concentration in solution fits a non-competitive Langmuir isotherm. In the case of K-jarosite, sorption data could not fit a Langmuir or Freundlich isotherm since sulphate-arsenate anion exchange is probably the sorption mechanism. Ionic strength and pH have little effect on the sorption capacity of goethite and jarosite in the small range of pH studied. The presence of sulphate, which is the main anion in AMD natural systems, has a negative effect on arsenic removal since sulphate competes with arsenate for surface sorption sites. Moreover, mobilization of arsenic in the transformation of schwertmannite to jarosite or goethite at pH 2-3 is proposed since the sorption capacities of goethite and K-jarosite are considerably lower than those reported for schwertmannite.

The role of pe, pH, and carbonate on the solubility of UO2 and uraninite under nominally reducing conditions

Abstract Experimental data obtained from uranium dioxide solubility studies as a function of pH and under nominally reducing conditions in a 0.008 mol dm −3 perchlorate medium and in a 1 mol dm −3 chloride solution are presented. The solubility of extensively characterized uraninite samples from Cigar Lake (Canada), Jachymov (Czech Republic), and Oklo (Gabon) was determined in a solution matching the composition of a groundwater associated with granitic terrain. The redox potential of the test solution was monitored throughout the experimental period. The results obtained were modeled using aqueous formation constants compiled by the NEA, using stability constants corrected to appropriate ionic strengths. The solubility curves have been adjusted by calculating the value of K s4 (UO 2(s) + 2H 2 O ⇔ U(OH) 4(aq) ) that gave the best fit with the experimental data. For a low temperature synthetic UO 2 , a value of logK s4 of −7.3 was determined, while for uraninites the best fit was obtained with a value of logK s4 of −8.5. A wide range of published UO 2 solubilities can be reproduced by the available database, where experimental conditions were adequately defined in the original experiments. A lower value of the solubility product of the uranium dioxide phase defined as fuel in the SKB uranium database provides reasonable solubilities for a wide span of experimental results at near to neutral pH. Based on the modeling and using the β 1,4 for the U(IV)-OH complexation given by Grenthe et al. (1992a) , a logK s0 (UO 2 (s) + 4H + ⇔ U 4+ + 2H 2 O) value of −2.3 ± 0.2 is proposed. Differences in solubility between natural and synthetic samples are attributed to the presence of carbonate in the experiments performed with uraninites, while differences in solubility observed among the natural samples can be correlated to radiation effects at atomic scale.

Sorption of Antimony (V) onto Synthetic Goethite in Carbonate Medium

Abstract The sorption kinetics of antimony(V) on synthetic goethite is very fast compared to the sorption of other metals on goethite (e.g. arsenic and selenium) and depends on temperature, with an activation energy of 49±9 kJ · mol−1 in the temperature range 15–35°C. Sorption isotherms have been developed at different temperatures and ionic strength values. The results have been modelled using a Langmuir isotherm and there is not a considerable influence of neither the temperature in the range studied (15°C–35°C), nor the ionic strength (between 0.001 and 0.01 mol · dm−3). Sorption is very high at pH values lower than 8, at more alkaline pH, the sorption decreases with pH, as expected considering the Antimony(V) predominating complex in solution, Sb(OH)6 −. Triple‐layer model successfully describes the data obtained by assuming a bidentate edge‐sharing surface complex of antimonate on the surface of goethite.

Reactive transport of arsenic(III) and arsenic(V) on natural hematite: experimental and modeling.

Natural hematite was used for the removal of arsenic(III) and arsenic(V) from aqueous solution. The experimental breakthrough curves were obtained in fixed-bed columns. The transport of arsenic in a simplified fixed-bed configuration was quantified by using the CXTFIT code, which was used to estimate the transport and sorption parameters of the convective-dispersive equation (CDE) and the two-site deterministic nonequilibrium (TSM/CDE) model by fitting the models to the experimental breakthrough curves (BTC). The prediction of the breakthrough curves performed by the two-site nonequilibrium sorption model resulted in a good fit, indicating that this model can properly describe the transport and sorption processes of arsenic on natural hematite. Additionally the parameters obtained indicate that nonequilibrium sorption governs the As(III) and As(V) uptake onto hematite in a fixed-bed column. No significant differences in the transport and sorption parameters of As(III) and As(V) on natural hematite were obtained; the retardation factor values were in the same order of magnitude for both species.

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.

Influence of β radiation on UO2 dissolution at different pH values

Summary In this work we studied the effect of external β radiation (90Sr- 90Y source with an activity of 7 mCi) on the dissolution rate of non-irradiated UO2 as a chemical analogue of spent nuclear fuel (SF). The experiments were carried out at three different pH values inside a glove-box in nitrogen atmosphere to avoid oxygen contamination. The MAKSIMA code was used to model the generation of radiolytic products, both molecular species and radicals. The formation of hydrogen peroxide in solution was observed in the experiments, as was predicted by the MAKSIMA code. When H2O2 could be quantified, its concentration was within the range predicted by this code. In addition, the application to our data of an empirical model for UO2 dissolution in the presence of H2O2 gave similar results. The UO2 dissolution rates obtained in this work were similar to the UO2 corrosion rates determined electrochemically and under γ irradiation with a similar dose rate. On the other hand, they were always lower than the ones obtained with fresh spent nuclear fuel, probably because in this case the dose rates to the solution are higher and, in addition, more than just β radiation is emitted by the fuel.

Cesium sorption on studtite (UO2O2 · 4H2O)

Abstract One of the mechanisms that may decrease the mobility of cesium released from spent fuel in a high level nuclear waste repository (HLNW) is its sorption onto uranyl-containing alteration phases formed on the spent fuel surface such as studtite (UO2O2·4H2O). The results obtained in this work show that sorption is a very fast process; cesium in solution is sorbed in less than one hour at pH 5. Sorption as a function of initial concentration in solution was also studied between initial cesium concentrations ranging from 7.6×10−9 mol dm−3 to 1.0×10−3 mol dm−3. The data have been modelled considering a Freundlich isotherm, with KF and n values of 10±1, and 1.4±0.1, respectively (r2=0.998). Sorption is very dependent on ionic strength, suggesting that cesium sorbs onto studtite by forming an outer-sphere complex involving electrostatic interactions. Sorption is observed to be very low at acidic pH, while relatively high at alkaline pH ( i.e. , almost 60% of the total cesium concentration in solution is sorbed at pH>9). The results point to the importance of sorption processes on uranyl alteration phases on the retention of radionuclides.

Stability of uranium (VI) peroxide hydrates under ionizing radiation

Abstract The uranyl peroxide, studtite (UO4⋅4H2O, C2/c, Z = 4), is expected to form as a consequence of alpha radiolysis of water in contact with spent nuclear fuel (SNF) in a geologic repository. Investigation of its stability is, therefore, of critical importance because secondary U(VI) phases may incorporate trace amounts of radionuclides and thus retard their mobility away from a repository site. To examine the effect of ionizing radiation on uranyl peroxides, electron-beam irradiation experiments have been conducted on two synthetic uranyl peroxides: studtite and metastudtite (UO4⋅2H2O, Immm, Z = 2). All experiments were done using a transmission electron microscope (TEM) with an acceleration voltage of 200 kV at room temperature. The fluence required to completely amorphize studtite was 0.51-1.54 × 1017 e/cm2, which is equivalent to an absorbed dose of 0.73-1.43 × 107 Gy. Metastudtite becomes amorphous at a higher absorbed dose (1.31 × 107 Gy) than studtite, most likely because it contains fewer water molecules in its structure. These uranyl peroxides partially amorphize at doses that are one-tenth of the dose required for complete amorphization. With continued irradiation, uraninite nanocrystals form that are a few nanometers in diameter, at 4-20 × 1010 Gy. In a geologic repository, for spent nuclear fuel, the estimated absorbed doses due to ionizing radiation may be as high as 108-1011 Gy after 106 years. This is well in excess of doses in the laboratory experiments that caused the uranyl peroxides to become amorphous and decompose.

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.

UO2 dissolution in the presence of hydrogen peroxide at pH > 11

Abstract The dissolution of non irradiated UO2 was studied in the presence of hydrogen peroxide (10−4 mol dm−3) at alkaline pH (11, 11.5, 12, and 13). Both hydrogen peroxide and uranium concentration in solution were determined as a function of time. The H2O2 consumption was modelled considering a pseudo first order reaction, the rate constants obtained were 0.126±0.004 h−1, 0.126±0.003 h−1, 0.078±0.002 h−1, and 0.056±0.002 h−1 at pH 11, 11.5, 12, and 13, respectively. The uranium concentrations measured at the end of the experiments were close to the solubility of sodium uranate (Na2U2O7). X-ray photoelectron spectroscopy showed that the surface of the solid was more oxidized at lower pH, indicating that at this pH the limiting step of the oxidative dissolution process is the dissolution of the U(VI) formed on the surface. At more alkaline pH values, the rate limiting step would be the oxidation of the UO2 surface.