Jordi Bruno

The solubility of (UO2)3(PO4)2 · 4H2O(s) and the formation of U(VI) phosphate complexes: Their influence in uranium speciation in natural waters

Abstract Uranium is commonly associated in natural waters with O-containing ligands both in the aqueous and the solid phases. Phosphate is present in most of these systems; however, the thermodynamics of the U(VI)-H3PO4 system are poorly known, particularly in the pH range of interest 6 to 9. Hence, the effect of phosphate on the migration of uranium in natural waters is not well understood. We have investigated the solubility of a well-characterized U (VI)-phosphate phase in the pH range 3 to 9: (UO2)3(PO4)2 · 4H2O(s) Analysis of these data indicates the formation of the predominant species UO2HPO4(aq) and UO2PO−4 in the pH range 4–9. Their formation constants as well as the solubility constant of (UO2)3(PO4)2 · 4H2O(s) have been determined. From the same experiments we have established the equilibrium constant of the hydroxide complex UO2(OH)3−. A discussion of the hydrolysis of U (VI) is also reported. The effect of phosphate on the mobility of U in natural waters is assessed in the light of these new data. They indicate that in the pH range of most natural waters, 6 to 9, U (VI) will be associated to aqueous phosphate complexes when the total concentration ratio [PO 3− 4 ] T [CO 2− 3 ] T is greater than 10−1.

The kinetics of dissolution of UO2 under reducing conditions and the influence of an oxidized surface layer (UO2+x): Application of a continuous flow-through reactor

Abstract We have studied the kinetics of dissolution of uranium dioxide, UO 2 (s), under strongly reducing conditions (H 2 (g)/Pd). We have investigated the dependence of the rate of dissolution as a function of critical geochemical parameters: pH, pCO 2 , and carbonate concentration. By using a stirred batch reactor, the kinetics of dissolution of an UO 2+x surface layer and the subsequent precipitation of UO 2 have also been studied. Although the initial UO 2 is pretreated before starting the experiments, it seems to be very difficult to avoid the formation of the oxidized surface layer. For this reason we have developed a thin layer continuous flow-through reactor in order to study the kinetics of dissolution of pure UO 2 (s). The dependence of the rate of dissolution of these solids on the proton concentration of the solution may be expressed according to the following equations: r diss ( UO 2 )(mol sec −1 m −2 ) = 1.4(±0.3) × 10 −8 [ H + ] 0.53 ± 0.08 (3 ⩽ pH ⩽ 7) r diss ( UO 2 )(mol sec −1 m −2 ) = 1.9(±0.8) × 10 −12 (7⩽ pH ⩽ 11) r diss ( UO 2+ x )(mol sec −1 m −2 ) = 1.1(±0.3) × 10 −12 [ H + ] −0.30 ± 0.02 (3 ⩽ pH ⩽ 9) mechanism of dissolution of these solids is surface controlled. rate of dissolution of UO 2 in acidic solutions may be described in terms of an integer dependence on the activity of the protonated surface complexes: r diss ( UO 2 )(mol sec −1 m −2 ) = 1.11 × 10 13 {> UOH 2 + } + (2 ⩽ pH ⩽ 6)

On the influence of carbonate in mineral dissolution: I. The thermodynamics and kinetics of hematite dissolution in bicarbonate solutions at T = 25C

The authors have studied the thermodynamics and kinetics of hematite dissolution in bicarbonate solutions under constant pCO{sub 2}. The solubility of hematite is increased in the presence of bicarbonate. They have established that the complexes responsible for this increase are FeOHCO{sub 3} (aq) and Fe(CO{sub 3}){sub 2}{sup {minus}}. The stability constants of these complexes at the infinite dilution standard state are log {beta}{sub 11} = {minus}3.83 {plus minus} 0.21 and log {beta}{sub 2} = 7.40 {plus minus} 0.11, respectively (all errors are given at 2 {sigma} confidence level through this work). The rate of dissolution of hematite is enhanced in bicarbonate solutions. This rate of dissolution can be expressed as R{sub diss} = k{sub 1}(HCO{sub 3}{sup {minus}}){sup 0.23} (mol m{sup {minus}2} h{sup {minus}1}), with k{sub 1} = 1.42 10{sup {minus}7} h{sup {minus}1}. The combination of the study of the surface complexation and kinetics of dissolution of hematite in bicarbonate solutions indicate that the dissolution of hematite is surface controlled and bicarbonate promoted. The rate of dissolution follows the expression R{sub diss} = k{sub HCO}{sub 3{sup {minus}}}{l brace}=FeOH {minus} HCO{sub 3}{sup {minus}}{r brace}, where k{sub HCO{sub 3}{sup {minus}}} = 1.1 10{sup {minus}3} h{sup {minus}1}. The implications of these findings in themorexa0» oxic cycle of iron in natural waters are discussed, most importantly in order to explain the high-Fe(III) concentrations measured in groundwaters from the Pocos de Caldas complex in Brazil.«xa0less

The corrosion of spent UO2 fuel in synthetic groundwater

Abstract Leaching of high burnup BWR fuel for up to 3 years showed that both U and Pu attain saturation rapidly at pH 8.1, giving values of 1–2 mg / l and 1 μg / l respectively. The leaching rate for Sr-90 decreased from about 10 −5 /d to 10 −7 /d but was always higher than the rates for U, Pu, Cm, Ce, Eu and Ru. Congruent dissolution was only attained at pH values of about 4. When reducing conditions were inposed on the pH 8.1 groundwater by means of H 2 /Ar in the presence of a Pd catalyst, significantly lower leach rates were attained. The hypothesis that alpha radiolytic decomposition of water is a driving force for UO 2 corrosion even under reducing conditions has been examined in leaching tests on low burn-up (low alpha dose-rate) fuel. No significant effect of alpha radiolysis under the experimental conditions was detected. Thermodynamically the calculated uranium solubilities in the pH range 4–8.2 generally agreed well with the measured ones, although assumptions made for certain parameters in the calculations limit the validity of the results.

The influence of dissolved carbon dioxide on trace metal speciation in seawater

Abstract Recent experimental work has shown that several trace metal ions (Be(II), Zn(II), Pb(II), Fe(III), Th(IV) and U(VI)) form mixed hydroxo-carbonate complexes in aqueous solution. The kinetics and thermodynamics of formation of these mixed complexes are discussed. Equilibrium calculations indicate that the mixed hydroxo-carbonate complexes are important in surface seawater conditions.

Beryllium(II) hydrolysis in 3.0 mol dm–3 perchlorate

The complex-formation equilibria in the system BeII–H2O have been studied at 25 °C by means of e.m.f. methods, using a coulometric titration technique, in 3.0 mol dm–3(Na)ClO4 solutions. The ranges of total beryllium(II) concentration and acidity were 80 B 1 mmol dm–3 and 2 ⩽–log h⩽ 6.2, respectively. Over the whole concentration range studied the data could be explained by assuming the equilibria (i)–(v). Previously reported information on beryllium(II) hydrolysis is 2Be2++ H2O ⇌[Be2(OH)]3++ H+ log β21=–3.23 ± 0.05 (i), 3Be2++ 3H2O ⇌[Be3(OH)3]3++ 3H+ log β33=–8.656 ± 0.002 (ii), 5Be2++ 6H2O ⇌[Be5(OH)6]4++ 6H+ log β56=–18.81 ± 0.03 (iii), 6Be2++ 8H2O ⇌[Be6(OH)8]4++ 8H+ log β68=–26.70 ± 0.05 (iv), Be2++ 2H2O ⇌ Be(OH)2+ 2H+ log β12=–11.09 ± 0.04 (v) reviewed and an attempt is made to correlate the data in different ionic media by using the Bronsted–Guggenheim–Scatchard specific ion interaction theory. The values of the equilibrium constants at infinite dilution derived from this approach are log β210=–3.47 ± 0.05, log β330=–8.86 ± 0.05, log560=–19.5 ± 0.1, and log β680=–26.3 ± 0.1. Tentative structures for the polynuclear hydroxo complexes are proposed and discussed. The solubility product of α-Be(OH)2(s) has been re-evaluated by using literature data and the hydrolysis constants of this study, at I= 0, log Kso=–6.87 ± 0.05.

Studies of metal carbonate equilibria. 20. Formation of tetra(carbonato)uranium(IV) ion, U(CO3)44−, in hydrogen carbonate solutions

Abstract The equilibrium between the tetracarbonate and pentacarbonate complexes of the U(IV) ion has been studied at 25 °C in CO 2 /HCO 3 − solutions of different ionic strength ( I = 0.5; 1.0; 2.0; 3.0 M NaClO 4 ), using Vis-UV spectrophotometry. The data have been explained by assuming the equilibrium U(CO 3 ) 4 4− + CO 3 2− ⇌ U(CO 3 ) 5 6− where K 5 is the (ionic strength dependent) equilibrium constant. The molar absorptivities for the penta- and tetracarbonate at 660 nm were constant and equal to ϵ 1 = 17.2 ± 0.2 and ϵ 2 = 35.5 ± 0.5, respectively. By using the specific ion interaction theory (S.I.T.), and the equilibrium constants at different ionic strength, the equilibrium at zero ionic strength: 1g K 5 (0) = −1.12 ± 0.25, was obtained. This stepwise equilibrium constant is small and the formation of U(CO 3 ) 5 6− seems to be mainly an ionic strength effect. Hence, very small amounts of U(CO 3 ) 5 6− (and possibly also similar complexes of the other tetravalent actinides) can form in natural waters.

Studies on the Radionuclide Coprecipitation-Solid Solution Formation. The UO 2 (S)-La(OH) 3 (S) Coprecipitation as an Analogue for The UO 2 (S)-Pu(OH) 3 (S) System

The aim of this work is to investigate to which extent coprecipitation will affect the solubility of minor radionuclides under repository conditions. The UO/sub 2/ (s)-La(OH)/sub 3/ (s) coprecipitation equilibria have been studied in 0.5 M NaClO/sub 4/ medium, at 25/sup 0/C under anoxic conditions. The data obtained in the pH range between 2 and 5, and at different initial La(III)/U(IV) ratios (10%, 5%, 3%, 2% and 1%), is explained by applying the Doerner-Hoskins/sup 8/ logarithmic law. The same law can also be used to describe previously collected UO(s)-Pu(OH)number(s) coprecipitation data/sup 2/. A method is proposed which uses the logarithmic distribution constants and the thermodynamic quantities for the pure phases to describe the solubility of radionuclides under repository conditions.

Constraints by Experimental Data for Modeling of Radionuclide Release from Spent Fuel

Comparison of spent fuel corrosion data from nuclear waste management projects in Canada, Sweden and the USA strongly suggests that the release of 90 Sr to the leachant can be used as a measure of the degradation (oxidation/dissolution) of the fuel matrix. A surprisingly quantitative similarity in the 90 Sr release data for fuel of various types (BWR, PWR, Candu), linear power ratings and burnups leached under oxic conditions was observed in the comparison. After 1000 days of leachant contact, static or sequential, the fractional release rates for 90 Sr (and for cesium nuclides) were of the order of 10 −7 /d. The rate of spent fuel degradation (alteration) under oxic conditions can be considered to be controlled either by the growth rates of secondary alteration products, by oxygen diffusion through a product layer, by the rate of formation of radiolytic oxidants or by solubility-controlled dissolution of the matrix. These processes are discussed. Methods for determining upper limits for long-term 90 Sr release, and hence fuel degradation, have been derived from the experimental data and consideration of radiolytic oxidant production.

Preliminary study of spent UO2 fuel corrosion in the presence of bentonite

Abstract The corrosion of spent UO 2 fuel in the presence of a dilute suspension (1.5%) of bentonite in synthetic groundwater has been studied. No significant changes in the uranium concentrations no indications of increased corrosion due to changes in solution chemistry, or due to sorption was found when bentonite was introduced to the system. The measured uranium concentrations were (3 ± 2) × 10 −6 M in the presence as well as in the absence of bentonite. The concentrations of plutonium and cationic fission products in the aqueous phase were lowered considerably, by up to two orders of magnitude in the case of plutonium due to sorption onto the bentonite.