Daqing Cui

The reduction of U(VI) on corroded iron under anoxic conditions

Summary The corrosion of iron and the interaction between corroded iron and U(VI) in anoxic conditions were investigated. The anoxic conditions were obtained by flushing an 99.97% Ar-0.03% CO2 gas mixture through the test vessel, in which an oxygen trap and six reaction bottles containing synthetic groundwater (10 mM NaCl and 2 mM HCO3−.) were placed. The dark-green coloured corrosion product, formed on iron surface after three months corrosion in synthetic groundwater solutions, was identified by powder X-ray diffraction to be carbonate green rust, Fe4IIFe2III(OH)12CO3. The iron foil that reacted in a solution (10 ppm U(VI), 10 mM NaCl and 2 mM HCO3−) for three months was analysed by SEM-EDS. The result shows that: (i) an uneven layer of carbonate green rust (1-5 µm thick) formed on the metallic iron; (ii) a thin (0.3 µm) uranium-rich layer deposited on top of the carbonate green rust layer; and (iii) some UO2 crystals (3-5 µm sized) on the thin uranium layer. The experimental results proved that the U(VI) removal capacity of metal iron is not hindered by formation of a layer of carbonate green rust on the iron. Tests with cast iron and pure iron indicate that they have similar U(VI) removal capacities. At the end of experiment, U concentrations in solution approached the solubility of UO2(s), 10−8 M. The stability of the carbonate green rust at the experimental conditions, pH, Eh, [Fe2+] and [HCO3−], is discussed.

The fate of radiolytic oxidants during spent fuel leaching in the presence of dissolved near field hydrogen

Summary In most European disposal concepts, one expects large amounts of dissolved hydrogen produced by the anoxic corrosion of iron inside a damaged container. At repository temperatures (<100 °C), dissolved hydrogen is quite inert and is not expected to contribute to the redox capacity of the deep groundwaters. In a previous work from this laboratory we observed a large impact of dissolved hydrogen on the dissolution of the spent fuel in stainless steel autoclaves. In this work we report data on spent fuel dissolution obtained using quartz lined autoclaves to ensure that there is no contact between the solution and the metallic parts; we also made a number of other experimental improvements. The autoclave was filled with a solution 10 mM NaCl, 2 mM NaHCO3 and with H2+0.03% CO2 at a pressure of 0.5 MPa in the remaining free volume. The leaching of PWR spent fuel powder, placed in a gold basket, was studied during more than one year by analysing solution samples taken at regular time intervals. Special care was devoted to the study of the initial phase of the dissolution that was not investigated previously. In spite of the ten times lower hydrogen concentration as compared to the study in stainless steel autoclaves, extremely low concentrations of uranium (less than 10-9 M) were again measured in the solution samples; this was the case also for most of the redox sensitive fuel components. The uranium levels in solution remained practically constant during the whole leaching period, indicating the absence of any oxidative dissolution of the spent fuel matrix. In order to follow the fate of radiolytic oxidants, gas phase analyses were also carried out. The radiolytic oxygen levels in the autoclave measured after one year leaching were below the detection limit. A discussion of the fate of the radiolytic oxidants in these experiments and the mechanism of the hydrogen activation is also presented. The main conclusion is that for concentrations of dissolved hydrogen above 4 mM, no measurable oxidative dissolution of the UO2 matrix seems to occur in the studied systems.

On Mo-Ru-Tc-Pd-Rh-Te alloy particles extracted from spent fuel and their leaching behavior under Ar and H2 atmospheres

Summary In the present work, we extracted Mo-Tc-Ru-Pd-Rh-Te alloy particles from spent fuel by a non-oxidative selective leaching of the fuel matrix, characterized them and studied their leaching behavior under different redox conditions. After selecting the optimal experimental conditions by using a synthetic alloy, the radioactive alloy particles were extracted from spent fuel and characterized by using SEM (scanning electron microscopy)-WDS (wavelength dispersive spectra). Micrometer sized alloy particles display very similar compositions, on average (wt.%): 32.7% Mo, 40.5% Ru, 7% Tc, 4.4% Rh, 12.2% Pd and 3.8% Te. The alloy particles were leached in 10 mM NaCl and 2 mM NaHCO3 solutions, purged with two kinds of gas mixtures. It was observed that, under purging with Ar + 0.03% CO2, the leaching rates of 99Tc and 100Mo were similar, about 1.5 ppb/day. The results obtained under purging with 89.97% Ar + 0.03% CO2 + 10% H2 indicate that the 4d alloy particles are very effective catalysts for reduction caused by hydrogen: the concentrations of 99Tc and 100Mo dropped to levels below 1 ppb. No significant leaching of Ru, Rh, Pd and Te was observed during purging with both gas mixtures.

The reduction of U(VI) by near field hydrogen in the presence of UO2(s)

Summary In several recent works a large impact of dissolved hydrogen on spent fuel dissolution has been observed: the solution concentrations of all redox sensitive components of the spent fuel decrease with time, while gas phase analyses indicate levels of radiolytic oxygen below the detection limit. This indicates that at the relatively low temperatures of these studies (≤70 °C) hydrogen is activated. The aim of this study is to test one of the proposed mechanisms for hydrogen activation, namely by UO2(s) surfaces, which may act as hydrogen catalysts. Dissolved U(VI) carbonate species were used as oxidized species in order to test the reducing ability of hydrogen in the presence and absence of UO2(s). In most cases parallel experiments with the same conditions, but under Ar(g) atmosphere were carried out as blanks. Quartz lined autoclaves avoiding any contact of the solution with metallic parts were used. The experiments were carried out in two stages: first the stability of the (10-7-10-5) M U(VI) in 10 mM NaCl, 2 mM NaHCO3 solutions in the presence of dissolved hydrogen was tested. Then a gold net basket containing UO2(s) fragments was introduced in the autoclave and the concentration of U(VI) was followed at different time intervals by ICP-MS. Various dissolved hydrogen concentrations, temperatures and U(VI) concentrations were used, always spanning the range of these parameters in a proposed deep rock repository. Especially at the lowest U(VI) concentrations tested (∼100 ppb) the activation of dissolved hydrogen by UO2(s) surfaces is clearly visible. As a conclusion of this work it may be stated that in the temperature range (20-70 °C) investigated a) dissolved hydrogen does not reduce U(VI) carbonate species in the absence of a catalyst and b) in the presence of UO2(s) surfaces dissolved hydrogen reduces U(VI) in carbonate solutions, very probably to UO2(s).

Surface mediated processes in the interaction of spent fuel α-doped UO2 with H2

In most deep disposal concepts, large amounts of hydrogen are expected to be produced by the anoxic corrosion of massive iron containers. At repository temperatures, hydrogen is quite inert and is not expected to contribute to the redox capacity of the deep groundwaters. In several recent works, a large impact of dissolved hydrogen on the dissolution of the LWR or MOX fuel and UO 2(s) doped with 233U or 238Pu has been observed. For hydrogen concentrations above a certain limit, the dissolution rates of these highly radioactive materials drop to very low values. A discussion of the results obtained with spent fuel or ?-doped UO 2 in the presence of a range of hydrogen concentrations is presented. Typical for all measurements under such conditions are the very low long term concentrations of uranium and other redox-sensitive radionuclides, such as Tc and the minor actinides. The concentrations of U are systematically lower than the values measured during UO2(s) solubility measurements carried out in the presence of strong reducing agents. Measurements of the radiolytic oxygen after long leaching periods result in values below detection limit. The investigation of the surface of spent fuel or UO2(s) pellets doped with 233U by XPS after long periods of testing shows absence of oxidation. The kinetics of the release of non-redox sensitive elements such as Sr and Cs, used to estimate fuel matrix dissolution rates, is also discussed. An attempt is made to propose potential mechanisms responsible for the observed behaviour, based mainly on data from studies on the interaction of water adsorbed on the surfaces of metal oxides or actinide oxides with radiation. Another important effect observed in recent studies is the existence of a threshold for the specific alpha activity below which no measurable influence of the alpha radiolysis on the uranium release from UO2 is observed. The importance of such a threshold for the behaviour of spent fuel under repository conditions encompassing very long time scales will be discussed, as well as the necessity to better investigate the mechanisms of recombination reactions in a thin water layer on the surface of actinide oxides affected by ?- radiolysis.

Redox Reactions of Iron and Uranium Dioxide in Simulated Cement Pore Water Under Anoxic Conditions

Results on the chemical behavior of Fe(0) and UO2(s), as well as the interaction between fresh and corroded iron with U(VI) in simulated cement contacting alkaline solution are reported. Batch experiments were conducted under anoxic conditions at different alkalinities and salt concentrations to investigate: (a) the corrosion of iron foils (b) the U(VI) removal by fresh, aged and pre-treated (with FeS or Fe3O4 layers) iron surfaces in a simulated cement pore fluid, (c) the dissolution rates of newly reduced UO2.00 slices in simulated cement pore fluid and KOH solutions (pH 12.7) and (d) the isotope exchange reactions between dissolved 235U(VI) and 238UO2(s). The reacted iron and UO2(s) surfaces were analyzed by X-ray diffraction (XRD), scanning electron microscopy-energy dispersive spectra (SEM-EDS), laser Raman spectroscopy and X-ray photoelectron, spectroscopy (XPS).

Immobilization of Radionuclides on Iron Canister Material at Simulated Near-field Conditions

This work is a continuation of a long-term spent fuel leaching and radionuclides immobilization (by iron canister) experiment under simulated near-field conditions, in deoxygenated 2 mM NaHCO3 solu ...

On the Interaction between uranyl carbonate and UO2(s) in anaerobic solution

It was repeatedly observed that in nearly neutral anoxic solutions (10 mM NaCl, 2 mM NaHCO3-, pH ~ 8.5) U(VI) concentration decreases from 10-5 M to 10-7 M after 24 days interaction with UO2(s) surfaces. The formation of a layer of hyper-stoichiometric uranium oxide UO2+x(s) on fresh UO2(s) surfaces is proposed as a possible reaction mechanism. Gibbs free energy calculations and other batch experiments were carried out to test the validity of this hypothesis. The calculated negative AG values indicate that the formation of a layer of UO2+x (x=0.25 or 033) is thermodynamically possible. The U(VI)/U(IV) ratio of the reacted uranium oxide surface is 0375, corresponding to UO2.27(s). The average U(VI)/U(IV) ratio obtained for the suspended particles worn off from the reacted UG2(s) surface on an ultrasonic bath resulted 0.162. Our preliminary results from this work support the above hypothesis, which might be also an explanation for some of the results obtained during spent fuel leaching experiments under anaerobic conditions.

Spent Fuel Leaching under Anoxic Conditions and the Effect of Canister Materials

The results of a spent fuel leaching experiment in which a fuel pin (17.7 g) was contacted with 380 mL of a 10 mM NaCl, 2 mM NaHCO 3 solution by taking special care to minimize atmospheric oxygen contamination are presented. During the first 287 days, the fractions of inventory in the aqueous phase per day (f/d) increased nearly constantly for all nuclides (except for 100 Mo), but were higher for fission products f/d( 137 Cs)=1.210 −6 , f/d( 99 Tc)=1.1·10 −6 and f/d( 90 Sr)= 6.7 · 10 −7 than for actinides: f/d ( 238 U) =1.0 · 10 −7 , f/d( 237 Np)= 2.6 · 10 −7 and f/d( 239 Pu) = 5.1 ·10 −9 . After adding iron, cast iron and copper foils (of ∼30 mm 2 size), the concentrations of 238 U, 237 Np and 99 Tc decreased by 80%, 97% and 88% to relatively stable levels (500ppb, 0.2 ppb and 0.6 ppb respectively). 239 Pu concentrations increased from a level around 0.05 ppb to PuO 2 solubility level, 0.5 ppb, and stabilized. The leaching process for 137 Cs, 100 Mo and 90 Sr seems not to be influenced by the addition of metal foils. The observations in the present work contribute to an improved understanding of the behavior of spent fuel under near field repository conditions.

The Surface Precipitation During UO 2 Leaching Process

To understand the influence of cations and silica in groundwater on spent fuel corrosion, the leaching behavior of newly reduced UO 2.00 in different solutions was investigated. Uranium concentrations in the solutions were measured and the leached UO 2 surface and precipitates on it were analyzed by XPS, laser Raman spectroscopy, SEM-EDS and TEM-EDS. It was found that, in air saturated 2 mM HCO 3 - solutions, the leaching rates of UO 2 (s) depend on the composition of the leaching solutions. Very similar leaching rates, 0.5 μg • cm -2 • day -1 , were obtained in 0.1 M NaCl and 0.1 M KCl solutions, while that in 0.45mM Ca 2+ -0.18 mM Mg 2+ -0.2 mM SiO 2 containing synthetic groundwater was 10 times smaller. Calcite and quartz like precipitates were detected on the corroded UO 2 (s) surface.