Uranium Concentration In Solution Research Articles

Dissolution of high burn-up spent nuclear fuel at high-pH

Abstract The release of radionuclides from high-burnup spent nuclear fuel (SNF) segments was studied at pH = 13.2 as well as the effect of the presence of calcium and silicon. The aim was to ascertain the dissolution of SNF in solutions corresponding to a high-level nuclear waste repository including concrete in different structural parts. The release of uranium at pH = 13.2 was higher than at pH = 8.4 in bicarbonate medium, while the presence of calcium resulted in a decrease of the uranium concentrations in solutions, assumed to be the consequence of the formation of a secondary solid phase such as Ca2U2O7. Caesium release was found higher at pH = 13.2 as well, but it was not influenced by the presence of Ca and Si at long term. On the other hand, actinide elements (plutonium, neptunium and americium) dissolution decreased at pH = 13.2, probably because of the formation of secondary solid phases. On the contrary, ruthenium and technetium release at pH = 13.2 was found to be much higher than the measured at lower pH, perhaps due to the higher dissolution kinetics of the metallic inclusions at such pH.

Studies of uranium leaching from core sample in tubes using an oxidizer

Clay content of ore, gangue and clayey cement governs in many ways the primary permeability of rocks and the consumption of chemical agents during uranium leaching. Clayey minerals neutralize sulfuric acid. Consequently, it is necessary to increase sulfuric acid consumption, power input and labor content, and the cost/performance ratio decreases therefore. The article describes application of the borehole uranium mining technology in Kazakhstan. The factors affecting permeability of productive layers are studied in detail. The processes occurring during borehole uranium mining using sulfuric acid as a solvent, as well as the conditions and causes of decrease in the borehole uranium production performance in complex mining and geological conditions in the fields of Kazakhstan are considered. The granulometric composition of core samples from the Chu-Sarysu uranium province was determined. The contents of uranium, calcium, aluminum, iron, magnesium and carbonate in the samples were found using the method of atomic emission spectrometry. The features of the mineral composition of ore-containing rocks were determined by the X-ray phase analysis. A method was developed and the experiments were carried out on a laboratory scale for leaching uranium from the core sample material in tubes, with iron (III) chloride used as an oxidizer in various modes of acidity of working solutions. The laboratory testing data are analyzed and discussed, and the change in the permeability factor Kp, uranium concentration in solution, uranium recovery and sulfuric acid consumption is plotted as function of L:S. The effectiveness of the mild acidity regime with ferric chloride added as an oxidizer is determined and shown.

Investigations on the Interaction of EDTA with Calcium Silicate Hydrate and Its Impact on the U(VI) Sorption

The interaction of EDTA with calcium silicate hydrate (C-S-H) and its impact on the sorption of U(VI) by C-S-H in the presence of EDTA at varying concentrations has been investigated under N2 and ambient atmosphere. The solid phase characterization was performed by FTIR, XRD and TGA measurements and the uranium concentration in solution has been determined by alpha-spectroscopy. At increased EDTA concentrations ([EDTA] > 0.1 M) calcium is complexed and extensively extracted from the solid resulting in a quantitative dissolution of the Ca(OH)2 phase and deterioration of C-S-H. At lower EDTA concentrations ([EDTA] ≤ 0.01 M), EDTA is sorbed into the solid phase and the associated adsorption capacity (qmax = 0.67 mol/kg) has been evaluated by fitting the corresponding data with the Langmuir isotherm model. The incorporation of EDTA in the C-S-H matrix was corroborated by FTIR, XRD and TGA measurements. Regarding the effect of EDTA on the U(VI) sorption by C-S-H, evaluation of the experimental data reveal a significant decrease of the Kd values in the presence of EDTA most probably due to the stabilization of U(VI) in the form of U(VI)-EDTA complexes in solution. Under ambient conditions a further decrease of the Kd values is observed because of the formation of U(VI)-carbonato complexes related to CO2 dissolution and hydrolysis.

FIBROUS ION EXCHANGERS FIBAN AS SORBENTS OF URANIUM(VI) COMPOUNDS PRESENT IN SULFATE SOLUTIONS

Some regularities of Uranium(VI) compounds sorption recovery from model sulfate solutions by fibrous ion exchangers FIBAN C-1 (strongly acidic cation exchanger with -SO3H groups), FIBAN AC-22V (polyampholyte with amino- and carboxyl groups) and FIBAN A-6 (anion exchanger with strongly basic and weakly basic aminogroups) under dynamic conditions were established. It was shown that effective recovery and concentration of Uranium(VI) are combined with a high rate of Uranium(VI) sorption. Quantitative characteristics of Uranium sorption under dynamic conditions such as dynamic exchange capacity, total dynamic exchange capacity, degree of Uranium recovery, Uranium content in the phase of the sorbent, sorption rate were calculated. It was established that the fibrous polyampholyte FIBAN АC-22V demonstrated the better indicators of Uranium(VI) sorption from sulfate solutions and concentration of these solutions (5.2 – 10.5 times) than other fibrous ion exchangers. It was shown that a 10-fold increase in the Uranium concentration in solutions leads to a considerable improvement of the sorption quantitative characteristics. It was found that the initial form of the fibrous ion exchangers can be completely restored by passing through the ion exchangers in Uranium form of a 0.5 M HCl or H2SO4 solutions. It was established that the fibrous polyampholyte FIBAN АC-22V advantageously be used for one cycle of Uranium(VI) sorption – desorption, and the fibrous cation exchanger FIBAN C-1 can be used during two cycles of Uranium(VI) sorption – desorption. The interaction mechanism of the Uranium(VI) compounds present in sulfate solutions (uranyl cations, neutral and negative sulfate complexes) with functional groups of studied sorbents was proposed.

Microstructural evolution of UO2 pellets containing metallic particles of Ru, Rh and Pd during dissolution in nitric acid solution: 3D-ESEM monitoring

Uranium dioxide containing 3 mol% platinum group metals (PGMs) (Ru, Rh, Pd) was synthesized by hydroxide precipitation. The powders were converted to oxides, pelletized and sintered to prepare dense pellets of UO2 incorporating PGMs particles. The characterization techniques performed revealed a microstructure similar to that of spent nuclear fuel (SNF). Dissolution tests in nitric acid demonstrated that in the presence of PGMs, the uranium dissolution rate was increased and the induction period was shortened. We used a new method based on the acquisition of Environmental Scanning Electron Microscopy (ESEM) micrographs recorded at three tilt angles and at different dissolution times. This method allowed the reconstruction of the topography of the solid/liquid interface. By monitoring the evolution of the solid/liquid interface during dissolution by means of 3D reconstructions, we were able to observe preferential dissolution zones in the vicinity of the PGMs particles and to determine microscopic dissolution rates for several regions of interest. PGMs particles were found mainly at the grain boundaries. In 0.1 M HNO3 solution at 60 °C, the normalized dissolution rate for uranium at the grain boundaries reached RL(U) = (7 ± 1) × 10−2 g·m−2·d−1, a value similar to the normalized dissolution rate determined for the whole image over the first 30 days of the experiment. This result showed that the dissolution occurred mainly at the UO2 grain boundaries in the vicinity of PGMs particles. Furthermore, the 3D reconstructions of the solid/liquid interface were used to determine the evolution of the surface area of the pellet. By combining the weight losses determined at the macroscopic scale using the uranium concentrations in solution with the reactive surface area values, it was possible to estimate an effective normalized dissolution rate for the whole pellet.

Uranium extraction from simulated salt lake water by an amidoxime-based UHMWPE fiber

A batch of amidoxime (AO)-based ultra-high molecular weight polyethylene (UHMWPE- g -(PAO-co-PAA)) fiber was prepared by a 7 L graft polymerization reaction kettle. The chemical structure and morphology were evaluated by means of Fourier transform infrared spectrometry and scanning electron microscope, respectively. The uranium adsorption behaviors of the UHMWPE- g -(PAO-co-PAA) fiber were studied by batch adsorption in simulated salt lake water. Effect of the total dissolved solids (TDS) and initial concentration of uranium on the adsorption behaviors of the fiber was investigated. It was found that after 28 d adsorption in the simulated DagzeCo salt lake water, UHMWPE- g -(PAO-co-PAA) fiber with an AO density of 6.0 mmol g−1 exhibited a uranium adsorption capacity of 8.29 mg g−1. The adsorption capacity decreased with the increment of TDS. In the simulated salt lake water in which TDS was 10 times as the sea water, the adsorption capacity of uranium was 6.63 mg g−1. The adsorption capacity increased with the increment of initial concentration of uranium in solution. The uranium adsorption capacity was 4.54 mg g−1 in a solution with initial concentration of 187 µg L−1 uranium, while the value increased to 17.48 mg g−1 with 1639 µg L−1 uranium simulated salt lake water solution. This study should shed light on improvement of uranium extraction from salt lake water and further application in uranium recovery from natural resources.

Dissolution of studtite [UO2(O2)(H2O)4] in various geochemical conditions

This study determined the dissolution rate of studtite, (UO2)O2(H2O)4, which can be formed by reaction between H2O2 and UO22+ that leaks from spent nuclear fuel (SNF) in deep geological repositories. The batch dissolution experiments were conducted using synthesized studtite under different solution conditions with varying pHs and concentrations of HCO3− and [H2O2] in synthetic groundwater. The experimental results suggested that carbonate ligand and H2O2 in groundwater accelerated the dissolution of studtite and uranium (U) release. Above 10−5 M of H2O2 initial concentration, the released uranium concentration in solution decreased, possibly as a result of reprecipitation of studtite due to reaction between uranium and H2O2. The results will be useful to assess the comprehensive transport of uranium from both nuclear waste and SNF stored in deep geological repositories.

Study on treatment of radioactive liquid waste from uranium ore processing by the use of nano oxide ferromagnetic

Nano oxide ferromagnetic Fe3O4 KT wich was produced by the Military Institute of Science and Technology ​​were used to adsorbed heavy metal elements in liquid waste. In this report, the nano oxide ferromagnetic Fe3O4 KT with the particle size of 80-100nm and the specific surface area ​​of 50-70m2/g was applied to study the adsorption of radioactive elements in the liquid waste of uranium ores processing. The effective parameters on adsorption process included temperature, stirring rate, stirring time, the pH value of the solution, the initial concentration of uranium in solution were investigated. The results showed that the maximum adsorption capacity for uranium of the nano Fe3O4 KT was 53.5 mgU/g with conditions such as: room temperature, stirring speed 120 rounds/minute, the pH value of solution was 8, stirring time about 2 hours . From the results obtained, nano Fe3O4 KT was tested to treatment real liquid waste of uranium ore processing after removing almost heavy metals and a part of radioactive elements by preliminary precipitation at pH 8. The results were analyzed on the ICP-MS and α, β total activity equipment, the solution concentration after treatment suitable for Vietnamese Technical Regulation on industrial wastewater QCVN 40: 2011 (concentrations of heavy metals; total activity of α and β).

상추, 배추, 무순, 미나리를 이용한 뿌리여과법(rhizofiltration)의 우라늄으로 오염된 지하수 정화 효율 규명

Lab scale rhizofiltration by using four plants was performed to investigate the uranium removal efficiency from groundwater. Lettuce (Lactuca sativa L.), Chinese cabbage (Brassica campestris L.), radish sprouts (Raphanus sativus L.) and buttercup (Oenanthe javanica) were cultivated during 3 weeks in the phytotron. Glass jar (<TEX>$12cm{\times}12cm{\times}8cm$</TEX> for each), containing 350 ml of the artificially uranium contaminated solution was used for 72 hours of the rhizofiltration. In experiments with different initial uranium concentration (<TEX>$18.00{\mu}g/L$</TEX>, <TEX>$31.00{\mu}g/L$</TEX>, <TEX>$84.00{\mu}g/L$</TEX> and <TEX>$116.00{\mu}g/L$</TEX>) in solution, more than 70% of the initial uranium were removed by using lettuce, Chinese cabbage and radish sprouts and the residual uranium concentration in solution maintained lower than USEPA water tolerance limit (<TEX>$30{\mu}g/L$</TEX>). From the rhizofiltration experiments at various pH conditions, the highest uranium removal for all four plants was acquired at pH 3 in solution. Rhizofiltration experiments testing two field samples of groundwaters having high uranium concentrations (<TEX>$86.00{\mu}g/L$</TEX> and <TEX>$173.00{\mu}g/L$</TEX>) were duplicated and more than 83% of the initial uranium were removed from the groundwater within 72 hours of rhizofiltration by using radish sprouts, which, suggests that the rhizofiltration can be a useful process to remediate uranium contaminated groundwater in the field. After the rhizofiltration experiment, the SEM and EDS analyses for the root surface of the radish sprouts were conducted, suggesting that the main mechanism of the rhizofiltration for the removal of uranium from groundwater would be surface precipitation on the root surface of the plant.

Sorption of uranium from underground leaching solutions with highly basic anion exchangers

Sorption of U on strongly basic anion exchangers AMP, AM-p, Lewatit K6267, Lewatit MonoPlus MP600XL, Purolite A500u, and Purolite A600u was studied at various temperatures in batch and dynamic experiments. The dynamic exchange capacities, (g U) dm−3, of the anion exchangers to a breakthrough of 0.5 (mg U) dm−3 at 6–8°C were determined: AMP (9.56) > Purolite A600u (9.26) > Lewatit K6267 (9.13) > Purolite A500u, Lewatit MonoPlus MP600XL (8.87) > AM-p (5.88). Also, the total dynamic exchange capacities, (g U) dm−3, of the anion exchangers at 6–8°C were determined: AMP (37.13) > Purolite A500u (31.79) > Lewatit MonoPlus MP600XL (28.88) > Lewatit K6267 (27.74) > AM-p (27.58) > Purolite A600u (25.24). The concentration equilibrium distribution coefficients (Kd) of U on these resins were determined in batch experiments. The highest value of Kd = 2869 at 6°C and initial uranium concentration in solution of 65 mg dm−3 was attained with AMP of commercial grain-size distribution. At 24°C and the same uranium concentration, the highest value of Kd = 5195 was obtained with separated 0.8–1.0-mm fraction of AMP. The kinetic experiments were performed, and the internal diffusion step was found to be the limiting step of the U sorption with all the resins under consideration. The internal diffusion coefficients in ion exchanger grains were calculated to compare the resins with each other.

Release of uranium from candidate wasteforms

AbstractLarge volumes of depleted natural and low-enriched uranium exist in the UK waste inventory. This work reports on initial investigations of the leaching performance of candidate glass and cement encapsulation matrices containing UO3 powder as well as that of uranium oxide powders. The surface areas of UO3 powder and the monolith samples of UO3 conditioned in the glass and cement matrices were very different making leaching comparisons difficult. The results showed that for both types of monolith conditioned samples a steady increase of uranium concentration in solution with time was generally not observed. The wt.% of uranium leached from UO3 conditioned in the lead borosilicate glass wasteform was approximately five orders of magnitude less than that leached from UO3 powder. Similarly, the quantities of uranium leached from UO3 conditioned in composite cement made with ordinary Portland cement, and from magnesium phosphate cement, were approximately four and three orders of magnitude, respectively, less than that leached from UO3 powder. The performance of a mixed oxide borosilicate glass wasteform was only slightly better than that of UO3 powder. This work shows that wasteforms based on encapsulation in lead borosilicate glass and cement matrices have the greatest potential for further development.

Uranium levels in Cypriot groundwater samples determined by ICP-MS and α-spectroscopy

The uranium concentration and the isotopic ratio 238U/234U have been determined in Cypriot groundwater samples by ICP-MS after ultrafiltration and acidification of the samples and α-spectroscopy after pre-concentration and separation of uranium by cation-exchange (Chelex 100 resin) and electro-deposition on stainless steel discs. The uranium concentration in the groundwater samples varies strongly between 0.1 and 40 μg l−1. The highest uranium concentrations are found in groundwater samples associated with sedimentary rock formations and the obtained isotopic ratio 238U/234U varies between 0.95 and 1.2 indicating basically the presence of natural uranium in the studied samples. The pH of the groundwater samples is neutral to weak alkaline (7 < pH < 8) and this is attributed to the carbonaceous content of the sedimentary rocks and the ophiolitic origin of the igneous rocks, which form the background geology in Cyprus. Generally, in groundwaters uranium concentration in solution increases with decreasing pH (7 < pH < 8) and this is attributed to the fact that at lower pH dissolution of soil minerals occurs, and uranium, which is adsorbed or forms solid solution with the geological matrix enters the aqueous phase. This is also corroborated by the strong correlation between the uranium concentration and the electrical conductivity (e.g. dissolved solids) measured in the groundwaters under investigation.

Sorption of 60Co from aqueous solutions on layered double hydroxides of Mg, Al, and Nd

Sorption of U(VI) from aqueous solutions of various compositions on layered double hydroxides (LDHs) of Mg, Al, and Nd, and also on layered double oxides (LDOs) of Mg and Al was studied. Experiments on sorption of UO22+ carbonate complexes from aqueous solution on LDO-Mg-Al showed that [UO2(CO3)3]4− ions are weakly captured by LDH-Mg-Al formed by contact of LDO-Mg-Al with water. With an increase in the uranium concentration in solution, Kd of U(VI) decreases. For 3.3 × 10−3 M [UO2(CO3)3]4− solutions, after 24-h contact of the solid and liquid phases, the distribution coefficients do not exceed 1.0 ml g−1 at V/m = 50 ml g−1. From aqueous nitrate solutions, U(VI) is sorbed on LDH-Mg-Al-NO3 poorly: Kd of U(VI) does not exceed 1.0 ml g−1 at 24-h contact of the solid and liquid phases and V/m = 50 ml g−1. At the same time, U(VI) is efficiently sorbed from aqueous 10−3 M UO2(NO3)2 solutions on LDH-Mg-Al and LDH-Mg-Nd with CO32− and OH− ions in the interlayer space. After 15-min contact of the solid and liquid phases, Kd of U(VI) exceeds 5 × 103 ml g−1. With an increase in the UO22+ concentration to 10−1 M, Kd of U(VI) decreases considerably, becoming lower than 25 ml g−1, but increases to ∼150 ml g−1 with an increase in the contact time of the solid and liquid phases to 24 h. The effect of the Na+, Ca2+, Cl−, and SO42− ions and of pH of the initial solution on the U(VI) sorption from aqueous UO2(NO3)2 solutions on LDH-Mg-Al-CO3 was examined.

Enhancing uranium solubilization in soils by citrate, EDTA, and EDDS chelating amendments

A systematic study was made of the effects of three soil amendments on the solubilization of uranium from a granitic soil. The aim was to optimize solubilization so as to enhance bioavailability for the purposes of remediation. The three amendments tested were with citrate, EDTA, and EDDS as chelating agents. The effects of pH, chelator concentration, and leaching time were studied. The most important factor in uranium solubilization was found to be the pH. In the absence of chelating agents, the greatest solubilization was obtained for alkaline conditions, with values representing about 15% of the total uranium activity in the bulk soil. There were major differences in uranium solubilization between the different amendments. The citrate treatment was the most efficient at acidic pH, particularly with the greatest concentration of citrate tested (50 mmol kg −1) after 6 days of treatment. Under these conditions, the uranium concentration in solution was greater by a factor of 356 than in the control suspension, and represented some 63% of the uranium concentration in the bulk soil. Under alkaline conditions, the EDTA and EDDS treatments gave the greatest uranium activity concentrations in solution, but these concentrations were much lower than those with the citrate amendment, and were not very different from the control results. The uranium extraction yield with EDDS amendment was greater than with EDTA.

Statistical analysis and isotherm study of uranium biosorption by Padina sp. algae biomass.

The application of response surface methodology is presented for optimizing the removal of U ions from aqueous solutions using Padina sp., a brown marine algal biomass. Box-Wilson central composite design was employed to assess individual and interactive effects of the four main parameters (pH and initial uranium concentration in solutions, contact time and temperature) on uranium uptake. Response surface analysis showed that the data were adequately fitted to second-order polynomial model. Analysis of variance showed a high coefficient of determination value (R (2)=0.9746) and satisfactory second-order regression model was derived. The optimum pH and initial uranium concentration in solutions, contact time and temperature were found to be 4.07, 778.48 mg/l, 74.31 min, and 37.47°C, respectively. Maximized uranium uptake was predicted and experimentally validated. The equilibrium data for biosorption of U onto the Padina sp. were well represented by the Langmuir isotherm, giving maximum monolayer adsorption capacity as high as 376.73 mg/g.

Resolving the H2 effect on radiation induced dissolution of UO2-based spent nuclear fuel

In recent years, the impact of H2 on α-radiation induced dissolution of UO2-based spent nuclear fuel has been studied and debated extensively. Experimental results on the effect of H2 on the concentration of H2O2 during α-radiolysis have been shown to disagree with numerical simulations. For this reason, the reaction scheme used in simulations of aqueous radiation chemistry has sometimes been questioned.In this work, we have studied the impact of H2 on the H2O2 concentration in α-irradiated aqueous solution using numerical simulations. The effects of H2 pressure, α-dose rate and HCO3- concentration were investigated by performing systematic variations in these parameters. The simulations show that the discrepancy between the previously published experimental result and numerical simulations is due to the use of a homogeneous dose rate (the energy is assumed to be equally distributed in the whole volume). Taking the actual dose rate of the α-irradiated volume into account, the simulation is in perfect agreement with the experimental results. This shows that the H2 effect is strongly α-dose rate dependent, and proves the reliability of the reaction scheme used in the simulations.The simulations also show that H2 influences the H2O2 concentration under α-radiolysis. The magnitude of the effect depends on the dose rate and the H2 pressure as well as on the concentration of HCO3-. The impact of the radiolytic H2 effect on the rate of α-radiation induced dissolution of spent nuclear fuel is discussed along with other (α- and γ-) radiation induced processes capable of reducing the concentration of uranium in solution. The radiolytic H2 effect is quantitatively compared to the previously presented noble metal catalyzed H2 effect. This comparison shows that the noble metal catalyzed H2 effect is far more efficient than the radiolytic H2 effect. Reduction of U(VI) in solution due to low dose rate γ-radiolysis in the presence of H2 is proposed to be the cause of the H2 effect observed in leaching experiments on α-doped UO2.

Hydrogen suppresses UO 2 corrosion

Release of long-lived radionuclides such as plutonium and caesium from spent nuclear fuel in deep geological repositories will depend mainly on the dissolution rate of the UO 2 fuel matrix. This dissolution rate will, in turn, depend on the redox conditions at the fuel surface. Under oxidative conditions UO 2 will be oxidised to the 1000 times more soluble UO 2.67. This may occur in a repository as the reducing deep groundwater becomes locally oxidative at the fuel surface under the effect of α-radiolysis, the process by which α-particles emitted from the fuel split water molecules. On the other hand, the groundwater corrodes canister iron generating large amounts of hydrogen. The role of molecular hydrogen as reductant in a deep bedrock repository is questioned. Here we show evidence of a surface-catalysed reaction, taking place in the H 2–UO 2–H 2O system where molecular hydrogen is able to reduce oxidants originating from α-radiolysis. In our experiment the UO 2 surface remained stoichiometric proving that the expected oxidation of UO 2.00 to UO 2.67 due to radiolytic oxidants was absent. As a consequence, the dissolution of UO 2 stopped when equilibrium was reached between the solid phase and U 4+ species in the aqueous phase. The steady-state concentration of uranium in solution was determined to be 9 × 10 −12 M, about 30 times lower than previously reported for reducing conditions. Our findings show that fuel dissolution is suppressed by H 2. Consequently, radiotoxic nuclides in spent nuclear fuel will remain immobilised in the UO 2 matrix. A mechanism for the surface-catalysed reaction between molecular hydrogen and radiolytic oxidants is proposed.

UO2dissolution in the presence of hydrogen peroxide at pH&gt;11

AbstractThe dissolution of non irradiated UO2was 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 H2O2consumption 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−1at 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 UO2surface.

Uranium chemistry in stack solutions and leachates of phosphogypsum disposed at a coastal area in Cyprus

The effect of the matrix composition (main constituents) on the concentration and chemical behavior of uranium in phosphogypsum stack solutions and leachates has been investigated. Solid and aqueous samples were taken from three different sub-areas of a phosphogypsum stack at a coastal area in Vasilikos (Cyprus). The sub-areas are characterized whether by their acidity (e.g. “aged” and “non-aged” phosphogypsum) or by their salt content, originating from pulping water during wet stacking or (after deposition) from the adjacent sea. Measurements in stack solutions and leachates showed that phosphogypsum characteristics affect both, the concentration and the chemical behavior of uranium in solution. Uranium concentration in solutions of increased salinity is up to three orders of magnitude higher than in solutions of low salinity and this is attributed to the effect of ionic strength on the solubility of phosphogypsum. Modelling showed that uranium in stack solutions is predominantly present in the form of uranium(VI) phosphate complexes (e.g. UO 2(H 2PO 4) 2, UO 2HPO 4), whereas in leachates uranium(VI) fluoro complexes (e.g. UO 2F 2, UO 2F 3 −) are predominant in solution. The latter indicates that elution of uranium from phosphogypsum takes places most probably in the form of fluoro complexes. Both, effective elution by saline water and direct migration of uranium to the sea, where it forms very stable uranium(VI) carbonato complexes, indicate that the adjacent sea will be the final receptor of uranium released from Vasilikos phosphogypsum.

Spent fuel UO2 matrix alteration in aqueous media under oxidizing conditions

The effects of αβγ radiolysis of water on oxidation and dissolution of the UO2 matrix of spent fuel were investigated in aerated media. The prospect of long-term interim storage of spent fuel assemblies in pools requires prior examination of an incident scenario in which a cladding failure could eventually lead to contact between water and the fuel. Radiolytic oxidation of the matrix together with the formation of secondary phases subject to volume expansion could then no longer be disregarded, and could enhance the degradation of the failed rod. To determine the amplitude of the alteration various spent fuel leaching experiments conducted at different degrees of oxidizing conditions and of accessibility of species at the reaction interface are discussed. This work is based on modeling the radiolysis processes and on various possible interpretations of leaching test results. Properly interpreting these results implies full knowledge of the fuel sample history, and the development of a rigorous experimental approach to eliminate any artifacts (problems due to the residual inventory at the grain boundaries, or to inaccurate estimates of the actual surface area). The results indicate that despite highly variable oxidizing conditions (presence of an external γ irradiation source, addition of H2O2, aerated water) the spent fuel matrix alteration determined from the release of tracer elements such as Cs and Sr is always the same. Conversely, allowing for the alteration conditions (no solution renewal and no complexants), the uranium concentrations in solution do indeed depend on the oxidizing conditions as does the nature of the secondary phases liable to precipitate.