Solubility Of Actinides Research Articles

Speciation of the rare earth element neodymium in groundwaters of the Nevada Test Site and Yucca Mountain and implications for actinide solubility : K. H. Johannesson, K. J. Stetzenbach & V. F. Hodge, Applied Geochemistry, 10(5), 1995, pp 565–572

Speciation of the rare earth element neodymium in groundwaters of the Nevada Test Site and Yucca Mountain and implications for actinide solubility : K. H. Johannesson, K. J. Stetzenbach & V. F. Hodge, Applied Geochemistry, 10(5), 1995, pp 565–572

Optimization of Lanthanide Borosilicate Frit Compositions for the Immobilization of Actinides Using a Plackett-Burman/Simplex Algorithm Design

ABSTRACTImmobilization by vitrification is one potential disposition option for a portion of the United States' excess plutonium inventory. Research has been performed at the Savannah River Site (SRS) to determine the optimum composition of a lanthanide borosilicate frit for the vitrification of plutonium using a Plackett-Burman design and simplex algorithm as a statistical tool. This technique uses various response variables to rank and optimize a composition. The variables used in this study correspond to homogeneity, durability, actinide solubility and devitrification after heat-treatment.The optimized frit composition was determined using a constant ThO2 loading of 20 wt%. No noticeable trends were followed with respect to the individual components which may indicate a relatively robust system able to accommodate variations in the feed.Batches containing various loadings of ThO2 were melted to determine if actinide solubility was improved in the optimized composition compared to that of a similar lanthanide borosilicate glass. No noticeable improvement in ThO2 solubility was realized as a result of using this optimization technique.

Speciation of the rare earth element neodymium in groundwaters of the Nevada Test Site and Yucca Mountain and implications for actinide solubility

Concentrations of the rare earth element (REE) Nd and major ions were measured in groundwater samples obtained from 4 wells on the Nevada Test Site and one well located 3 miles east of the proposed high-level nuclear waste repository at Yucca Mountain, adjacent to the Nevada Test Site. The speciation of dissolved Nd was modelled using a combined specific ion interaction (i.e., Pitzer Model) and ion pairing model. The model provided a means of determining the activity coefficients of the Nd 3+ ion (ym), complexing anions (γx), and Nd 3+ -anion ion pairs (γM MX). The modelling indicated that between 92% and > 99% of dissolved Nd is complexed with CO 3 2− as either carbonato (NdCO 3 +) or dicarbonato (Nd(CO 3) 2 −) complexes. An elevated Nd concentration in one of the groundwater samples was associated with higher calculated free CO 3 2− concentrations (i.e., [CO 3 2−] F, where the subscript F denotes free concentrations). Based on the relatively large stability constants for Nd CO 3 complexes as compared with stability constants for other inorganic ligands, it is proposed that elevated [CO 3 2−] F concentrations in groundwaters can result in enhanced dissolved Nd concentrations. The analogous chemical behavior of the trivalent REEs, especially Nd 3+, and the trivalent actinide series elements (Am 3+, Cm 3+, Cf 3+), suggests that these actinides will be primarily inorganically complexed with CO 3 2− in the local groundwaters and that elevated [CO 3 2−] F concentrations may increase the ability of these groundwaters to solubilize and transport these actinides.

Thermodynamic Modeling of Neptunium(V) Solubility in Concentrated Na-CO3-HCO3-Cl-ClO4-H-OH-H2O Systems

AbstractSafety assessments of nuclear waste repositories often require estimation of actinide solubilities as they vary with groundwater composition. Although a considerable amount of research has been done on the solubility and speciation of actinides,1,2relatively little has been done to unify these data into a model applicable to concentrated brines. Numerous authors report data on the aqueous chemical properties of Np(V) in NaClO4, Na2CO3, and NaCl media, but a consistent thermodynamic model for predicting these properties is not available. To meet this need, a model was developed to describe the solubility of Np(V) in Na-Cl-ClO4-CO3aqueous systems, based on the Pitzer activity coefficient formalism for concentrated electrolytes. Hydrolysis and/or carbonate complexation are the dominant aqueous reactions with the neptunyl ion in these systems. Literature data for neptunyl ion extraction and solubility are used to parameterize an integrated model for Np(V) solubility in the Np(V)-Na-CO3-HCO3-Cl-ClO4-H-OH-H2O system. The resulting model is tested against additional solubility and extraction data, and compared with Np(V) solubility experiments in complex synthetic brines.

Geochemical Modeling of a Bentonite-Groundwater-Rock System in the Presence of Uranium

Many concepts for deep underground storage of high level radioactive waste include the use of bentonite (sodium-montmorillonite) as a backfill material surrounding the waste. It is therefore necessary to model the chemical conditions in the system sodium-montmorillonite-granitic groundwater-granitic rock in order to be able to predict the speciation and solubility of actinides and fission products in this environment. In this study, the chemical interactions and the evolution of the chemical composition of the water in such a system have been modeled using the geochemical computer code PHREEQE. The parameters considered are the pH, Eh, and the chemical composition in aqueous solution. Mineral formation has been taken into account in some cases. The speciation and solubility of uranium in this system have also been calculated.

Dependence of Actinide Solubility and Speciation on Carbonate Concentration and Ionic Strength in Groundwater

Article Dependence of Actinide Solubility and Speciation on Carbonate Concentration and Ionic Strength in Groundwater was published on May 1, 1992 in the journal Radiochimica Acta (volume 58-59, issue 1).

Thermodynamic modelling of cement degradation: Impact of redox conditions on radionuclide release

Portland cement, used as a solidification matrix and/or as backfill and construction material will determine the aqueous chemistry of a low- and intermediate level radioactive waste repository for very long periods of time. Based on a recently developed cement degradation model, the present work aims to investigate the influence of Eh on the solubility of actinides during the degradation of cement. Uranium is used as a representative for the actinides. Literature data are used to estimate the solubility of uranium bearing solid phases, which were found to be stable in cement-like systems. Log K so values of -41 and of -66 are estimated for a solid called x-phase and for uranophane. In a strongly reducing environment (−450 mV) uraninite is the stable solid phase and uranium solubilities in the range 10 −8 to 10 −10 M are modelled. Under moderately reducing conditions (−300/−200 mV) the solubility limiting phase depends on the pH of the cement pore solution, and at pH∼12 a phase transformation is predicted. Modelled solubilities lie between 10 −7 and 10 −10 M. Uranium speciation in solution and the nature of the solubility limiting phase are very sensitive to both Eh and pH, whereas total dissolved uranium is not. It is concluded that the predictability is fair for strongly reducing conditions, but the uranium-cement system is not very well understood at moderately reducing conditions. The study raises a series of still unresolved questions, which clearly indicate the need of further experimental data.

The solubility of actinides in a cementitious near-field environment

The solubilities of thorium, protactinium, uranium, neptunium, plutonium, and americium under the chemical conditions representative of a cementitious near-field environment have been studied in many national radioactive waste management research programmes. This paper presents the results and conclusions from the UK Nirex Safety Assessment Research Programme, which has drawn on this body of data and conducted a number of validation experiments for the conditions typical of the UK repository design. The solubilities range from about 10 −6 M for hexavalent uranium to about 10 −10 M for americium at pH 12. The experimentally obtained data have been supported by interpretive simulations using the geochemical speciation code HARPHRQ.

Geochemical modelling of the long-term dissolution behaviour of the French nuclear glass R7T7

The long-term dissolution behaviour of the French nuclear reference glass R7T7 was studied by means of the geochemical code DISSOL. New experimental data which support some of the assumptions of DISSOL are presented: namely, that the dissolution is congruent and that the altered layer can be considered as an assemblage of secondary phases. At 100°C the main results of modelling are that the altered layer is essentially formed of a pure siliceous phase (amorphous silica or chalcedony) associated with smectites and zeolites. This sequence of secondary minerals is closely linked to the chemical composition of the glass. For high degrees of reaction, corresponding to high B concentration, the ionic strength reaches 1 and the pH varies from 9 to 10 depending on the CO 2 fugacity; B, Li and Na are essentially found in solution and their concentrations depend on the amount of dissolved glass. On the contrary, Fe, Al and Zn have low solution concentrations which are controlled by solubility products of secondary minerals. Silicon and Ca have an intermediate behaviour which depends on the choice of selected secondary minerals. The total volume of the secondary phases is always lower than that of the corresponding dissolved glass. The results of modelling compared to static leaching experimental results show only minor differences which can be explained by kinetic control or colloid formation. It is concluded that the altered layer is not a barrier to diffusion. The consequences of this work on actinide solubility are also discussed.

Colloid Formation During Waste Form Reaction: Implications for Nuclear Waste Disposal

Insoluble plutonium- and americium-bearing colloidal particles formed during simulated weathering of a high-level nuclear waste glass. Nearly 100 percent of the total plutonium and americium in test ground water was concentrated in these submicrometer particles. These results indicate that models of actinide mobility and repository integrity, which assume complete solubility of actinides in ground water, underestimate the potential for radionuclide release into the environment. A colloid-trapping mechanism may be necessary for a waste repository to meet long-term performance specifications.

Solubility in the rat lung of actinides associated with estuarine silt from West Cumbria

The solubility of actinides, associated with contaminated estuarine silt from West Cumbria, was measured in the rat lung. This material is similar mineralogically to airborne particles of silt involved in the sea-to-land transfer of radioactivity. A total of 25 mg of silt was administered to rats by multiple intratracheal instillations over a period of 7 weeks. There was no evidence of the rapid lung clearance phase normally observed in the rat and, after an initial period of stasis, most of the plutonium and americium was cleared from the lung with a half time of approximately 240 days. Some 15% of the retained plutonium isotopes and 10% of the 241Am were transferred to the skeleton over a period of about 3 months, indicating that both behave as ICRP Class W materials. Uptake by the liver was about 2% for both Pu and Am. Levels of actinides in the lung-associated lymph nodes were either at or blow the minimum level of detection. The significance of these results in relation to the ICRP lung model and to generalised derived limits for plutonium is discussed.

Americium and plutonium in concentrated salt solutions and their interactions with geological material specific to a waste repository in a salt environment

Experiments have been carried out into the leaching by salt brines of real and simulated intermediate level waste conditioned in cement. In addition, standard actinide solutions have also been prepared by dissolving aged hydroxides in these salt solutions. After contact with the cement, the pH of the solutions rose rapidly from their natural values of about 6 to a value of 12 or more and in the case of a quinary solution from 4 to about 7. As a consequence, maximum actinide solubilities of about 10 −8M and 10 −6M respectively were obtained. Little colloidal material was observed. The solubilities of the standard hydroxides varied between 10 −3 and 10 −8M at pH values between 1 and 10. A differential pulse polarography investigation of a plutonium-containing solution at pH 1 indicated that the predominant species present were Pu III and/ or Pu IV. Actinide-containing solutions were subsequently passed through columns filled with natural rock salt. Results indicated that at least two types of actinide species were present; one was strongly retarded and the other migrated with the solvent front. The relative proportions of these depended on the pH of the system and the solution composition. For actinide mixtures results indicate that americium is more effectively retained than plutonium. The presence of lanthanides such as cerium were found to affect dramatically migration properties.

The influence on actinide solubilities of candidate materials for use in repositories for low and intermediate level nuclear wastes

The solubilities of uranium, plutonium, americium and neptunium have been measured under oxidizing conditions as a background to a programme of leach testing of intermediate level wastes under simulated repository conditions. The waters used were prepared by prior equilibration with candidate matrices and backfills, which included two cements, a polyester resin and bentonite. The solubilities measured were generally low (10 −10-10 −5M). The experimental values were compared with those calculated using a simple model. Good agreement was achieved with americium and plutonium; however, the behaviour of uranium and neptunium proved difficult to predict.

Substitutional solid solutions from heterotypic lanthanide dititanates Ln 2- xLn′ xTi 2O 7

The formation of substitutional solid solutions of dititanates containing two lanthanide ions from heterotypic mixtures, monoclinic and cubic, at temperatures up to 1500 °C was studied by Raman spectroscopy and X-ray diffractometry. All the results were fitted to a stability diagram by means of two parameters: the percentage difference δ in the ionic radii of each lanthanide pair and the unit cell volume V of the resulting crystal. The boundary between cubic phases and the region of restricted miscibility provides information on the solubility of the monoclinic phase in cubic lanthanide dititanates. This can be used to predict the solubility of tervalent actinides in such compounds which is relevant to the chemistry of nuclear waste immobilization.

The theoretical derivation of solubilities of long-lived radionuclides in disposal

Calculations have been carried out on the solubilities of several long-lived radionuclides under disposal conditions, including several geologies. The effects of the matrix for immobilisation on pH and of the corroding canister on Eh are taken into account. An approach involving Pourbaix diagrams has been used for the fission and activation products, but a more complex model has had to be employed with the actinides. Comparison with assumed upper concentrations in drinking water suggests that, of the radionuclides studied, plutonium, neptunium, americium, caesium-135, iodine-129 and technetium-99 merit high priority in source term studies. Extensions are suggested to the model used to calculate actinide solubilities.

The Solubility of Actinides in the Near-Field

AbstractThe solubility of some of the radiologically important actinide elements have been determined in a water whose chemistry was representative of that believed to exist in potential repositories in the U.K. The solubilities of the actinides have been determined as a function of the pH of the water. The results of these experiments have been compared with the results of predictions made using the PHREEQE geochemical modelling code in order to test and validate the data base used in the model. In the light of these comparisons, suggestions are made for alternative values for the thermodynamic data and for further studies.

Natural Analogues—A Way to Increase Confidence in Predictions of Long-Term Performance of Radioactive Waste Disposal

ABSTRACTNo exact analogue to a radioactive waste disposal system exists but by studying natural analogues of the most important components or subsystems of a waste repository, confidence in long-term predictions can be increased. The U.S. Nuclear Regulatory Commission (NRC) is supporting research on igneous intrusions into proposed repository-type host rock (basalt, tuff, crystalline hard-rock, and salt), on uranium ore body analogues, on actinide solubility as a function of natural complexing agents in closed basin lakes and on field migration of radionuclide species at low-level waste and uranium mill tailings disposal sites. The NRC is studying analogues to ensure that important parameters which might be overlooked in laboratory tests and in modeling because of problems with scaling time, space and geologic complexity can be identified and their importance bounded.

Evaluation of Solubility and Speciation of Actinides in Natural Groundwaters

Groundwater circulation is in the long term probably the most important mechanism for the transport of radionuclides from deep underground repositories to the biosphere. The fate of leached long-lived nuclides, such as the actinides, is of particular interest to repository safety analysis. Both experimental and theoretical studies should lead to a comprehensive description of their behaviour in the geosphere.

Adsorption of Co and selected actinides by Mn and Fe oxides in soils and sediments

Iron and Mn oxides and associated radionuclides in soils and sediments from the radioactive waste burial grounds at Oak Ridge National Laboratory have been selectively extracted using wet chemical techniques. Product-moment-correlation analyses have demonstrated that 60Co and various actinides, principally 244Cm, 241Am and 238Pu are dominantly associated with Mn oxides. Correlation coefficients between these radionuclides and Fe oxides and organic C are generally very low. The important role of Mn oxides in radionuclide adsorption is attributed to their unique surface and colloidal properties. The data illustrate the importance of the Mn oxide component of soils and sediments in controlling transition metal and actinide solubility. These results suggest two major implications for the disposal of radioactive waste. First, in order to minimize future 60Co and actinide mobilization from disposal sites, a chemical environment in which Mn oxides are least soluble should be maintained. Second, the liberal use of Mn oxides in waste management operations might improve long-term retention of these radionuclides. Deep-sea Mn modules, which may in the future be mined for their trace metal contents, could serve as a ready supply of Mn oxide for waste disposal applications.