The behaviour of the platinum metals in nuclear waste glass produced in a lab scale liquid fed ceramic melter was investigated. The platinum metals form two kinds of metallic conductive particles in the glass melt: RuO2 (preferentially needle shaped) and a Pd—Rh—Te alloy (tear—shaped). These phases decrease the electrical resistivity of the glass melt. Resistivities were measured at 950º and 1150 ºC. Powdered glass was used as starting material. For the same temperatures resistivities were calculated applying an equation derived for an embedment structure of highly conductive particles in a matrix of low conductivity. To match calculated and measuredresistivities one must assume needle—shaped particles. This gives evidence that mainly RuO2 contributes to the observed decrease in resistivity. If the measurements are started with glass blocks instead of glass powder, remarkably lower resistivities are measured in the melt. A comparison of both results indicates a partial destruction of particles or particle contacts by grinding. Thus, resistivities measured with ground samples are not representative of those effective in a melter.

The investigation of solid glass samples yielded the following results: Within a few hours, Pd—Rh—Te and RuO2 particles form fine sediments in the melter. The development of stable coarse sediments needs some 1000 h of melter operation. The Pd—Rh—Te alloy particles reach a mean diameter of 5 μm in the fine sediment and of 20 μm in the coarse one. With time, the alloy dissolves more Rhand Te. Finally, the alloy separates into two phases. RuO2 dissolves Cr and Rh.

The diffusion of Te and Pd was measured in the German waste glass VG 98/012 usingion implantation to introduce these elements, and the Rutherford backscattering (RBS) method to measure their diffusion. Both elements tend to precipitate during manufacture ofnuclear waste glasses, and the Te-content of the precipitates depends on the annealing time. Diffusion of these elements is a basic property of the separation phenomenon. Diffusion of both elements was found to be very slow. Te-diffusion coefficients were in the range of 10-17 to 10-15 cm2s-1 between 540 and 590 αC. Surface evaporation of Te was observed at all heat treatments. Pd diffusion was not observed in this study, most likely due to limited solubility of Pd and formation of metal precipitates.

Volatility of 137Cs and 106Ru at 600ºC over high-level waste glass in a canister is numerically analyzed. The time-dependence of normalized concentrations of each radionuclide is traced by a differential equation based on a simple model consisting of competitive forward and backward steps. It is assumed that, as time proceeds, a surface layer forms on the glass and serves as a protective barrier against volatilization. It is found that the forward rate of volatilization for 137Cs is about 140 times greater than that of 106Ru at 600ºC. The effects of the glass surface area and plenum volume on the concentration excursion inside the canister are closely estimated by the differential equation.

The microstructural changes of a simulated waste glass irradiated with doping of 238pu and 244Cm were observed by use of a preshadowed carbon replica technique in combination with scanning electron microscopy (SEM). The irradiated glass was annealed and its microstructural changes after annealing were observed by use of the sametechnique.

In the glass irradiated at a dose of 2.75x1025 α-decays/m3, bubbles with a radius from 0.15μrm to 0.35μm were observed. The average bubble radius and the bubble density were 0.23μm and 1x1017 bubbles/m3, respectively. Using these observed values, the volume change of the glass resulting from the bubble formation was estimated to be + 0.51 %. This valuoof + 0.51 % was close to the volume change of the irradiated glass measured in our previous study, which suggests a large portion of the volume change by α-irradiation results from bubble formation. In the glass annealed after irradiation the average bubbleradius was observed to decrease with annealing time. The bubble radius as a function of annealing time was analyzed on the basis of the helium diffusion model with two chemicalprocesses, i.e. trapping at bubbles and re-solution from bubbles into glass matrix. The values of the diffusion coefficient of helium, the trapping parameter and the re-solution parameter, which had been obtained experimentally in our previous study on the helium release, were applied to the calculation. The calculated curve was in good agreement with the observed data.

An on-line method is described for the near-continuous monitoring of the composition of a molten radioactive waste glass or, alternatively, for signaling a deviation from the target composition of a waste glass. The principle of this method, proposed by A. Schneider in 1986, is founded on the relation between two specific physical properties and composition in a ternary system. Most glasses currently considered as waste forms can be represented as pseudo-ternary systems. The pair of properties especially well suited for this purpose are the viscosity and density of the molten glass. A novel viscometry method was developed which uses the remotely determined rise velocity of carefully metered gas bubbles. The monitoring method was tested successfully with simulated Savannah River waste glasses. An integrated probe was conceived for a Joule-heated melter for the on-line determination of viscosity, temperature, density, and liquid level. A computer program calculates the glass composition from the measured data, using information from a previously developed data base.

Aluminum silicate based ceramics were studied as matrix materials for the immobilizationof real TRU—wastes, as well as simulated high-level and TRU—waste mixtures. As observed by XRD and electron microscopy, isolation of radionuclides is achieved either by encapsulation of radiophases by aluminumsilicate matrix phases and by incorporation into host phases in the matrix. TRU—leach rates were found to be on the order of 5*10–3 and 1*10–2 gm–2 d–1 at room temperature after one year in MgCl2 —rich salt solution. By doping the ceramic matrix with 20 weight percent PuO2 an accumulated α-dose of 8.33*109 Gy (1*1019 α-decays*g–1) was reached after 3 years. The microstructure and the lattice constants ofthe matrix phases were not affected by a-irradiation.

Fission products and α-emitters from reprocessing plants ( HAW ) are incorporatedinto a ceramic matrix consisting of 90 w/o α-AI2O3 The amount of HAW in the product is 10 w/o. The matrix is produced by a sol-gel technique using a boehmite powder and a SiO2 sol as starting materials. To reduce the sintering temperature seeding with α-AI2O3 particlesis applied. The first step of the process is mixing of the starting materials with the nitric acid containing HAW. This results in the production of a gel. The gel is extruded and subsequently dried at 105ºC. Calcination is carried out in a hydrogen containing atmosphere. Finally, the material is sintered at 1300ºC. During heat treatment only minor volatilization is observed. Even cesium is almost completely retained sinceit forms an aluminosilicate (Cs2O. xSiO2. yAI2O3; feldspar type). The waste product has a total leach rate of 4 * 10-7g * cm-2 d-1 at 97ºC as determined in a Soxhlet apparatus. Under the same conditions, specific leaching rates of radiologically important nuclides, such as cesium and strontium, are approximately 10-6 g * cm-2 d-1. The process has the advantage that commercially available starting materialscan be used. In addition, only a moderate sintering temperature is required and pressingbefore or during sintering is not necessary.

Perovskite is the least durable of the resistate minerals comprising Synroc-C and it is desirable to reduce its abundance in Synroc. Kinetic limitations and competition with Csapparently affect the incorporation of Sr into hollandite during hot-pressing at 1200ºC/20 MPa so that ∼ 10% of perovskite (a value below the percolation limit) is probably an optimum target. Zirconolite-rich Synroc formulations have been prepared for actinide-rich wastes. Background XRD and TEM studies have also been performed to study the crystal-chemical behaviour of Nd (a simulant of trivalent actinides) in zirconolite. Either rare-earth compensated perovskite or freudenbergite in Synroc can evidently be used to immobilise Na-bearing HLW.

Synroc formulations containing 10 wt% simulated PW-4b type HLW in combination with a single transuranic dopant (Np, Pu, Am or Cm) were fabricated and characterised using alpha-track methods, SEM, AEM and dissolution testing. The Pu, Am and Cm-doped samples were calcined in 3.5%H2/96.5%N2. One of the Np-doped Synrocs was calcined in an atmosphere of pure H2, whilst another Np-doped Synroc was calcined in a mixture of 3.5%H2/96.5%N2. All the different formulations displayed microstructures similar to those of Synroc containing actinide-free simulated HLW. AEM derived phase compositions and phase distributions are given. All the Synrocs showed good durability when exposed to leachants (including doubly deionised water, and pH- and Eh-buffered solutions with pHs between 2 to 10). There were no obvious differences between the two Np-doped samples. The improved appearance and behaviour of allthe Synrocs in this study compared with similar formulations described by other authors is attributed to improved mixing techniques during fabrication.

Results of further studies on phase equilibria in system CaO—ZrO2—TiO2 along the “pyrochlore” line ( Ca Zrx Ti3-x 07 ) are reported concerning the possible substitution of cerium for either calcium or zirconium in nominal “Ca Zr2 Ti 07”. The dominant tetragonal phase in this composition was identified as calzirtite. The structural role of zirconium and its tendency towards 7—coordination environments are discussed. Possible implications on high—level nuclear waste immobilizers like SYNROC are brieflyconsidered.

We have found that lanthanum zirconate (La2Zr2O7) with a pyrochlore structure can be used as a superior host phase for actinide and lantanide elements contained in high-level nuclear waste. The solubility of Ce, Nd and Sr in La2Zr2O7 is determined because Ce and Nd may simulate the segregation of actinide and lanthanide elements at the grain boundaries of polycrystalline La2Zr2O7. The results show that La2Zr2O7 is a promising material for confining actinide and lanthanide elements in its crystal structure.

This study deals with the incorporation of Eu3+ ions in aluminates with magnetoplumbite (MP) or mixed β alumina - MP type structure. The synthesis procedure is the coprecipitation from a nitric solution. Partial substitution of Eu3+ byLa3+ increases the thermal stability range of the phases. In these structuresEu3+ occupies several sites and in the case of Na+ containing phases, Eu3+ and Na+ coexist at some mirror planes. Some results concerning Cs+ ion immobilization in mixed β - MP structures, by ionic exchange in molten salt are also reported.

This paper reports the preparation and characterization of a series of oxides containing3+ or 4+ lanthanide (M = Ce, Pr, or Tb) ions, with different ionic sizes and varying M4+/M3+ reduction potentials, in nearly cubic coordination. The objective of the study was to demonstrate how oxidation-reduction characteristics and ionic-size trends explain the properties of these oxides and to compare the oxidation-reduction stability of M3+ and M4+ lanthanide ions in high (CN 7 or 8) coordination in fluorite-related oxides versus low (CN 6) coordination in perovskite oxies. Efficient preparative methods are reported, as well as powder diffraction and thermogravimetric measurements for oxides CaMTi2O7-x and CaMZr2O7-x. These oxides were characterized by X-ray powder diffraction and by thermogravimetric analysis. CaCeTi2O7 is a pyrochlore, a = 10.142(4) Å with Ce4+ much more easily reducible than in the perovskite BaCeO3. By contrast, a preparation with the stoichiometry “CaPrTi2O7-x” is a two-phase mixture of perovskite CaTiO3 and a presumably Pr3+-rich pyrochlore Pr2Ti2O7(?). CaTbTi2O7-x appears to be a Tb3+ pyrochlore, a = 10.149(2) Å. CaCeZr2O7 is a pyrochlore, a = 10.524(1) sÅ. A preparation of “CaPrZr2O7-x” also appeared to yield a two-phase mixture, perovskite CaZrO3 and pyrochlore Pr2Zr2O7. In this paper, the structures, f-element ion sites, and M(IV)- MM1I stability trends in the CaMTi2O7-x and CaMZr2O7-x oxides are compared with the structural and stability trends in the perovskites BaMO3 which have M4+ ions in sixfold (tilted octahedra) coordination.

This paper reviews the literature on rare earth(III) and actinide(III) hydroxide thermodynamics, in particular the determination of their enthalpies of formation at 25 ºC. The hydroxide unit-cell volumes, lanthanide/actinide ion sizes, and solid-solution stability trends have been correlated with a generalized acid-base strength model for oxides to estimate properties for heterogeneous equilibria that are relevant to nuclear waste modeling and to characterization of potential actinide environmental interactions. Enthalpies of formation and solubility-product constants of actinide(III) hydroxidesare estimated

Solids obtained from laboratory solubility experiments in two different groundwaters from the Yucca Mountain region, Nevada, are described. The solubility study provided limiting solubility concentrations for neptunium(V), plutonium(IV), and americium(III) in groundwaters from Wells J-13 and UE-25p#l. The solubility-controlling solids are compared to relevant radionuclide compounds that are reported in the literature. The preparations and some characteristics of published solids that possibly may form in actinide-groundwater systems are described. The solids formed in the experiments are sodium neptunium(V) carbonates, polymeric Pu(IV) that contained small amounts of carbonate, and hexagonal or orthorhombic americium(III) hydroxycarbonates.

UK Nirex Ltd are seeking to develop a deep underground repository for the disposal of solid ILW and LLW in the UK. The formation of water-soluble complexants from the degradation of the solid organic polymers and cellulosic materials present in such wastes may influence the solubility and sorption properties of a number of radioelements. A number of materials have been chemically degraded under alkaline, anaerobic conditions representative of the near field of the repository. Measurements of plutonium solubility in the leachates obtained show that the degradation of cellulose is of particular concern; the effects on the solubility can be several orders of magnitude greater than those of other organic materials. Products formed from the degradation of ion-exchange resins were found to have little effect. Solubilities of a number of other radioelements in aerobically degraded cellulose leachates are also given. The effect of reducing the ratio of celluloseto cement in the degradation experiments results in a decrease in plutonium solubility in the corresponding leachate. Several of the likely degradation products of cellulose have been individually synthesised including isosaccharinic acid, a key degradation product of cellulose. Enhanced plutonium solubilities have been measured in 10-3M solutions of these compounds. Sorption of plutonium onto cement from leachates obtained by the degradation of 10% cellulose in cement is reduced by about two orders of magnitude. Interaction with cement removes significant concentrations of complexants from solution. Thermodynamic modelling studies show that hydroxyl groups present in the compounds which result from the alkaline degradation of cellulose can strongly complex with plutonium.

The experimental investigation and modelling of radionuclides sorption processes is an essential part of safety analyses of an underground repository of radioactive wastes in geological formations. Physico—chemical phenomena at oxide/solution interfaces have been studied in the case of the system constituted by amorphous silica and americium (III) in the presence of humic substances as organic complexing agents. The retention behaviour of Am(III) onto silica appears dependent on pH, ionic strength and in particular humic concentration. At a ionic strength of 0.1 M, the presence of low concentration of humic acids (1 mg/l) enhances the Am(III) retention in the pH range 3-5.5, whereas at higher pH, its uptake is largely decreased. At lower ionic strength (0.001 M), Am(III) retention on silica is highly lowered, in the pH range 3-7 in the presence of humic acids. Reversibility experiments show the desorption of americium in the absence of humic substances, and also the desorption of the humics on the silica with respect of the pH. The surface complexation model has been applied to our results in the different systems and apparent surface complexation constants have been determined.

The speciation of UO22+ and UO22+/ TBP mixtures has been investigated in solution and intercalated with the reference smectite clay SAz-1 using x-ray absorption, Raman, andluminescence spectroscopies. Neither aquated UO22+ nor its TBP complex undergoes any detectable changes in uranium oxidation state on intercalation. Further, at the pH values employed in this work, there is no evidence for hydrolysis of the uranium species to generate dimeric or higher order uranium oligomers. However, we do find indications that the structures of the solution complexes are altered on intercalation, particularly for the UO22+TBP system and for more dilute UO22+/aqueous systems. In addition, several lines of evidence suggest that, at the loading levels used in this study, the uranyl species is interacting with two or more spectroscopically distinguishable sites on SAz-1.

This report covers the investigation of mechanisms for the retardation of U(VI) migration. It describes (1) the measurement of the solubility of UO22+as afunction of pH and equilibration time, (2) the calculation of solubility using solution thermochemical data, (3) the measurement of the sorption of UO22+by silica as a function of pH and supporting electrolyte concentration, and (4) attempts to simulate the uranium sorption isotherms in terms of the U solution thermochemical data, measured properties of the silica and proposed sorption reaction mechanisms.

The solubility products of uranyl hydroxide and uranyl carbonate and the formation constants of the complexes present in the solution have been determined at 25ºC and I=0.1 mol dm−3 by potentiometry and solubility experiments. The uranyl complexes U02(OH)2' (UO2)2(OH)22+,(UO2)3(OH)5+, UO2CO3', UO2(CO3)22− and UO2(CO3)34− were identified. For the system uranyl-hydroxide the results were verified by analysis of conductance measurements. A similar investigation was carriedout for thorium. Using Pu(IV)-chloride complexes as an example the potential of differential pulse voltarnmetry is demonstrated for the determination of small formation constants.