The transport of radionuclides in indigenous rock is greatly affected by the sorption of cations in the porous rock matrix. For the determination of sorption coefficients, batch experiments have traditionally been used. For these experiments, the rock sample is crushed into fine particles to reduce the experimental time. However, this procedure increases the specific surface area of the sample and the new surfaces created could have different sorption qualities than the naturally occurring surfaces, which may impair the results of sorption coefficient determinations. A new method for determining sorption coefficients in intact rock is being developed, using electromigration as a means to speed up the transport process, thereby allowing for faster equilibration between the rock sample and the tracer solution. Here, we report results from preliminary experiments, using cesium as a sorbing tracer, showing a consistent difference between sorption coefficients obtained using electromigration methods on intact rock samples and traditional batch experiments on crushed samples.

Many countries have recognized vitrification as the suitable solidification method for conditioning high level radioactive waste before they are isolated from biosphere in geologic disposal. This paper studies the leaching behavior of a simulated HLW-glass (90Nd/10) under low oxygen repository condition during 1.5 years. The leaching results showed that the glass was corroded less under low oxygen condition than under ambient atmosphere. XRD analyses showed some little diffraction peak appeared on the leached glass surface at 150°C. At the same time, no diffraction peak appeared on the leached glass surface at 90°C. The secondary products formed on the surface of the leached glass contained mainly Si, Al, Ca and Fe.

Over the past decade or so there has been an explosion in the number of sorption modelling approaches and applications of sorption modelling for understanding and predicting solute transport in natural systems. The most widely used and simplest of all models, however, is that employing a constant distribution coefficient (Kd) relating the sorbed concentration of a solute on a mineral surface and its aqueous concentration.

There are a number of reasons why a constant partitioning coefficient is attractive to environmental modellers for predicting radionuclide retardation, and in spite of all the shortcomings and pitfalls associated with such an approach, it remains the leitmotif of most performance assessment transport modelling.

This paper examines the scientific basis underpinning the Kd-approach and its broad defensibility in a performance assessment framework. It also examines sources of epistemic and aleatory uncertainty that undermine confidence in Kd-values reported in the open literature. The paper focuses particularly upon the use of so-called “generic” data for generalised rock types that may not necessarily capture the full material property characteristics of site-specific materials.

From the examination of recent literature data, it appears that there are still a number of outstanding issues concerning interpretation of experimental laboratory data that need to be considered in greater detail before concluding that the recommended values used in performance assessments are indeed conservative.

In a highly confined medium corresponding to geological repository conditions, the alteration rate of the French SON68 inactive nuclear reference glass drops by about four orders of magnitude below the initial rate. However, extended experiments lasting months or years provide evidence of a virtually constant or slowly decreasing residual alteration rate. Although very low, this rate could account for most of the altered glass thickness after 10 000 years. Experiments at high temperatures and especially high glass-surface-area-to-solution-volume ratios were performed to reveal and quantify the predominant mechanisms underlying the residual rate. The authors describe the characterization of the solution chemistry, the crystallized secondary phases, and the amorphous gel observed after alteration of the French SON68 inactive reference glass, and discuss their implications in terms of long-term behavior modeling. A slow diffusion mechanism is identified in the solid, involving alkalis in particular but also boron. This mechanism results in higher concentrations in solution that affect the system chemistry, not only by slightly modifying the pH and element speciation in solution (e.g. silicon), but also by inducing the precipitation of new crystallized secondary phases that can consume glass constituent elements in the same way as simple solution renewal. Diffusion and the precipitation of secondary phases are two mechanisms to be considered to account for the residual rate.

An archaeological glass initially fractured and altered for 1800 years in a marine environment is now being examined by the CEA because of its strong morphological similarity to the nuclear glasses used for immobilization of long-lived radionuclides (i.e. the presence of fractures and cracks formed during cooling, which significantly increase the surface area accessible to water). The issue concerns glass alteration by water, and in particular the different behavior of the external surfaces in contact with a solution highly renewed and the internal surfaces, which constitute a much more confined medium.

The preliminary results of this study are discussed. The cracks in the archaeological glass have been filled by crystallized alteration products formed jointly by elements from the glass and elements dissolved in seawater. The glass is distinctly less altered (by a factor of 10 to 100) on the internal surfaces generated by the cracks than on the external surfaces. The forward glass dissolution rate was measured at different temperatures on pristine glass samples and under conditions that allowed us to estimate the dissolution rate of the external surfaces under realistic conditions at about 200 μm in 1800 years. The implications of this study are then discussed.

Solidification of molten ceramics during molten state processing of High-Level Waste (HLW) was studied to investigate a possibility of a simpler and flexible process for immobilisation of HLW. Simulated HLW was heated together with different reducing agents (TiN with AlN or TiO2) either at 1673 K or 1873K for 1 to 3 hours under Ar atmosphere. Cooling rates of 5 K/min and 10 K/min did not result in any obvious differences in the products obtained under the standard condition used in the present study. The composition and quantity of the reducing agent had a significant effect on the melting and solidifying behaviour of the simulated HLW. The results suggest the possible further studies to improve the solidification process and the final product.

Sulphur is present in several kinds of nuclear waste destined to confinement in a glass matrix.. These species are hardly incorporated in borosilicate glasses, due to the poor miscibility of molten sulphates and glass melt at high temperature. This point constitutes one of the main technological difficulties in the current fabrication of nuclear glasses. Several studies carried out in the last two decades indicate that the liquid state miscibility of sulphates and glass melt, determining the waste incorporation in the final glass, is dependent on the composition of the glass and on the temperature and duration of its preparation. In particular, the ratios K = [SiO2] / [B2O3] and R = [Na2O] / [B2O3] (where [.] indicates the mol %), as well as the presence of V2O5 in the glass matrix play an important role in the determination of both kinetics and thermodynamics of sulphates incorporation in the vitreous matrix. In this work a study on the physico-chemical effects of the ratio R, the melt reactivity, the type of sulphate added, and the presence of V2O5 on the solubility of sulphates in borosilico-vanadate glasses is presented.

Glass samples were prepared at the laboratory scale (up to 50-100 g) by melting oxide and sulphate powders in Pt/Au or Pt/Rh crucibles at a fixed viscosity level of 100 Po in air. Several glass specimens were powdered and re-melted at 1200°C in the presence of sulphates in order to simulate conditions as close as possible to those of an industrial melter. XRF and ICP AES chemical analysis, SEM EDS and Raman spectroscopy were employed to characterise the fabricated samples. Raman spectra, in particular, reveal the structural modifications that condition the sulphate incorporation in the glass. A critical review of the obtained results is compared with the literature data, to yield a more systematic description of the mentioned factors effects on sulphate behaviour in the borosilico-vanadate / system.

Ceramics in systems with rare earth (REE = Y, Sm, Gd), manganese, and titanium oxides are considered as potential matrices for immobilization of REE- and actinide-containing radioactive wastes. Ceramics of this type were synthesized from oxide mixtures by a cold pressing and sintering method at temperatures of 1300-1400 °C as well as plasma heating. Major phases in the ceramics were found to be the phases of the pyrochlore-murataite polysomatic series with three- (3C), five- (5C), and seven-fold (7C) elementary fluorite unit cell. Perovskite/pyrophanite, hibonite/loveringite as well as garnet in the samples contained extra Fe and Al oxides were present as minor phases. Formation of the murataite, to be exact – structures built from murataite and pyrochlore modules, is more characteristic of the Y-bearing systems rather than the systems with heavier lanthanides. Increase of ionic radius of the REE trends towards formation of the pyrochlore structure phases whereas occurrence of large-size cations (La) stimulates formation of the perovskite structure phase. In the Y-bearing system murataite formation takes place at relative low Y2O3 concentrations (<15 mole %) and TiO2 content as high as ∼60 mole % and MnOx – 15-20 mole %. Higher Y2O3 concentrations and Gd2O3 substitution for Y2O3 yields preferably the pyrochlore phase or may favor formation of the phase with the 7C polytype.