The relative leaching behavior of seven alkali borosilicate glasses considered for immobilization of high level radioactive wastes was compared using a static 90°C leach test similar to MCC-l. All compositions (Table I) were compatible with the low melting temperature, 1150*C, viscosity and other process variables required for the French AVM process. The quantity of simulated waste products was from 10.9-15.9 weight %.

The corrosion stability of six simulated HLW forms proposed by several European countries (five borosilicate glasses and one borosilicate glass ceramic) in contact with different media relating to the clay disposal has been investigated for periods up to 80 days and at a surface area to solution volume ratio of 1 cm−1 under non de-aerated conditions. In the reference medium distilled water the corrosion stabilities are largely determined by saturation effects for elements such as Si, Ca, Mg, Sr, Fe and U. Si saturation is found to be enhanced by a large Al2O3 concentration in the glass. In the clay-water mixture these saturation effects are much less predominant. Wet clay attacks the waste forms faster than the clay-water mixture, although the corrosion rates tend to decrease with time in both clay media. In general, no simple Arrhenius-type temnnerature dependence for the corrosion behaviour is found for the interval between 40 and 200°C.

Part of nuclear fuel cycle waste is highly dangerous, and must be safely isolated from people. Although the site of the final waste disposal must be the main safety barrier, the form of the waste and its properties are also important considerations.

The compositions of borosilicate glasses used to solidify fission product solutions are selected for their low leach rates. Fracturing of the glass blocks, caused primarily by thermal and residual stresses during cooling, increases the potential leaching surface area and the number of small particles, both of which are undesirable since long term container failure must be considered. It is therefore important to know the state of fracture of the glass occurring at different stages of its thermal history. A theoretical appraisal of the fracture state of the glass is presented, complimented by an experimental study using industrial scale glass blocks under realistic thermal conditions.

The crystallization of fission product glasses is generally caused by maintaining these glasses at high temperatures (600° to 900°C) for several days or weeks in order to evaluate the effects of crystallization on leach resistance and mechanical properties.

In order to assess the longterm chemical and physical stability of high level waste glasses it is important to know the amounts of crystalline phases present.

In principal all information about the crystalline state can be derived from its X-ray powder diffraction pattern,provided the spectral resolution is sufficiently high. Powderdiffraction techniques are the only means available to determine the concentration of crystalline phases in the presence of equivalent amorphous components. In addition it permits the determination of specific crystalline phases in a multicomponent system.

A recent study concluded that addition of Fe203 to a soda borosilicate nuclear waste glass may significantly reduce damage by water attack due to formation of a Fe-rich film on the glass surface. However, differences in SiO2, B203, CaO, and concentration of fission products in previous glass compositions make it impossible to ascribe the improved leach resistance solely to Fe203 content. In the present work, leaching behavior of two glasses are compared which differ only by the substitution of Fe203 for some of the ZnO in the glass. Both glass compositions, Table 1, are compatible with the French AVM process and contain 9% (by weight) of simulated waste products characteristic of the Swedish nulcear waste program.

Knowledge about the oxidation state of molybdenum in glasses and glass ceramics which contain nuclear waste is important because of the danger of formation of water-soluble Cs-molybdate with Cs-137. In a previous paper , the behaviour of molybdenum in glasses with and without fission products was investigated with the aid of ESR and optical spectroscopy.

In the vitrification of HLW the aim is to create a product, the stability (thermal, mechanical) of which is, to the greatest possible degree, independent of fluctuations in its composition. Process control during vitrification of HLLW is necessary in order to assure the quality of the final glass product. The chemical processes are controlled by quantitative analysis of all material streams entering and leaving the melting device. There are six material streams to consider: frit input, waste input, corrosion products (generated by the vitrification process), off-gas, recycled off-gas, glass output. Additionally, the capacity of the glass for buffering short-term fluctuations in the incoming streams must be known. Sufficient sampling during non-radioactive vitrification in addition to materials tests can provide the necessary data basis. The results are combined in a materials balance.

In October 1981 SYNROC-D was selected as the reference alternate waste form to borosilicate glass for immobilization of defense wastes. A total of eight candidate waste forms competed in this selection process and the decision of which alternate waste form to choose was based primarily on performance properties.

The Synroc concept was introduced by Ringwood in 19781 and, as is well known, the material now proposed for the disposal of hiah-level civilian nuclear waste consists primarily of hollandite (BaAl2 Ti6016), perovskite (CaTiO3) and zirconolite (CaZrTi207).

Glass ceramics and ceramics based on the mineral sphene (CaTiSiO5) are being developed to host the wastes arising from possible future CANDU* fuel reprocessing. Results from leaching tests in deionized water and in synthetic groundwater indicate that these materials are highly durable. Secondary Ion Mass Spectrometry (SIMS) depth profiling of leached specimens suggests that leaching in the glass ceramics is predominantly confined to the glass phase. The high ionic strength and composition of the groundwater have a significant passivating effect on leaching and surface alteration phenomena, and encourage the precipitation of new phases on the ceramic surface. Leaching results, scanning electron microscope (SEM) observations and SIMS depth profile measurements are compared and discussed.

The results of recent measurements of optical absorption, etching rate and transmission electron microscopy in pure silica and borosilicate glasses are reported and discussed. At dose saturation conditions, the dependence of the optical density associated with the production of single atomic defects from the mass of the impinging particle shows a marked saturation at masses ≥ 20 amu. The corresponding etching rates increase by about 4 times with respect to unirradiated samples. In borosilicate glasses, the temperature dependence of the threshold dose rate of electrons to initiate the nucleation of bubbles shows a marked increase from about 300° to 600° K.

Ion implantation into simulated nuclear waste glasses is a rapid means of producing near-surface energy deposition similar to that produced by α-recoil nuclei after long storage times (typically 103 – 106 years). For example, Dran, Maurette, and Petit1 used 200 keV Pb-ion implantations in glass at a fluence of 5 × 1012/cm2 to produce surface damage. This fluence is equivalent to approximately 2 × 1018 alpha-decays/cm3 which corresponds to approximately 106 years storage for glass containing Savannah River Plant (SRP) defense high-level waste (DHLW). These authors1 found that this fluence value corresponded to a critical fluence (Φc) for enhanced etching (a factor of 20 increase as inferred by step-height changes) for several silicate glasses when etched in a NaCl solution at 100°C. This critical fluence value also corresponds very well with the fluence at which significant overlap of individual ion tracks occurs.

Radiation effects studies in both glass and glass ceramic nuclear waste forms have identified a rare-earth titanate phase of the general formula (RE) 2Ti207 which is capable of acting as a host phase for actinides.1,2 Ringwood and co-workers3 have also proposed a structurally similar phase, zirconolite (CaZrTi2 07), as one of the primary host phases in the SYNROC waste form. Data from these and other previous studies, as well as mineralogical information available on these titanate phases, have not provided an unambiguous interpretation of the effects of radiation damage relative to nuclear waste forms. This paper reports new laboratory data concerning radiation damage effects in both of these phases.

The long-term stability of nuclear waste forms is an important consideration in their selection for safe disposal of radioactive waste. Stability against long-term radiation damage is particularly difficult to assess by short-term laboratory experiments. Much of the displacement damage in high-level waste forms will be generated by heavy recoil nuclei emitted during the α-decay process of long-lived actinide elements. Hence, an accelerated aging test which reliably simulates the α-recoil damage accumulated during thousands of years of storage is desirable. One recent approach to this simulation is to implant the waste form with heavy Pb-ions.I- 6 If the validity of this approach is to be fully assessed, two important questions which have not yet been investigated must be answered.(1) Is the structural damage, including cumulative effects, similar for irradiation by Pb-ions and a-recoil nuclei in a given material? (2) Is the dose-dependence of the accumulated damage similar? The purpose of this investigation was to assess the extent of these similarities in selected materials. We utilized transmission electron microscopy (TEM) to characterize the radiation damage and measure its dose-dependence.

The variation in the leachability of naturally occurring zircons (ZrSiO4) as a function of total alpha dose has been measured at 87°C. For calculated doses in the range of 1016 to 1018 alphas/gm there is an increase of one order of magnitude in the weight percent of zircon that is dissolved and an increase in the leach rate of the zircon 2.9 × 10−8 to 2.3 × 10×7 gm/cm2 day. Totally metamict specimens (dose ≥ 1019 alphas/gm) have leach rates as high as 1.8 × 10−6 gm/cm2 day.

It has been proposed recently that high-level radioactive wastes from spent nuclear reactor fuels be converted into a crystalline phosphate form prior to their disposal in geological repositories. This approach is based on the known durability in nature of the mineral monazite, (Ce,La)P0, which contains appreciable concentrations of thorium and uranium. Of significance is the demonstrated resistance of monazite to excessive metamictization, in spite of the fact that the mineral has been subjected since its formation to radiation from the radioactive disintegration of U, Th and their many intermediate decay products. Nevertheless, radiation effects on the leaching behavior of the material in possible groundwater environments must be studied in the development and testing of monazite-like phases for the long-term disposal of fission product and actinide wastes.