Hirotaka Furuya

Corrosion behavior of a powdered simulated nuclear waste glass: A corrosion model including diffusion process

Abstract Static corrosion tests were performed with a powdered simulated waste glass in deionized water at 90°C for periods of up to 130 days. It was observed that normalized elemental mass loss (NL) values for soluble elements (Li, B, Na and Mo) were larger than those for Si by a factor of three and continued to increase after saturation of Si. A corrosion model (diffusion-combined model), where a diffusion model is combined with a dissolution/precipitation model (reaction path model), was developed and applied to the analysis of experimental results. In the diffusion-combined model, it is assumed that less-soluble elements dissolve into the solution congruently with the silica glass matrix (glass matrix dissolution). On the other hand, it is assumed that soluble elements diffuse through the glass to the surface and dissolve into the solution, in addition to the glass matrix dissolution. The diffusion-combined model can explain the experimental results well, and it is found that the diffusion coefficient is the most effective parameter determining the corrosion behavior.

Volumetric change of simulated radioactive waste glasses irradiated by the 10B(n, α)7Li Reaction as simulation of actinide irradiation

Abstract The density change of simulated radioactive waste glasses irradiated by the 10 B(n,α) 7 Li reaction was determined by a sink-float method as a function of irradiation exposure. Simulated waste glasses P0500, P0798 and GP98/12 swelled, while P0504 shrinked. The magnitude of the density change was less than 0.6% up to a fluence of 6.6 × 10 25 reactions/m 3 , which corresponds to the cumulative irradiation during a few tens of thousand years after disposal of the waste glass from the spent fuel irradiated up to 33000 MWD/MTU. The processes which play an important role on the density change have not been clarified, but it is likely that one of the processes is helium bubble formation which was clarified by a carbon replica technique, in association with transmission electron microscopy.

Oxidation kinetics of Zircaloy-2 between 450°C and 600°C in oxidizing atmosphere

Abstract The oxidation kinetics of Zircaloy-2 have been studied in the temperature range 450–600°C under the atmosphere of flowing Ar/5%H2, CO2/1%CO, and CO2. Using the micro-balance technique, the weight change of the specimen has been measured as a function of time. The results showed that the oxidation kinetics of Zircaloy-2 obeyed the cubic rate law rather than the parabolic one. The effect of oxygen partial pressure on the rate constant was not found under the present experimental conditions. On the other hand, the activation energies of the oxidation were 145, 171, and 188 kJ/mol for Ar/5%H2, CO2/1%CO, and CO2 atmospheres, respectively. It was shown from the X-ray diffractometry that the specimens oxidized under the conditions of this study consisted mainly of monoclinic zirconia and, to a minor degree, of tetragonal one. It is suggested that the lateral cracks observed with scanning electron microscopy (SEM) may cause the slow diffusion of oxygen in the oxide phase.

Radiation effect of simulated waste glass irradiated with ion, electron and γ-ray

The density change of simulated waste glass has been measured as a function of electron fluence, using the high voltage electron microscope. A very large value of swelling 0.3, was observed. It is likely that the swelling is due to the formation of oxygen bubbles. It was observed that the swelling saturated at about 0.2 displacements per atom (dpa) and additional swelling initiated beyond 0.5 dpa. Simulated waste glass was irradiated in a nuclear reactor using the 10B(n, α)7Li reaction. The shrinking of the waste glass w observed and it saturated at about 0.2 dpa in which the value of shrinking was 0.0012. Several kinds of glass specimen were irradiated by γ-rays and density changes were measured. The glass of high silica content compacted, while waste glass of low silica content swelled by 0.0022 at 1.0 × 109 R. It can be concluded that the density change induced by γ-rays cannot be neglected, compared with the density change due to irradiation with the 10B(n, α)7Li re α-decay of actinides.

Behavior of metallic fission products in uranium–plutonium mixed oxide fuel

Abstract Metallic fission products, ruthenium, rhodium, technetium, palladium, and molybdenum, exist in irradiated oxide fuels as metallic inclusions. In this work, the radial distributions of metallic inclusion constituents in the fuel specimen irradiated to a peak burnup of 7–13 at.% were observed with an electron probe microanalysis. Palladium concentration is high at the periphery in all the specimens. Molybdenum shows the same tendency for the 13 at.% burnup specimen. These results showed the significant difference between experimental data and calculations with ORIGEN-2 at such high burnups, which suggested that the migration of palladium and molybdenum was controlled mainly by diffusion of gaseous species containing each metal along the fuel temperature gradient.

Review of waste glass corrosion and associated radionuclide release as a part of safety assesment of entire disposal system

Abstract Current knowledge on high-level nuclear waste glass corrosion is summarized, and remaining problems are discussed for meaningful predictions of the glass corrosion and associated radionuclide release as a part of safety assessment of entire disposal system. In recent years, much progress has been made in understanding the mechanism of waste glass corrosion in aqueous environments. Glass corrosion models based on the mechanism have been developed for predicting the long-term glass performance, and they are incorporated as part of radionuclide source term in safety assessments of the disposal system. However, these results have not yet allowed meaningful predictions for the long-term release of individual radionuclides from the glass in repository environments, because mechanism of the long-term glass corrosion has not been fully understood and solubilities of actinoids and fission products under disposal conditions are rather uncertain. In addition, the most serious problem is that the effects of various reactions and interactions occurring in the engineered barrier system, such as corrosion of overpack, alteration of backfill and chemical interactions of the released glass constituents with them have not been fully coupled with the glass performance. These reactions may be dominant processes controlling the glass corrosion and associated radionuclide release for the long-term. For the meaningful predictions, we must evaluate the waste glass performance in combination with the effects of various reactions and interactions occurring in the engineered barrier system on the basis of fully understanding of the chemical and geochemical mechanisms.

Isotope effect in heat of transport of H, D and T in Nb

Abstract The thermal diffusion of hydrogen isotopes, H and D, in Nb was studied at an average temperature of 168°C. By analyzing the redistribution of hydrogen in Nb on the basis of the irreversible thermodynamics, the heat of transport Q∗ was determined for H and D as 9.5 kJ/mol and 16.0 kJ/mol, respectively. The large isotope dependence of Q∗ was concluded by comparing these values with the value of 18.8 kJ/mol for T, which was previously reported by the present authors. The diffusion coefficients of H and D were also determined from the transient process of redistribution and found to be in good agreement with those based on the Gorsky effect.

The oxidation kinetics and the structure of the oxide film on Zircaloy before and after the kinetic transition

Abstract Oxidation kinetics of Zircaloy-4 have been measured using a micro-balance technique in CO–CO 2 gas mixtures between 450°C and 600°C. Oxidation kinetics of Zircaloy-4 obeyed a cubic rate law with time at 450–600°C up to 24 h. At 600°C, the kinetic transition occurred after about 36 h. After the transition, oxidation kinetics obeyed a linear rate law. X-ray diffraction patterns for the samples oxidized at 600°C showed that the volume fraction of tetragonal phase of zirconia decreased with time until the kinetic transition occurred and was almost constant after that. In addition, stresses in the oxide films were found to be larger for the pre-transition samples than for the post-transition ones.

Behavior of Fission Products Zirconium and Barium in Fast Reactor Fuel Irradiated to High Burnup

Barium and Zr generated in nuclear fuels can precipitate as multi-component oxide with some other fission products. In addition, the solubility of Ba in the fuel depends on the oxygen potential and the temperature and Zr can easily dissolve into the fuel matrix. Therefore, the behavior of the Ba-Zr oxide inclusions during irradiation is rather complex. In this work, the composition of multi-component oxides and the distributions of Ba and Zr as a function of relative radius were evaluated with X-ray microanalysis. As results, the oxide inclusions containing both Ba and Zr and containing only Ba were observed in the fuel irradiated to the burnup of 13.3 and 10.6 at%, respectively. These results were discussed in terms of the solubility of Ba and Zr in the fuel and in terms of the rO2–UO2 phase diagram, together with the radial distributions of Ba and Zr in fuel matrix.

Distribution of molybdenum in FBR fuel irradiated to high burnup

Abstract Molybdenum is one of high yield fission products and has a chemical affinity for oxygen in uranium-plutonium mixed oxide fuel. In this work, radial distributions of molybdenum were investigated in high burnup fuels irradiated up to 13.33 at.%. It is found that the distributions are different from those expected from diffusion process in low burnup. This suggests that molybdenum in high burnup fuel migrates by not only a diffusion process but also by a gaseous molybdenum transport mechanism.