Tatsumi Arima

Manufacturing of zirconia microspheres doped with erbia, yttria and ceria by internal gelation process as a part of a cermet fuel

Zirconium oxide is an inert matrix candidate for the transmutation of plutonium in light water reactor (LWR). The thermal conductivity of cubic zirconia is however lower than the conductivities of UO2 and MOX. Special designs are therefore necessary to avoid high peaking temperatures close to the melting point in the zirconia pellet. Cermet would be a favorable design to improve the thermal conductivity. The suggested cermet fuel consists of fine plutonium doped stabilized zirconia particles dispersed in a metallic inert matrix. Manufacturing tests on cubic zirconia microspheres were carried out by using the internal gelation process developed at the Paul Scherrer Institute. Gelation was conducted successfully and the sintered spheres had a homogeneous single cubic structure. The lattice parameter of the cubic zirconia was estimated as a function of the Er, Y and Ce atomic fraction using a simplified semi-quantitative formula. On the experimental side, it is necessary to further investigate the ideal fabrication conditions, because some gel spheres were opaque and fragile and most of the sintered spheres were cracked, nicked and porous.

Thermal conductivity of zirconia based inert matrix fuel: use and abuse of the formal models for testing new experimental data

An inert matrix fuel material based on yttria-stabilized cubic zirconia: Er x Y y Pu z Zr 1-x y z O 2 1-x y /2 (x + y= 0.15, z: [0.05-0.15]) was proposed for burning excess plutonium in light water reactors. The studied inert matrix fuel is made of cubic stabilized zirconia. The limited number of experimental thermal conductivity data justifies this formal and intensive study. Approaches derived from Klemens theory were revisited and the derived conductivity model applied for zirconia, accounting the effects of phononic scattering centers. The hyperbolic thermal conductivity trend with temperature known for pure zirconia, is reduced by isotopes, impurities, dopants and oxygen vacancies, which act as scattering centers and contribute to conductivity reduction to a flat plot with temperature for stabilized zirconia. It is experimentally observed that the thermal conductivity derived from laser flash measurements for Er x Y y M z Zr 1-x y z O 2 x y /2 (with M=Ce or Pu, z = 0 or ∼0.1 and x+y = 0.15) is rather constant as a function of temperature in the range 300-1000 K. The thermal conductivity was observed to depend on the concentration of dopants such as YO 1.5 and/or ErO 1.5 , CeO 2 (analogous of PuO 2 ) or PuO 2 . The bulk material conductivity of Er 0.05 Y 0.10 Pu 0.10i Zr 0.75 O 1.925 is about 2 W m 1 K 1 .In this study, the thermal conductivity data of both monoclinic and stabilized cubic zirconia based IMF are tested with the model approach in order to understand the experimental data in a semi-quantitative way.

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.

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.

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.

Initial dissolution rate of a Japanese simulated high-level waste glass P0798 as a function of pH and temperature measured by using micro-channel flow-through test method

Aqueous dissolution tests were performed for a Japanese type of simulated high-level waste (HLW) glass P0798 by using a newly developed test method of micro-channel flow-through (MCFT) method, and the initial dissolution rate of glass matrix, r 0, was measured as a function of solution pH (3–11) and temperature (25–90°C) precisely and consistently for systematic evaluation of the dissolution kinetics. The MCFT method using a micro-channel reactor with a coupon shaped glass specimen has the following features to provide precise and consistent data on the glass dissolution rate: (1) any controlled constant solution condition can be provided over the test duration; (2) the glass surface area actually reacting with solution can be determined accurately; and (3) direct and totally quantitative analyses of the reacted glass surface can be performed for confirming consistency of the test results. The present test results indicated that the r 0 shows a “V-shaped” pH dependence with a minimum at around pH 6 at 25°C, but it changes to a “U-shaped” one with a flat bottom at neutral pH at elevated temperatures of up to 90°C. The present results also indicated that the r 0 increases with temperature according to an Arrhenius law at any pH, and the apparent activation energy evaluated from Arrhenius relation increases with pH from 54 kJ/mol at pH 3 to 76 kJ/mol at pH 10, which suggests that the dissolution mechanism changes depending on pH.

Oxidation behavior of modified SUS316 (PNC316) stainless steel under low oxygen partial pressure

Abstract Oxidation behaviors of modified SUS316 (PNC316) and SUS316 stainless steels were investigated under the low oxygen partial pressure of 10 −31 −10 −22 atm at 600–800 °C. Oxygen uptake by these materials parabolically increased with time, and the kinetic rate constants depended on both oxygen partial pressure and temperature. Thus, semi-empirical equations of the parabolic rate constants were obtained to be 2.70×10 4 exp(−109/ RT ) P O 2 0.279 for PNC316 and 9.23×10 4 exp(−98/ RT ) P O 2 0.313 for SUS316. For the duplex layer formed under the low oxygen partial pressure, the inner layer consisted of such oxides as Cr 2 O 3 and FeCr 2 O 4 , while the outer layer consisted of non-oxidized α-Fe. Furthermore, oxidation along the grain boundaries was observed for samples oxidized for a long time. From the point of view of fuel cladding chemical interaction evaluation at high burn-up fuel for fast reactors, it is interesting that formation of non-oxidized α-Fe was observed under the low oxygen partial pressure.