Tetsuo Shiratori

Chemical form of the solid fission products in (Th, U) O2 simulating high burnup

The chemical form of the solid fission products has been studied for (Th0.81U0.19) O2 simulating 21.5% FIMA in an HTGR environment. Experiments have been performed with X-ray diffraction, electron-probe microanalysis, ceramography and hardness measurement. The results showed that fission-product phases of two types, Mo-Ru-Pd and (Ba, Sr)(Zr, Ce) O3, are present in the simulated fuel pellet. The fuel matrix comprises (Th1-xUx, Zr, Ce, RE) O2 with an x value of 0.067(RE = Nd, La, Pr, Y, Sm). Dissolution of the rare earths (RE) and the residual Zr plus Ce in (Th0.933U0.067) O2 was accompanied by contraction of the unit cell of the oxide matrix. Reaction behavior in the selected fission product system BaO-ZrO2-Nd2O3 was also investigated. The results showed that in the presence of BaO, Nd2Zr2O7 is converted to barium zirconate: at 1630°C, Nd2Zr2O7 + 2 BaO → 2 BaZrO3 + Nd2O3. This fact, combined with thermochemical assessment, confirms the relative stability of (Ba, Sr)(Zr, Ce) O3 against Nd2Zr2O7 in the simulated (Th0.81U0.19) O2. From these results and fission product inventories, it is inferred that the chemical state of high-burnup ThO2 is very similar to that of (Th0.81U0.19) O2.

Preparation of rock-like oxide fuels for the irradiation test in the Japan Research Reactor No. 3

Three types of uranium-based rock-like oxide (ROX) fuel were prepared for an irradiation test in the Japan Research Reactor No. 3 (JRR-3). The first one was a particle dispersed type fuel where (U,Zr,Y)O 2 particles of 250 μm in diameter were dispersed in a matrix of spinel or corundum. The second one was a homogeneous mixture type of (U,Zr,Y)O 2 and spinel or corundum. The third one was a (U,Zr,Y)O 2 single phase fuel. In all case, the sintering of the pellets was carried out at 2020 K for 4 h in a stream of 75%H 2 /25%N 2 mixed gas. The characterization tests showed that the pellets fabricated had small void volume fraction (<9%). The homogeneous distribution of the particles/grains of (U,Zr,Y)O 2 in the matrix was confirmed by the ceramography. The X-ray diffraction analysis (XRD) showed the formation of homogeneous (U,Zr,Y)O 2 solid solutions and no appreciable interactions between the (U,Zr,Y)O 2 phase and inert matrix materials (spinel or corundum).

Rim structure formation of isothermally irradiated UO2 fuel discs

UO2 fuel discs, which had been irradiated at an isothermal condition of 550-630 degreesC to 51, 86 and 90 GWd/t without any restraint pressure, have been subjected to detailed microstructure observations. elemental analyses and density measurements. Their data were compared with previously reported results of high-burnup Zircaloy-clad type fuel pellets, so as to clarify the effect of pellet-cladding interaction (PCI) restraint on the rim structure formation. The porous rim structure, accompanying huge bubbles and high porosity, was recognized for the high-burnup discs of 86 and 90 GWd/t, but not for the 51 GWd/t disc. From the good coincidence between porosity increase and density decrease for the high-burnup discs, it was concluded that the precipitation and growth of coarsened rim bubbles substantially caused fuel swelling. The wide variety of rim bubble sizes and porosities at a given local burnup, which was recognized in the present PCI-free discs and Zircaloy-clad type pellets reported in other literature. possibly resulted from the external PCI restraint effect. The pressure difference between bubble internal and external pressures, and vacancy diffusivity would rate-control the growth of rim bubbles

Solubility of magnesium in uranium dioxide

Abstract The solubility of magnesium in uranium dioxide under low oxygen pressures was studied at 1200°C. Magnesium was found to dissolve up to y > 0.1 (and below y = 0.15) of the apparent formula, MgyU1−yO2 + x ( x ⪋ 0 ) on heating at po2 = 10−15 and ≤ 10−19 atm. The formed solid solution in such a low po2 region was of the type (MgaU1−a){Mgb}O2 + c, in which the magnesium atoms partly occupy the interstitial sites together with the substitutional sites for uranium atoms. The ration of interstitial atoms to the total magnesium atoms increased from 0.23 (y = 0.05) or 0.39 (y = 0.1) at po2 = 10−15 atm with decreasing oxygen partial pressure to 0.62–0.63 (y = 0.05 and 0.1) at po2 ≤ 10−19 atm. The lattice parameter of the (MgaU1−a){Mgb}O2 + c solid solutions was represented as a linear equation of a, b and c. The interstitial magnesium caused an increase in the lattice parameter, in contrast to the substitutional magnesium which largely decreases the lattice parameter. It is possible that the uranium atoms in the solid solutions prepared at low oxygen partial pressures (≤ 10−19 atm) were reduced to slightly less than the tetravalent state.

Thermal conductivities of irradiated UO2 and (U, Gd)O2

The evaluation of thermal conductivity of irradiated fuel is very important since it directly affects the fuel operating temperature. The disk-shaped UO2 and UO2-10 wt%Gd2O3 samples were prepared and irradiated to about 4%FIMA to measure the thermal diffusivities by the laser flash method. The burnup was almost uniform within each sample. The irradiation temperature was almost constant and uniform within each sample except the temperature escalation that occurred during the irradiation. The thermal conductivity, determined from the thermal diffusivity, density and specific heat capacity, decreased by irradiation, while it partly recovered after the thermal diffusivity measurement at temperatures up to about 1800 K. The thermal conductivity reduction attributable to the irradiation-induced point defects was small in the samples which experienced higher temperature than 1273 K during the temperature escalation. The present results were compared with the reported models

Fabrication of very high density fuel pellets of thorium dioxide

Abstract Very high density ThO 2 pellets were prepared without binders and lubricants from the ThO 2 powder originated by the thorium oxalate, which was aimed to simplify the fabrication process by skipping a preheat treatment. The as-received ThO 2 powder with a surface area of 4.56 m 2 /g was ball-milled up to about 9 m 2 /g in order to increase the green pellet density as high as possible. Both of the single-sided and the double-sided pressing were tested in the range from 2 to 5 t/cm 2 in the green pellet formation. Sintering temperature was such low as 1550°C. The pellet prepared in this experiment had a very high density in the range from about 96 to 98% TD without any cracks, in which a difference of the pellet density was not recognized in the single-sided pressing methods.

Thermodynamic properties of Th0.80U0,20O2 + x solid solution

Oxygen-potential (ΔGO2) measurements employing a thermogravimetric method have been performed for Th0.80U0,20O2 + x. A complete set of data is presented at 1273–1473 K in the ranges 2.000 ≲ OM ≲ 2.024 and −95 ≲ ΔGO2

Post-irradiation examination of high burnup mg doped UO2 in comparison with undoped UO2, Mg-Nb doped UO2 and Ti doped UO2

−32 kcal/mol. Partial molar entropies and enthalpies of solution of oxygen in the mixed oxide were derived from the temperature variation of ΔGO2. Vapor pressures over Th0.80U0,20O2 + x at 2000 at 2300 K were calculated from our experimental ΔGO2 data and the known free energies of formation for gaseous and condensed oxides. It is predicted that with an increase in O/M ratio the vapor pressure of UO3(g) increases rapidly while maintaining an extremely lower pressure of ThO2(g).

Study on the development of the lattice strain in (Mg,U)O2+x solid solution

Abstract The pellets of UO 2 , magnesium doped UO 2 (Mg–UO 2 ), magnesium and niobium doped UO 2 (Mg–Nb–UO 2 ) and titanium doped UO 2 (Ti–UO 2 ) were irradiated to burnups ranging from 19 to 94 GWd/tU at temperatures 550–930°C. The solubility of magnesium in UO 2 was low around 2 mol%. The addition of magnesium and titanium caused to form large grain sized pellet on sintering. The swelling of pellets during irradiation was unchanged by magnesium addition below 60 GWd/tU in agreement with the literature rate for UO 2 . The thermal conductivity of unirradiated Mg–UO 2 was higher than that of undoped UO 2 , which seemed to also hold for irradiated specimens. Pellet fracturing occurred by irradiation mainly by thermal stress. The undoped and metal doped UO 2 pellets in the 84–94 GWd/tU range at the irradiation temperatures of 560–640°C showed large bubbles and sub-divided grains of sub-micron size and the rim structure formation all over the surface. The xenon release from the pellets during irradiation increased with increasing burnup. In the fuels of close burnups, the xenon release increased rapidly with increasing temperature above about 600°C. At high burnups, the effect of metal addition seemed to recede unclear perhaps due to the formation of heavily damaged fuel matrix.

Dissolution of ThO2-Based Oxides in Nitric Acid Solutions at Elevated Temperatures

Abstract Lattice strain observed in magnesia-doped UO 2 was studied. Solid solution Mg y U 1− y O 2+ x was prepared by the reaction of mixtures of MgUO 4 , MgU 3 O 10 and UO 2 . The composition of the solid solution was analysed by X-ray diffraction (XRD) and electron probe microanalysis. Homogeneous and inhomogeneous strains were evaluated by XRD analysis. The homogeneous strain was explained as caused by the cation defects in the solid solution. The development of the strain was dependent on the number of Mg 2+ -U 5+ complexes. The variance method was applied to obtain the inhomogeneous strain. The average crystallite size of the product was also given in the variance analysis. The evolution of the inhomogeneous strain and the decrease in the size of the crystallite were attributed to the precipitation of MgO accompanied by the reduction of the sample.