Akinori Itoh

Fabrication of nitride fuels for transmutation of minor actinides

At the Japan Atomic Energy Research Institute, the concept of the transmutation of minor actinides (MA: Np, Am and Cm) with accelerator-driven systems is being studied. The MA nitride fuel has been chosen as a candidate because of the possible mutual solubility among the actinide mononitrides and excellent thermal properties besides supporting hard neutron spectrum. MA nitrides of NpN, (Np, Pu)N, (Np, U)N, AmN, (Am, Y)N, (Am, Zr)N and (Cm, Pu)N were prepared from the oxides by the carbothermic reduction method. The prepared MA nitrides were examined by X-ray diffraction and the contents of impurities of oxygen and carbon were measured. The fabrication conditions for MA nitrides were improved so as to reduce the impurity contents. For an irradiation test of U-free nitride fuels, pellets of (Pu, Zr)N and PuN + TiN were prepared and a He-bonded fuel pin was fabricated. The irradiation test started in May 2002 and will go on for two years in the Japan Materials Testing Reactor.

Stability and structure of the δ phase of the U-Zr alloys

Abstract Homogeneity range and crystal structure of the intermediate δ phase in the UZr alloy were examined by electron probe microanalysis, X-ray diffraction and differential thermal analysis, using the alloys prepared by arc-melting. The homogeneity range of the δ phase was found to be 64.2–78.2 at% Zr at 600°C and 66.5–80.2 at% Zr at 550°C. The powder diffraction patterns of the δ phase agreed with an unusual structure of modified C32, in which the (0,0,0) sites are preferentially occupied by Zr atoms, and the ( 2 3 , 1 3 , 1 2 ) and ( 1 3 , 2 3 , 1 2 ) sites randomly shared by U and Zr atoms.

Reactions of U–Zr alloy with Fe and Fe–Cr alloy

Abstract The reaction zones formed in two kinds of diffusion couples: U–23at.%Zr/Fe and U–23at.%Zr/Fe–12at.%Cr, at 908, 923, 953, 973 and 988 K have been examined using the electron-probe microanalysis. In the U–Zr/Fe–Cr couples, diffusion of Cr to the U–Zr side is slower than that of Fe, and the Cr-rich phase is formed adjacent to the unreacted Fe–Cr alloy. Except for the Cr-rich phase, the measured compositions of the phases in the reaction zones in both U–Zr/Fe and U–Zr/Fe–Cr couples have corresponded well to those in the U–Zr–Fe ternary system. Each reaction zone can be divided to several layers. For the U–Zr side of the reaction zone, the configurations of the schematic diffusion paths, which are the curves connecting the average compositions of these layers on the U–Zr–( Fe + Cr ) composition triangle, are independent of the annealing temperature and the Cr addition to Fe. For the Fe(–Cr) side, however, the paths depend on the annealing temperature and the Cr addition to Fe. Some of the phases that are expected to emerge considering the schematic diffusion path and the U–Zr–Fe phase diagram have not been found at 988 K.

Interdiffusion in uranium-zirconium solid solutions

Abstract Interdiffusion coefficients in the (γ-U, β-Zr) solid solutions have been measured in the temperature range of 973 to 1223 K and in the Zr concentration range of 0.1 to 0.95 atomic fraction. The data measured at lower temperature than 1223 K have shown notable depression in the Zr concentration range of 0.2 to 0.4 atomic fraction. The dependencies of the obtained data on alloy composition and temperature are consistent with variation of the thermodynamic factor which is calculated based on the evaluation of the activity coefficient in the U Zr system. The interdifusion coefficient in the neighborhood of the Zr atomic fraction of 0.3 cannot be expressed by a single set of a frequency factor and an activation energy because of the significant influence of the thermodynamic factor on the interdiffusivity in the U Zr solid solutions.

Irradiation behavior of high uranium-density alloys in the plate fuels

Abstract The U 6 Mn and U 6 Fe 0.6 Mn 0.4 alloys were irradiated to ∼54% of the initially contained 19.6% 235 U at around 190°C for 216 days. The volume swelling of the miniplates with such fuel dispersions in an Al matrix was reduced to the previously reported U 6 Fe plate. Irradiation tests using U 6 No 0.6 Fe 0.4 and U 3 Si 0.8 Ge 0.2 proved that the cladding restraint is more effective to suppress the gas-bubble growth in the foil-type than in the dispersion-type plates. The fuel-aluminium reaction was also investigated.

Carbothermic synthesis of (Cm,Pu)N

Abstract Nitrides are being considered as fuel candidates for the transmutation of minor actinides. Carbothermic reduction of oxides is an effective method to fabricate nitride fuels. In this study, (Cm 0.40 ,Pu 0.60 )N was synthesized by the carbothermic reduction of the mixed oxide in N 2 . By applying excess carbon, an oxide-free nitride was obtained at 1773 K. The lattice parameter of the oxide-free sample was 0.4948 nm, and that of the nitride with oxides was 0.4974 nm. The former value agreed well with that estimated from the literature values for CmN and PuN. The larger lattice parameter of the latter sample is considered to be due to the oxygen dissolved in the nitride.

Oxygen solubility in dysprosium mononitride prepared by carbothermic synthesis

Abstract Dysprosium mononitride was prepared by carbothermic reduction of the sesquioxide in nitrogen gas stream. A variety of the initial C/Dy molar ratios were chosen to prepare the series of Dy(N, O) solid solutions. The residual carbon in the products was removed by additional heating with nitrogen and hydrogen mixed gas stream. The lattice parameter and composition of Dy(N, O) were determined by X-ray diffraction and quantitative analyses on N, O, and C. The lattice parameter of Dy(N, O) decreased with increasing oxygen content. The oxygen solubility in DyN–Dy2O3 pseudo-binary system under 1 atm of nitrogen increased linearly with increasing temperature from ∼9 mol% DyO at 1623 K to ∼14 mol% DyO at 2075 K.

The mechanisms of fission gas release from (Th, U)O2

The releases of xenon from three (Th, U)O2 specimens with different U contents were measured over a wide range of fission dose from 2.9 × 1019 to 2.2 × 1022 fissions m−3 by using a post-irradiation technique. The releases were found to decrease with dose and to level off at higher doses. Measurements of the changes in lattice parameter and specific surface area of the same specimens enabled one to conclude that the decrease in release originates in the trapping of xenon by the vacancies and vacancy clusters induced by fission fragments. And the release mechanisms of fission gas were proposed based on the proper evaluation of the observation on radiation damage and recovery in oxide fuel.

Irradiation behavior of microspheres of U-Zr alloys

Abstract In order to understand the fundamental behavior of metallic fuels, microspheres of U-Zr alloys were prepared and irradiated. The microspheres of about 0.8 mmφ were formed by solidifying fused droplets of U-Zr alloys. Irradiation temperatures were controlled by external electric heating. Burnup reached 1.5 at%. After the irradiation, fractional release of fission-product gases, dimensional change and microstructure were examined. The results were compared with the analysis by a metal-fuel performance code ALFUS.

Behavior of neutron-irradiated U3Si

Abstract The behavior of U 3 Si-based alloys has been studied under neutron irradiation. Maximum burnup reached ∼62% of the initially contained 19.6% 235 U and irradiation temperatures were in the approximate range of 190–280°C. Postirradiation examinations revealed the following. The dimensional instability of the high uranium-density fuels is attributed to the radiation-induced amorphization and plastic deformation. The occurrence of amorphization is suggested by the liquid-like behavior of U 3 Si under irradiation at temperatures well below the melting point of crystalline material. The accelerated swelling of U 3 Si due to the large fission-gas-bubble growth can be suppressed by the cladding restraint. The reaction layer is formed at the U 3 Si Al interface. The thickness of the reaction layer of surface-oxidized U 3 Si is significantly reduced in comparison with that of non-oxidized U 3 Si. This reduction in thickness is caused by the thin film of UO 2 that has been formed at the surface of the U 3 Si by oxidation.