Mitsuo Akabori

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.

X-ray absorption study of molten yttrium trihalides

The local structure and structural changes in molten YCl3-LiCl-KCl and molten YBr3-LiBr systems have been investigated by using a high-temperature extended X-ray absorption fine structure (EXAFS) technique. The behaviour of octahedral coordination of the halide ion (Cl(-) and Br(-)) around the Y(3+) ion has been studied by EXAFS of the Y K-absorption edge. The nearest Y(3+)-Cl(-) and Y(3+)-Br(-) distances and coordination numbers of halide ions around the Y(3+) ion do not change by mixing with the alkali halides. The stabilization of the (YCl6)(3-) and (YBr6)(3-) octahedral coordination by adding alkali halides was suggested by decreasing the Debye-Waller factor and the anharmonicity in the nearest Y(3+)-Cl(-) and Y(3+)-Br(-) interactions. The bridging structure of the (YBr6)(3-) octahedra sharing a Br(-) ion in the molten YBr3-LiBr system was studied by EXAFS of the Br K-absorption edge. The coordination number of Y(3+) around the Br(-) ion decreases from 2 in the pure melt to almost 1 in the 30mol% and 15mol% YBr3 melts. This suggests that the bridging is almost broken and the stable octahedron exists freely in the LiBr-rich melts.

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.

Release behavior of cesium in irradiated (Th, U)O2

Abstract The release behavior of fission cesium in low burnup (Th, U)O 2 was studied in post-irradiation annealing at temperatures from 1473 to 1773 K, where the released Cs was directly collected and its quantity was measured by using a new technique. The release rate was controlled by the diffusion process in (Th, U)O 2 , and the rate decreased with increasing grain size and burnup. The diffusion coefficients were determined by the equivalent sphere model, which is based on the assumption that the equivalent sphere for the Cs release is equal to the grain. The activation energy for Cs diffusion increased with increasing burnup: 464 kjmol −1 at 0.005% FIMA and 664 kJmol −1 at 0.15% FIMA.

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.

Thermochemical modeling of actinide alloys related to advanced fuel cycles

Abstract Innovative approaches for improved management of transuranium elements are being sought. The use of either metallic or nitride fuels, coupled with pyrochemical reprocessing techniques, are being studied. As our experience with these advanced fuels is limited, their development will be promoted by thermochemical analysis. This paper reviews our thermochemical modeling of technical problems related to advanced fuel cycles and presents models of the actinide alloys and nitrides together with experimental verification of these thermochemical predictions.