Shuhei Miwa

Research and Development of Minor Actinide-containing Fuel and Target in a Future Integrated Closed Cycle System

Research and development of minor actinide-containing fuels and targets, i.e., (Pu,Am)O2–MgO, (Pu,Np)O2–MgO, (U,Pu,Np)O2, (U,Pu,Np)N and (Pu,Np,Zr)N, for use in a future integrated closed cycle system that includes fast reactor and accelerator driven sub-critical system is underway. The present statuses of fabrication test and property measurements are given. Design concept of the oxide target is described in detail together with a screening of the support material. A new apparatus for the measurement of mechanical properties at the elevated temperature is installed for use in evaluating the fuel-cladding mechanical interaction. Development histories with future prospects of two types of Np-containing fuels for the fast reactor are mentioned. Preliminary test results for a new nitride target for the accelerator driven sub-critical system are given. Finally, an irradiation test plan in the experimental fast reactor JOYO is briefly described.

Low temperature liquid-phase-assisted sintering of Si3N4 ceramics as an inert matrix for confinement of minor actinides

Silicon nitride ceramics is a candidate for inert matrix, since it is very tolerant for neutron irradiation and it has relatively high thermal conductivity. For these reasons, Si3N4 ceramics containing large amounts of CeO2, as a simulant of transuranium elements, were manufactured by sintering. The low-temperature sintering behavior and chemical and thermal properties of the obtained ceramics are reported. CeO2 (16 or 30 mass%), SiO2 (3 mass%) and MgO (5 mass%) were mixed with fine Si3N4 powder. The mixed powder was uniaxially pressed into a pellet under 60 MPa. These pellets were sintered at 1400~1750°C for 2~4 h in flowing N2 (2 L/min). The pellet density rapidly increased after the sintering between 1400 and 1430°C in the case of 4 h keeping and bulk density of 3.34 g/cm3 was attained at 1450°C (Relative density ~95%), in the case of 16mass% CeO2. Room temperature thermal conductivity increased with sintering temperature from ~10 to ~40W/mK after sintering at 1450°C and 1650°C, respectively. High temperature thermal conductivity of sintered ceramics at 1650°C was approximately four times as high as that of UO2 up to 800°C. The dissolution rate of the sintered pellets for 3mol/L HNO3 solution at 80°C was measured. After 200 h, almost 25% of grain boundary phase was eluted from the pellet containing 16mass% CeO2.

Effects of interaction between molten zircaloy and irradiated MOX fuel on the fission product release behavior

As a first step for obtaining experimental data on the effects of high-temperature chemical interaction on fission product release behavior, we focused on the dissolution of irradiated uranium plutonium mixed oxide (MOX) fuel by molten zircaloy (Zry) and carried out a heating test under the reducing atmosphere. Pieces of an irradiated MOX fuel pellet and cladding were subjected to the heating test at 2373 K for five minutes. The fractional release rate of cesium (specifically 137Cs) was monitored during the test and its release behavior was evaluated. The observation of microstructures and measurements of elemental distribution in the heated specimen were also performed. We demonstrated experimentally that the fuel dissolution by molten Zry accelerated the release of Cs from the fuel pellets.

Oxygen Potential of (Th0.7Ce0.3)O2- x

Oxygen potentials of (Th 0.7 Ce 0.3 )O 2- x were experimentally determined by means of thermogravimetric analysis as a function of non-stoichiometry at 1173 and 1273 K. Oxygen potentials of (Th 0.7 Ce 0.3 )O 2- x at each temperature increased with increase of oxygen to metal (O/M) ratio (=2- x ) and steep increases of the oxygen potentials when approaching O/M ratio = 2 were observed. These characteristics are typical for non-stoichiometric fluorite-type actinide dioxides. The oxygen potentials of (Th 0.7 Ce 0.3 )O 2- x were similar to those of CeOO 2- x when they were plotted as a function of average Ce valence.

Modeling and Simulation of Thermophysical Properties of Minor Actinides-Containing Oxide Fuels

The molecular dynamics (MD) calculation was performed for minor actinide (MA: Np and Am)-containing mixed oxide (MOX) fuels, U 0.7-x Pu 0.3 MA x O 2 , in the temperature range from 300 to around 2500 K to evaluate the thermal expansion, heat capacity, and thermal conductivity. The MD results showed that the calculated heat capacity and thermal conductivity were similar in all the composition ranges, indicating that MA scarcely affected the thermal properties of the MOX fuel in the perfect crystal system.