Masahiko Osaka

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.

A Novel Concept for Americium-containing Target for Use in Fast Reactors

A novel concept for the americium-containing target, which consists of (Th,Am)O2-x and molybdenum support material, is proposed that can fulfill several requirements such as reduction of the environmental burden and effective use of natural resources. In particular, deployment of recovered molybdenum from the spent nuclear fuel is highlighted and its nuclear effects are focused on in comparison to other types of molybdenum, namely natural and enriched ones. A preliminary investigation for effects of molybdenum isotopic composition on fast reactor core characteristics were carried out using nuclear calculation codes. The results indicated the feasibility of the fast reactor core loaded with the present target by optimizing its amount and composition. Moreover, several additional gains were indicated such as possible use of the radioactive waste and production of a new energy source, 233U.

Molecular Dynamics Studies of Americium-Containing Mixed Oxide Fuels

The molecular dynamics (MD) calculation was performed for americium-containing mixed oxide fuels, (U0-7--x Pu0.3Am x )O2 (x=0,0.016; 0.03; 0.05; 0.1; 0.15), in the temperature range from 300 to around 2,500 K to evaluate the lattice parameter, heat capacity and thermal conductivity. The MD results reveal that the calculated heat capacity and thermal conductivity are at a similar level in the entire composition range, in other words they are scarcely influenced by adding americium up to 15%. This behavior was examined from a view point of a phonon-impurity scattering mechanism.

Protected Plutonium Production by Transmutation of Minor Actinides for Peace and Sustainable Prosperity—Irradiation Tests of Np and Np-U Samples in the Experimental Fast Reactor JOYO (JAEA) and the Advanced Test Reactor at INL—

A project on Protected Plutonium Production (P3) was proposed by Tokyo Institute of Technology as part of a nonproliferation research program for plutonium (Pu) utilization in nuclear reactors. The project is aimed at the production of inherently protected Pu by the addition of 237Np to uranium (U) fuel. In order to validate this P3 concept, two irradiation tests were performed. In the first, a determination of Pu isotopes in 237Np samples irradiated in the experimental fast reactor JOYO was done to evaluate 238Pu production from 237Np under fast neutron spectra. The fast reactor can undertake P3, which can be better performed in the reflector region. In the test, the percentage of the total amount of 238Pu atoms transmuted from 237Np atoms by irradiation was around 90%. In the second test, 2, 5, and 10% Np-containing U samples were irradiated in the Advanced Test Reactor at the Idaho National Laboratory (INL) to evaluate 238Pu production in the thermal neutron region. The fuel specimens were removed from the core at 100, 200, and 300 effective full power days (EFPDs), and then a postirradiation examination was completed at an analytical laboratory in the Materials & Fuels Complex (MFC) at the INL. For the samples after irradiation for 300 EFPDs, Np depletions were about 60% for 2% neptunium (Np)-U samples and about 50% for 5 and 10% Np-U samples. The 238Pu-to-Pu ratios were about 20, 30, and 45% for 2, 5, and 10% Np-U samples, respectively.

Analysis of Curium in Mixed Oxide Fuel Irradiated in the Experimental Fast Reactor JOYO for the Evaluation of Its Transmutation Behavior

Curium isotopes generated in the MOX fuel irradiated in the experimental fast reactor JOYO were analyzed by applying a sophisticated radiochemical technique. Curium was isolated from the irradiated MOX fuel by anion-ex- change chromatography using a mixed medium of nitric acid and methanol. The isotopic ratio of curium and its content were determined by thermal ionization mass spectroscopy and alpha-spectrometry, respectively. The curium content was less than 0.004 at% even at high burnup of 120GWd/t, which is much smaller than that of PWR-MOX at 60 GWd/t. On the basis of present analytical results, the transmutation behavior of curium isotopes in a fast reactor was discussed from various viewpoints. Transmutation rates of curium isotopes were estimated; the rate for 246Cm, which is known to be a key nuclide in the transmutation of curium, was larger than the previously reported value. It was concluded from these evaluations that the fast reactor was suitable for the incineration of curium.

Densification of Molybdenum-Cermet Fuel by Sintering with Metal Additives

Fabrication tests of UO 2 - and PuO 2 -containing Mo-cermet pellets with Pd or Ni additives were carried out. It is concluded that Pd is a useful and effective sintering additive since it can provide dense Mo-cermet fuels with a good thermal performance

Analysis of curium isotopes in mixed oxide fuel irradiated in fast reactor

Curium is one of the key elements in recycling and transmutation of minor actinides (MA) because of its high ra- diotoxicity and difficulty of transmutation. In order to make isotopic analysis of curium in heavily irradiated fuel, the isolation technique of curium was developed by adopting anion exchange chromatography in nitric acid-methanol mixed media. The technique was successfully applied to the analysis of curium in mixed oxide (MOX) fuel irradiated in the experimental fast reactor “JOYO”. The transmutation behaviors of curium in fast reactor are discussed on the basis of observed isotopic ratio of curium.

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.

Molecular Dynamics Study on Defect Structure of Gadolinia-Doped Thoria

A molecular dynamics study on the defect structure of Gd-doped thoria was carried out. The partially ionic two-body potential was used with empirically obtained parameters for Th, Gd and O ions in the fluorite structure. Different defect structures were assumed and their effects on the lattice parameter were investigated. Molecular dynamics simulations were carried out with Gd fractions and temperatures ranging from 0 to 45% and from 298 to 1200 K, respectively. Diffusion coefficients of O ions in the Gd-doped thoria having various Gd fractions were also calculated at 1273 K. Comparison of the calculated results with the experimental ones lead to the conclusion that the Gd-oxygen vacancy-Gd type cluster is the most likely defect structure in the Gd-doped thoria.

Penetration behavior of water solution containing radioactive species into dried concrete/mortar and epoxy resin materials

Penetration behavior of radionuclides such as 137Cs into dried concrete material, dried mortar material and epoxy paint for a few dozen days was observed using a solution containing fission products extracted from irradiated fuels to obtain fundamental information on the radionuclide penetration rate and depth. Hardly any radionuclides could penetrate into the epoxy paint. The radionuclide solution penetrated into concrete and mortar materials to a depth of a few millimeters for a few dozen days. The penetration behavior observed near the surface of concrete and mortar materials was similar to the diffusion of nuclides in media such as water-saturated concrete, bentonite and cement materials.