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Kato, Masato; Hirooka, Shun; Ikusawa, Yoshihisa; Takeuchi, Kentaro; Akashi, Masatoshi; Maeda, Koji; Watanabe, Masashi; Komeno, Akira; Morimoto, Kyoichi
Proceedings of 19th Pacific Basin Nuclear Conference (PBNC 2014) (USB Flash Drive), 12 Pages, 2014/08
Uranium and plutonium mixed oxide (MOX) fuel has been developed for Japan sodium-cooled fast reactors. Science based fuel technologies have been developed to analyse behaviours of MOX pellets in the sintering process and irradiation conditions. The technologies can provide appropriate sintering conditions, irradiation behaviour analysis results and so on using mechanistic models which are derived based on theoretical equations to represent various properties.
Kawamura, Hideyuki; Kobayashi, Takuya; Furuno, Akiko; Usui, Norihisa*; Kamachi, Masafumi*; Nishikawa, Shiro*; Ishikawa, Yoichi*
Proceedings of 19th Pacific Basin Nuclear Conference (PBNC 2014) (USB Flash Drive), 7 Pages, 2014/08
Numerical simulations on oceanic dispersion of the radioactive cesium in the North Pacific from March 2011 to September 2013 were conducted to clarify the concentration of the radioactive cesium released from the Fukushima Daiichi Nuclear Power Plant. We implemented the oceanic dispersion simulations with two independent ocean reanalysis dataset. It was suggested that the 
Cs concentration in the North Pacific was lower than the pre-Fukushima background level about two years after the Fukushima disaster. The intercomparison revealed that meso-scale eddies in the Kuroshio Extension region may have efficiently diluted the radioactive cesium concentration at the sea surface. In addition, it was suggested that the enhanced downward current accompanied by the meso-scale eddies played an important role in transporting the radioactive cesium into the intermediate layer.
Hirooka, Shun; Kato, Masato; Morimoto, Kyoichi; Washiya, Tadahiro
Proceedings of 19th Pacific Basin Nuclear Conference (PBNC 2014) (USB Flash Drive), 8 Pages, 2014/08
Since the severe accident at Fukushima Daiichi Nuclear Power Station, technologies to remove fuel debris from the damaged core have been developed. However, many subjects such as how to access to the core, cut the fuel debris, control criticality safety, estimate fissile materials, store removed debris and so on are still in existence. Purpose of this work is to evaluate the fuel debris properties by using analysis of simulated fuel debris and to estimate the inner state such as temperature profile and density profile which depend on elapsed time after the accident. The reported properties such as melting temperature, thermal conductivity and thermal expansion were obtained by the simulated fuel debris manufactured from UO
and zircaloy.
Yokota, Hideharu; Tanaka, Shingo
Proceedings of 19th Pacific Basin Nuclear Conference (PBNC 2014) (USB Flash Drive), 10 Pages, 2014/08
For understanding mass transport mechanism in the rock as natural barrier, 
dipole tracer migration experiments were implemented for a fracture in mudstone of Wakkanai Formation at the G.L.-250m gallery of the Horonobe URL, northern Hokkaido, Japan. Results of its preliminary analysis by the analytical solution retrieved from one-dimensional advection-dispersion equation indicate that a fracture surface in mudstone has both reversible and irreversible sorption characteristics for Cs
and Sr
but only irreversible that for Co and Eu
.