Recent activities in the field of nuclear materials and nuclear fuels

The research and development of nuclear materials and fuels is aggressively advanced for the improvement of reliability or performance of nuclear systems, such as light water reactor, fast breeder reactor (FBR), fusion reactor and so on. Many of outstanding researches are presented by the Journal of Nuclear Science and Technology. The latest activity in nuclear materials and fuels is introduced in this summary by surveying the recent journal.

Several ceramic materials possess many superior properties for fusion reactors or high-level radioactive waste forms. Though thermal property is one of the most important factors, neutron-irradiated ceramics show severe degradation in thermal diffusivity [1–3]. A positron annihilation lifetime (PAL) has been measured to investigate the irradiation defects which primarily controlled thermal diffusivity in ceramics [4]. The correlation between PAL and thermal diffusivity was clarified with a-Al₂O₃ and AlN in doses of 0.01–0.02 dPa [5]. Nowadays, ceramics as alternative waste forms of glasses have been developed to immobilize actinide elements extracted from high-level radioactive nuclear wastes. Thermal properties of some appropriate ceramics (Y₆WO₁₂, Yb₆WO₁₂, Y₆UO₁₂) were reported as the waste form [6–8]. Since ceramic materials have the excellent characteristic, it is expected that these activities will be promoted more.

There have been considerable researches of zirconium alloys for the cladding tube, low-alloy steels for the reactor pressure vessel and stainless steels for the core structural materials. The behavior of hydrogen diffusion in zirconium oxide was reported [9] because the hydrogen in zirconium oxide affected the corrosion resistance of the cladding tube made of zirconium alloy. The hydrogen diffusion behavior in monoclinic and tetragonal zirconium oxides was estimated by performing electronic state calculation [10,11]. As a result, it was estimated that compression stress reduced the hydrogen diffusion coefficient by 40%/GPa and oxygen defects were thought to act as trapping sites for hydrogen [12]. Shortage of surveillance test specimens for irradiation embrittlement of the reactor pressure vessel steel for extension of plant lifetime is a problem and it is considered that small specimen test technology is useful. Since Charpy impact properties depend on the specimen size as well as the notch location of the weld joint [13,14], the effects of specimen size, V-notch location and chemical contents on impact properties of small specimen were evaluated. Kim et al. [15] were reported that the ductile to brittle transition temperature (DBTT) and the upper shelf energy of standard size specimen could be predicted from the empirical equations obtained from the sub-sized specimens. The effects of notch location and chemical contents on DBTT were almost independent of specimen size. It is believed that these further activities contribute to improvement in the reliability of nuclear systems.

Measurement and analysis of the thermal properties of nuclear fuels were performed by some groups. To explain the difference in the thermal conductivity degradation between mixed oxide (MOX) fuel and UO₂ fuel after burnup [16], the quasi-two phase material model was proposed [17]. The numerical result showed the gradual increase of the difference with burnup which was consistent with the results from the irradiation tests. Moreover, the results indicated that the inhomogeneity of Pu content in the MOX fuel could be one of the major reasons for the moderation of the thermal conductivity degradation of the MOX fuel. Actinide nitrides (AnN) are considered to be advanced fuel materials for FBR or target materials for an accelerator-driven system [18]. The thermochemical and thermophysical properties of AnN were reviewed [19,20]. Recently, the thermal and mechanical properties of neptunium mononitride (NpN) were investigated by first-principles calculations [21]. The obtained specific heat capacity reproduced the experimental data well. It was clarified that the specific heat capacity of NpN consisted of the lattice and electronic specific heat capacities and the contribution of the lattice dilatation. These activities will contribute to the development of the fuel performance code which estimated the fuel performance in case of operation or an accident.