The behavior of Nd(III)/Nd(0) redox reaction at a solid tungsten (W) and liquid cadmium (Cd) electrodes in the LiCl-KCl-NdCl 3 system at 773 K was analyzed by using a cyclic voltammetry. The Nd(III)/Nd(0) reaction in the solid W electrode was found to occur in two steps through the process of Nd(II) ion generation. As the analytic results of the maximum current value according to scan rates in the region of 0.05 V/s to 0.3 V/s, the Nd(III)/Nd(II) reaction showed reversible reaction characteristics over the whole range, while the Nd (II)/Nd(0) reaction showed the reversible reaction in the region of 0.05 V/s to 0.2 V/s, as well as the quasi-reversible in the range of 0.2 V/s or higher. Furthermore, the diffusion coefficients of Nd(III) ion and Nd(II) ion were calculated by using the maximum current value, which were considerably consistent with the previous results of other researchers. The reduction potential of the Nd (III)/Nd(0) reaction in the liquid Cd electrode was measured as approximately 0.54 V higher than that of the solid W electrode because the reduced Nd formed an Nd-Cd intermetallic compound with a liquid Cd electrode, lowering the activity of Nd. To confirm this mechanism, experiments using Cd-coated W electrodes were conducted by using cyclic voltammetry and chronopotentiometry. The experimental results showed that 6 Nd-Cd intermetallic compounds were present. When the ratio of Cd is the highest, NdCd 11 is formed, and the potential in this case was very similar to the potential at the liquid Cd electrode. Furthermore, the Gibbs free energy and activity coefficient of Nd-Cd intermetallic compounds were calculated by analyzing the results of chronopotentiometry. The results could thermodynamically explain the potential difference of the Nd(III)/Nd(0) reactions occurring between the solid W electrode and the liquid Cd electrode.

Background: Radiation dose rates in PRIDE facility is evaluated quantitatively for assessing radiation safety of workers because of large amounts of depleted uranium being handled in PRIDE facility. Even if direct radiation from depleted uranium is very low and will not expose a worker to significant amounts of external radiation. Materials and Methods: ORIGEN-ARP code was used for calculating the neutron and gamma source term being generated from depleted uranium (DU), and the MCNP5 code was used for calculating the neutron and gamma fluxes and dose rates. Results and Discussion: The neutron and gamma fluxes and dose rates due to DU on spherical surface of 30 cm radius were calculated with the variation of DU mass and density. In this calculation, an imaginary case in which DU density is zero was added to check the self-shielding effect of DU. In this case, the DU sphere was modeled as a point. In case of DU mixed with molten salt of 50-250 g, the neutron and gamma fluxes were calculated respectively. It was found that the molten salt contents in DU had little effect on the neutron and the gamma fluxes. The neutron and the gamma fluxes, under the respective conditions of 1 and 5 kg mass of DU, and 5 and 19.1 g.cm-3 density of DU, were calculated with the molten salt (LiCl+KCl) of 50 g fixed, and compared with the source term. As the results, similar tendency was found in neutron and gamma fluxes with the variation of DU mass and density when compared with source spectra, except their magnitudes. Conclusion: In the case of the DU mass over 5 kg, the dose rate was shown to be higher than the environmental dose rate. From these results, it is concluded that if a worker would do an experiment with DU having over 5 kg of mass, the worker should be careful in order not to be exposed to the radiation.