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- Author or Editor: H. Ahn x
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Abstract
The distillation behaviour of cadmium at a reduced pressure was investigated to develop an actinide recovery process from a liquid cadmium cathode in a laboratory scale cadmium distiller. The apparent evaporation rate of cadmium increased with an increasing temperature whereas the rate decreased with an increasing vacuum pressure. The evaporation rate of cadmium varied within 9.7–40 g/cm2/h in the temperature range of 500–650 °C and pressure range of 0.5–10 Torr (0.0667–1.33 kPa). The theoretical values calculated by the Hertz–Langmuir relation were much higher than experimentally measured values. The deviation was compensated by an evaporation coefficient (α) obtained empirically. About 0.02–0.20 wt% of residue was left in the crucible after distillation and found to be CdO. It could be concluded that the temperature range of 500–650 °C is favourable for the cadmium distillation process if residual eutectic salt does not exist in the cadmium alloy surface.
Abstract
It is important to increase a throughput of the salt removal process from uranium deposits which is generated on the solid cathode of electro-refiner in pyroprocess. In this study, it was proposed to increase the throughput of the salt removal process by the separation of the liquid salt prior to the distillation of the LiCl–KCl eutectic salt from the uranium deposits. The feasibility of liquid salt separation was examined by salt separation experiments on a stainless steel sieve. It was found that the amount of salt to be distilled could be reduced by the liquid salt separation prior to the salt distillation. The residual salt remained in the deposits after the liquid salt separation was successfully removed further by the vacuum distillation. It was concluded that the combination of a liquid salt separation and a vacuum distillation is an effective route for the achievement of a high throughput performance in the salt separation process.
Abstract
A computational fluid dynamics (CFD)-based multiphysics model of a molten-salt electrorefiner is presented for the computational electro-fluid analysis. A target model of the electrorefining cell presented here has a structure arranged concentrically with the cathode annulus surrounding a rotating cruciform anode inside it. This comprehensive approach of a multiphysics model solves the convective and diffusive transport of ionic uranium and allows for a prediction of the concentration present in the LiCl–KCl eutectic electrolyte between the electrodes at a current driven condition. The coupling of the local overpotential distribution and uranium concentration gradient makes it possible to predict the local current density distribution at the electrode surfaces.
Abstract
This paper describes ongoing research into the multi-physics model development of an electrorefining process for the treatment of spent nuclear fuel. A forced convection of molten eutectic (LiCl–KCl) electrolyte in an electrorefining cell is considered to establish an appropriate electro-fluid model within the 3-dimensional framework of a conventional computational fluid dynamic model. This computational platform includes the electrochemical reaction rate of charge transfer kinetics which is described by a Butler–Volmer equation, while mass transport is considered using an ionic transport equation. The coupling of the local overpotential distribution and uranium concentration gradient makes it possible to predict the local current density distribution at the electrode surfaces.
Summary
It is impossible to detect 14C and 3H by direct methods such as γ-spectroscopy because they are pure b-emitters and thus they are classified as hard to measure nuclides (HTM). In this paper the analysis results of 14C and 3H in the low level radioactive wastes (LLWs), including spent ion exchange resin, evaporated bottom and sludge are presented. The LLWs were generated by three nuclear power plants (NPPs), in Korea all with pressurized water type reactors (PWRs). A simultaneous separation procedure for 14C and 3H in LLWs was established by wet oxidation-acid stripping. A liquid scintillation analyzer was used for the measurement of 14C and 3H. It was found that the recovery of 14C and 3H was 82-99 and 78-103%, respectively, by wet oxidation-acid stripping with diluted standard solutions. At the lowest injection of 14C and 3H, i.e., at 1.44 Bq for 14C and 1.22 Bq for 3H, the minimum detectable activity (MDA) of 14C and 3H was calculated as 0.88 and 0.78 Bq/g, respectively, for the minimum allowable sample weight, using wet oxidation and 16 wt% H2SO4 acid. By the wet oxidation-16 wt% H2SO4 stripping method no interfering nuclides were detected in the trapping solution of 14CO2 and the distillate of 3H. The activity concentration range of 14C in the analyzed samples, i.e., spent ion exchange resin, evaporated bottom and sludge, was 0.17-110,000, 8.4-1380 and 0.1-10,006 Bq/g, respectively, and that of 3H in the same was from no detectable to 769, 134-14,383 and 0.7-4820 Bq/g, respectively.
This study was aimed at evaluating the efficacy of different mineral separation procedures to validate the EN1788 (2001) European Union standard protocol for better identification of irradiated fish and shellfish. The silicate minerals were isolated with physical density separation method from two types of non-irradiated freeze-dried fish and shellfish that included Pacific saury (Cololabis saira), mackerel (Scomber japonicus), shrimp (Penaeidae metapenaeus), and mussel (Mytilus coruscus). Radiation-specific thermoluminescence (TL) peaks (glow curve 1) were observed between 150–250 °C. The peaks are typical for the irradiated food; despite the samples being not irradiated. Apparently it showed that the isolated minerals were contaminated with organic materials such as bone, etc. Acid-hydrolysis digestion was employed to remove the possible contaminants. The minerals obtained through alternative pre-treatment showed no TL curves in radiation specific temperature range. Moreover, acid hydrolysis extraction resulted in producing higher mineral yields and lower background luminescence. Results were also confirmed by calculating TL ratios (glow curve 1/glow curve 2) to confirm the irradiation history of samples. Furthermore, different time and temperature treatments on TL intensity of irradiated standard quartz (SiO2) minerals showed that the acid-hydrolysis can be adjusted to 50 °C and 3 h for better luminescence determinations.
Abstract
A pyrochemical processing has become one of the potential technologies for a future nuclear fuel cycle. An integrated multi-physics simulation and electrotransport model of a molten-salt electrolytic process are proposed and discussed with respect to the recovery of pure uranium when using thermochemical data. This study has been performed to provide information for diffusion boundary layers between the molten salt (KCl-LiCl) and electrode. The diffusion-controlled electrochemical model demonstrate a prediction of the electrotransport behaviors of LWR spent fuel as a function of the time up to the corresponding electrotransport satisfying a given applied current based on a galvanostatic electrolysis.