• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.309-313
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• 2018

In this study, we prepared Li-K-Cd alloy, which meets the requirement of eutectic ratio of Li:K, to maintain the operating temperature of the drawdown process at $500^{\circ}C$ and to achieve the reuse of LiCl-KCl molten salt. The prepared Li-K-Cd alloys were added to LiCl-KCl salt bearing U and Nd at $500^{\circ}C$ to investigate the removal of $UCl_3$ in the salt. The reduction of $UCl_3$ in the salt was examined by measuring the OCP value of salt and analyzing the salt composition by ICP-OES. Reduction was also visually confirmed by change of salt color from dark purple to white. The experimental results reveal that the prepared Li-K-Cd alloy has reductive extractability for $UCl_3$ in salt. By improving the preparation method, the Li-K-Cd alloy can be applied to the drawdown process.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.2
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• pp.161-176
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• 2021

In this study, the electrochemical behavior of Sm on the binary liquid Al-Ga cathode in the LiCl-KCl molten salt system is investigated. First, the co-reduction process of Sm(III)-Al(III), Sm(III)-Ga(III), and Sm(III)-Ga(III)-Al(III) on the W electrode (inert) were studied using cyclic voltammetry (CV), square-wave voltammetry (SWV) and open circuit potential (OCP) methods, respectively. It was identified that Sm(III) can be co-reduced with Al(III) or Ga(III) to form Al_zSm_y or Ga_xSm_y intermetallic compounds. Subsequently, the under-potential deposition of Sm(III) at the Al, Ga, and Al-Ga active cathode was performed to confirm the formation of Sm-based intermetallic compounds. The X-ray diffraction (XRD) and scanning electron microscopy-energy dispersive spectroscopy (SEM-EDS) analyses indicated that Ga₃Sm and Ga₆Sm intermetallic compounds were formed on the Mo grid electrode (inert) during the potentiostatic electrolysis in LiCl-KCl-SmCl₃-AlCl₃-GaCl₃ melt, while only Ga₆Sm intermetallic compound was generated on the Al-Ga alloy electrode during the galvanostatic electrolysis in LiCl-KCl-SmCl₃ melt. The electrolysis results revealed that the interaction between Sm and Ga was predominant in the Al-Ga alloy electrode, with Al only acting as an additive to lower the melting point.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.347-357
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• 2018

The applicability of room temperature ionic liquids (RTILs), 1-alkyl-3-methylimidazolium bis(trifluoromethylsulfonyl)imide ([$C_nmim$] [$Tf_2N$]), was investigated for the extraction of Am(III) and Eu(III) from nitric acid using n-octyl(phenyl)-N,N-diisobutyl carbamoylmethyl phosphine oxide (CMPO) and tri-n-butylphosphate (TBP) as extractants. The distribution ratios of Am(III) and Eu(III) in CMPO-TBP/[$C_nmim$][$Tf_2N$] were measured as a function of various parameters such as the concentrations of nitric acid, CMPO, and TBP. The results were compared with those obtained in CMPO-TBP/n-dodecane (n-DD). With comparable concentrations of the extractants, the distribution ratios obtained with RTILs were much higher than those obtained with n-DD. It was observed that the extraction efficiency was less for Eu(III) than for Am(III). The extraction of Am(III) and Eu(III) decreased with increases in the feed acidity for all three RTILs. The results suggest that the extraction of Am(III) and Eu(III) by CMPO in RTILs from nitric acid proceeds through the cation-exchange mechanism. The distribution ratios of Am(III) and Eu(III) increased with increases in the concentration of CMPO for all three RTILs. A linear regression analysis of the extraction data resulted in a straight line with a slope of about 3, suggesting the involvement of 3 molecules of CMPO during the extraction process.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.2
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• pp.125-131
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• 2009

An electrolytic system (zinc anode-gallium cathode) was setup to evaluate the performance of several stirrers prepared for this study, where stirrers have been used to prevent uranium from forming dendrite on the cathode in pyrochemical process. In the case of no-stirring condition, zinc dendrites began to grow on the gallium surface in 1 hour and some dendrite grew out of the cathode crucible around 6 hours. When a rectangular stirrer or a tilt stirrer was rotated, at 40${\sim}$150 rpm, to mix the liquid gallium cathode, dendritic growth of zinc metal was prevented irrespective of revolution speed, but some of the deposits overflowed out of the cathode crucible owing to the large centrifugal forces at 150 rpm. The harrow stirrer did not nearly retard the dendrite growth at 40 rpm, but the dendrite growth was retarded at higher than 100 rpm and the zinc deposits also did not overflow at 150 rpm. Pounder could also prevent the dendrite growth to some extent but it had some difficulties in operation compared with other types of stirrers.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.2
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• pp.109-114
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• 2009

Absorption spectra for a quantitative analysis of actinide elements such as U(VI) and Pu(V) were measured by using a liquid waveguide capillary cell (LWCC) which has an optical path length of 1.0 meter. In order to investigate radioactive elements, a LWCC is installed in a glove box and is coupled to a spectrophotometer with optical fibers. Limits of detection (LOD) for the system were determined as 0.74 and 0.35 M with molar absorption coefficients of 8.14${\pm}$0.07 (414 nm) and 17.00${\pm}$0.16 (569 nm) $M^{-1}cm^{-1}$ for U(VI) and Pu(V) ions, respectively. The measured LOD values are about 30 times more sensitive when compared to those achievable by using a conventional quartz cell with an optical path length of 1.0 cm. As an application with an enhanced sensitivity, a quantitative analysis for micromolar concentrations of Pu(V) has been performed to decrease the uncertainty in the formation constant of the Pu(VI)-OH complex.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.1
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• pp.9-17
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• 2010

The LCC (Liquid Cadmium Cathode) structure to be developed for inhibiting the formation and growth of the uranium dendrite has been known as a key part in the electrowinning process for the simultaneous recovering of uranium and TRU (TRans Uranium) elements from spent fuels. A zinc-gallium (Zn-Ga) experimental system which is able to be functional in aqueous condition and normal temperature has been set up to observe the formation and growth phenomena of the metal dendrites on liquid cathode. The growth of the zinc dendrites on the gallium cathode and the performance of the existing stirrer type and pounder type cathode structure were observed. Although the mechanical strength of the dendrites appeared to be weak in the electrolyte and easily crashed by the various cathode structures, it was difficult to effectively submerge the dendrite into the bottom of the liquid cathode. Based on the results of the aqueous phase experiments, a lab-scale electrowinning experimental apparatus which are applicable to the development of LCC srtucture for the electrowinning process was established and the performance tests of the different types of LCC structure were conducted to prohibit the uranium dendrite growth on LCC surface. The experimental results of the stirrer type LCC structures have shown that they could not effectively remove the uranium dendrites growing at the inner side of the LCC crucible and the performances of the paddle and harrow type LCC structure were similar. Therefore a mesh type LCC structure was developed to push down the uranium dendrites to the bottom of the LCC crucible growing on the LCC surface and at the inner side of the crucible. From the experimental results for the performance test of the mesh type LCC structure, the uranium was recovered over 5 wt% in cadmium without the growth of uranium dendrites. After completion of the experiments, solid precipitates of the bottom of the LCC crucible were identified as an intermetallic compound (UCd11) by the chemical analysis.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.1
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• pp.103-111
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• 2020

Pulling-type cutting devices, which use a diamond wire saw, have been used generally for cutting concrete structures. In this study, a pushing-type cutting device with a collection cover was developed by overcoming the disadvantages of pulling-type devices. In this device, dry or liquid methods can be selected to cool frictional heat. Operation and leakage tests of the dust generated during the dismantling of a concrete structure were carried out, confirming the suitable operation of the fabricated cutting device; the leakage rate was approximately 1.7%. For a conservative evaluation, the internal dose of workers was estimated in dismantling the core center part of biological shield concrete with a specific activity of 99.5 Bq·g^-1. The committed effective dose per worker was 0.25 mSv. The developed cutting device contributed to reducing radioactive concrete waste and minimizing worker exposure due to its easy installation. Therefore, it can be utilized as a cutting apparatus for dismantling not only reinforced concrete structures but also radioactive biological shield concrete in nuclear power plant decommissioning efforts.

• Korean Chemical Engineering Research
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• v.47 no.3
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• pp.303-309
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• 2009

The effects of zeolite type catalysts addition on the thermal decomposition of low density polyethylene(LDPE) and ethylene vinyl acetate(EVA) resin have been studied in a thermal analyzer(TGA, DSC) and a small batch reactor. The zeolite type catalysts tested were natural zeolite, FCC catalyst, used FCC catalyst, and catalyst A. As the results of TGA experiments, addition of antifogging-agent decreased the pyrolysis point to $250^{\circ}C$, but addition of longevity-agent and clay reduced the pyrolysis rate in EVA resin. Addition of the zeolite type catalysts in the LDPE resin increased the pyrolysis rate in the order of catalyst A > used FCC catalyst > natural zeolite > LDPE resin. Addition of the zeolite type catalysts in the EVA resin increased the pyrolysis rate in the order of used FCC catalyst > natural zeolite > catalyst A > EVA resin. In the DSC experiments for LDPE resin, addition of zeolite type catalysts decreased the melting point and the heat of pyrolysis reaction in the order of catalyst A > used FCC catalyst > natural zeolite> LDPE resin. In the batch system experiments, the mixing of natural zeolite enhanced the yield of liquid fuel oil.

• Particle and aerosol research
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• v.15 no.1
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• pp.1-13
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• 2019

Current desulfurization and denitrification technologies have reached a considerable level in terms of reduction efficiency. However, when compared with the simultaneous reduction technology, the individual reduction technologies have issues such as economic disadvantages due to the difficulty to scale-up apparatus, secondary pollution from wastewater/waste during the treatment process, requirement of large facilities for post-treatment, and increased installation costs. Therefore, it is necessary to enable practical application of simultaneous SOx and NOx treatment technologies to remove two or more contaminants in one process. The present study analyzes a technology capable of maintaining simultaneous treatment of SOx and NOx even at low temperatures due to the electrochemically generated strong oxidation of the wet-pulse complex system. This system also reduces unreacted residual gas and secondary products through the wet scrubbing process. It addresses common problems of the existing fuel gas treatment methods such as SDR, SCR, and activated carbon adsorption (i.e., low treatment efficiency, expensive maintenance cost, large installation area, and energy loss). Experiments were performed with varying variables such as pulse voltage, reaction temperature, chemicals and additives ratios, liquid/gas ratio, structure of the aeration cleaning nozzle, and gas inlet concentration. The performance of individual and complex processes using the wet-pulse discharge reaction were analyzed and compared.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.1
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• pp.19-27
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• 2007

Combined method of extraction chromatography and liquid scintillation analysis was established for determinating $^{55}Fe\;and\;^{63}Ni$ radioactivity in solid samples. Activated concrete wastes generated from dismantling KRR-2 were analysed. The sequential separation including precipitation and extraction chromatography resulted in the above 90% chemical recoveries of Fe. Above 62% recoveries of Ni were obtained by this procedure exception to 43.6 and 46.5% recoveries. The seperation and counting procedure was also confirmed with spiked samples of known quantity. The measured and spiked quantity were agreed with the 3.7% and 0.7% variations in the $^{55}Fe\;and\;^{63}Ni$ experiments, respectively. The radioactivities of $^{55}Fe$ in the dismantled concretes are shown from below MDA to maximum 362 Bq/g. The radioactivities of $^{63}Ni$ in all concrete samples are below MDA. The $^{63}Ni$ doesn't exist in dismantled concretes from KRR-2. The radioactivity of $^{55}Fe$ is decreased rapidly as the sampling depth is increased from the concrete surface.