• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.2
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• pp.63-71
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• 2011

A large volume of uranium electrokinetic leachate has been generated during the electrokinetic decontamination to remove uranium from contaminated soil. The treatment technology for the reuse of the uranium leachate was developed. The concentration of uranium in the generated uranium leachate was 180 ppm and concentrations of Mg(II), K(I), Fe(II), and Al(III) ions ranged from 20 ppm to 1,210 ppm. The treatment process for uranium leachate consisted mainly of mixing and cohesion, precipitation, concentration, and filtration. In order to obtain the pH=11 of a precipitate solution, the calcium hydroxide needs to be 3.0g/100ml and the sodium hydroxide needed to be 2.7g/100ml. The results of several precipitation experiments showed that a mixture of NaOH+0.2g alum+0.15g magnetite was an optimal precipitant for filtration. The average particle size of precipitate with NaOH+alum+0.15g magnetite was $600\;{\mu}m$. Because the total value of metal concentrations in supernatant at pH=9 was the smallest, sodium hydroxide should be added with 0.2g alum and 0.15g magnetite for pH=9 of leachate.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.3
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• pp.211-221
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• 2016

Waste salt generated from a pyro-processing for the recovery of uranium and transuranic elements has high volatility at vitrification temperature and low compatibility in conventional waste glasses. For this reason, KAERI (Korea Atomic Energy Research Institute) suggested a new method to de-chlorinate waste salt by using an inorganic composite named SAP ($SiO_2-Al_2O_3-P_2O_5$). In this study, the de-chlorination behavior of waste salt and the microstructure of consolidated form were examined by adding $B_2O_3$ and $Fe_2O_3$ to the original SAP composition. De-chlorination behavior of metal chloride waste was slightly changed with given compositions, compared with that of original SAP. In the consolidated forms, the phase separation between Si-rich phase and P-rich phase decreases with the amount of $Al_2O_3$ or $B_2O_3$ as a connecting agent between Si and P-rich phase. The results of PCT (Product Consistency Test) indicated that the leach-resistance of consolidated forms out of reference composition was lowered, even though the leach-resistance was higher than that of EA (Environmental Assessment) glass. From these results, it could be inferred that the change in the content of Al or B in U-SAP affected the microstructure and leach-resistance of consolidated form. Further studies related with correlation between composition and characteristics of wasteform are required for a better understanding.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.2
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• pp.127-137
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• 2019

Zr electrorefining is demonstrated herein using Zirlo tubes in a chloride-fluoride mixed molten salt in the presence of $AlF_3$. Cyclic voltammetry reveals a monotonic shift in the onset of metal reduction kinetics towards positive potential and an increase in intensity of the additional peaks associated with Zr-Al alloy formation with increasing $AlF_3$ concentration. Unlike the galvanostatic deposition mode, a radial plate-type Zr growth is evident at the top surface of the salt during Zr electrorefining at a constant potential of -1.2 V. The diameter of the plate-type Zr deposit gradually increases with increasing $AlF_3$ concentration. Scanning electron microscopy-energy-dispersive X-ray spectroscopy (SEM-EDX) and X-ray photoelectron spectroscopy (XPS) analyses for the plate-type Zr deposit show that trace amount of Al is incorporated as Zr-Al alloys with different chemical compositions between the top and bottom surface of the deposit. Addition of $AlF_3$ is effective in lowering the residual salt content in the deposit and in improving the current efficiency for Zr recovery.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.279-298
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• 2019

This study comprehensively addresses recent progress at KAERI in waste treatment technology to cope with waste produced by pyroprocessing, which is used to effectively manage spent fuel. The goal of pyroprocessing waste treatment is to reduce final waste volume, fabricate durable waste forms suitable for disposal, and ensure safe packaging and storage. KAERI employs grouping of fission products recovered from process streams and immobilizes them in separate waste forms, resulting in product recycling and waste volume minimization. Novel aspects of KAERI approach include high temperature treatment of spent oxide fuel for the fabrication of feed materials for the oxide reduction process, and fission product concentration or separation from LiCl or LiCl-KCl salt streams for salt recycling and higher fission-product loading in the final waste form. Based on laboratory-scale tests, an engineering-scale process test is in progress to obtain information on the performance of scale-up processes at KAERI.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.3
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• pp.255-263
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• 2006

Vol-oxidizer is a device to convert $UO_2$ pellets into $U_3O_8$ powder and to feed a homogeneous powder into a Metal Conversion Reactor in the ACP(Advanced Spent Fuel Conditioning Process). In this paper, we propose a design model of the vol-oxidizer, develop the new vol-oxidizer with a capacity of 20 kg HM/batch in $UO_2$ pellets, and conduct a verification for the device. Design considerations include the internal structure, the capacity, the heating position of the device, and the size. The dimensions of the new vol-oxidizer are decided by the design model. We determine a permeability test of the $U_3O_8$ measuring the temperature distribution, and the volume of $UO_2$ and $U_3O_8$. We manufactured the new vol-oxidizer for a 20 kg HM/batch in $UO_2$ pellets, and then analyzed the characteristics of the $U_3O_8$ powder for the verification. The experimental results show that the permeability of the $U_3O_8$ throughout mesh enhance more than old vol-oxidizer, the oxidation time takes only 8 hours when compared with the 13 hours of the old device, and the average distribution of particle size is $40{\mu}m$. The capacities of new vol-oxidizer for a 20 kg HM/batch in $UO_2$ pellets were agree well with the predictions of design model.

• 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.