• Analytical Science and Technology
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• v.16 no.6
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• pp.431-437
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• 2003

Metalization yield of uranium oxide to uranium metal from lithium reduction process of spent pressurized water reactor (PWR) fuels was measured using thermogravimetric analyzer. A reduced metal produced in the process was divided into a solid and a powder part, and each metalization yield was measured. Metalization yield of the solid part was 90.7~95.9 wt%, and the powder being 77.8~71.5 wt% individually. Oxidation behaviour of the quartemary alloy was investigated to take data on the thermal oxidation stability necessary for the study on dry storage of the reduced metal. At $600{\sim}700^{\circ}C$, weight increments of alloy of Mo, Ru, Rh and Pd was 0.40~0.55 wt%. Phase change on the surface of the alloy was started at $750^{\circ}C$. In particular, Mo was rapidly oxidized and then the alloy lost 0.76~25.22 wt% in weight.

• Proceedings of the IEEK Conference
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• 2001.10a
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• pp.179-182
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• 2001

The chloride volatilization method for the recovery of zirconium and removal of uranium from zirconium containing metallic wastes formed in spent fuel reprocessing was studied using the simulated alloy waste, i.e. the mixture of Zr foil and UO$_2$/U$_3$O$_{8}$ powder. When the simulated waste was heated to react with chlorine gas at 350- l00$0^{\circ}C$, the zirconium metal changed to volatile ZrCl$_4$showing high volatility ratio (Vzr) of 99%. The amount of volatilized uranium increases at higher temperatures causing lowering of decontamination factor (DF) of uranium. This is thought to be caused by the chlorination of UO$_2$ with ZrCl$_4$vapor. The highest DF value of 12.5 was obtained when the reaction temperature was 35$0^{\circ}C$. Addition of 10 vol.% oxygen gas into chlorine gas was effective for suppressing the volatilization of uranium, while the volatilization ratio of zirconium was decreased to 68% with the addition of 20 vol.% oxygen. In the case of the mixture of Zr foil and U$_3$O$_{8}$, the V value of uranium showed minimum (44%) at 40$0^{\circ}C$ with chlorine gas giving the highest DF value 24.3. When the 10 vol.% oxygen was added to chlorine gas, the V value of zirconium decreased to 82% at $600^{\circ}C$, but almost all the uranium volatilized (Vu=99%), which may be caused by the formation of volatile uranium chlorides under oxidative atmosphere.ere.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.42-47
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• 2024

The electrochemical extraction of uranium in ternary low melting LiCl-KCl-CsCl eutectic on inert and reactive electrodes via different electrochemical techniques was investigated. It was established that the electrochemical reduction process of U(III) ions on the inert W electrode was irreversible and proceeded in one stage. On reactive liquid Ga and liquid Cd electrodes the reduction of uranium ions took place with the considerable depolarization with the formation of UGa₂, UGa₃ and UCd₁₁ intermetallic compounds. Thermodynamic characteristics of uranium compounds and alloys were calculated. The conditions for the extraction of uranium from the electrolyte in the form of alloys on both liquid reactive electrodes via potentiostatic electrolysis were found.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.225-225
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• 1998

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.4
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• pp.239-244
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• 2004

In order to enhance an oxidation resistance of the pure uranium metal under air condition, a small quantity of niobium(Nb) which is known to mitigate metal oxidation is added into uranium metal as an alloying element. A simulated metallic uranium alloy, U-Nb has been fabricated and then oxidized in the range of 200 to $300^{\circ}C$ under the environment of the pure oxygen gas. The oxidized quantity in terms of the weight gain(wt%) has been measured with the help of a thermogravimetric analyzer. The results show that the oxidation resistance of the U-Nb alloy is considerably enhanced in comparison with that of the pure uranium metal. It is revealed that the oxidation resistance of the former with the niobium content of 1, 2, 3, and 4 wt% is : 1) 1.61, 7.78, 11.76 and 20.14 times at the temperature of $200^{\circ}C$ ; 2) 1.45, 5.98, 10.08 and 11.15 times at $250^{\circ}C$ ; and 3) 1.33, 4.82, 8.87 and 6.84 times at $300^{\circ}C$ higher than that of the latter, respectively. Besides, it is shown that the activation energy attributable to the oxidation is 17.13~21.92 kcal/mol.

• Proceedings of the Materials Research Society of Korea Conference
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• 2003.03a
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• pp.238-238
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• 2003

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.395-400
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• 2003

For the long term storage safety study of the metallic spent fuel, U-Nb, U-Ti, U-Ni, U-Zr, and U-Hf simulated metallic uranium alloys, known as corrosion resistant alloys, were fabricated and oxidized in oxygen gas at $200^{\circ}C~300^{\circ}C$. Simulated metallic uranium alloys were more corrosion resistant than pure uranium metal, and corrosion resistance increases Nb, Ni, Ti in that order. The oxidation rates of uranium alloys determined and activation energy was calculated for each alloy. The matrix microstructure of the test specimens were analyzed using OM, SEM, and EPMA. It was concluded that Nb was the best acceptable alloying elements for reducing corrosion of uranium meta] considered to suitable as candidate.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1855-1862
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• 2023

Graphite crucibles are used for melting uranium and its alloys in VIM furnace. Various coating materials namely Al₂O₃, ZrO₂, MgO etc. are applied on the inner surface of the crucibles using paint brush or thermal spray technique to mitigate U-C interaction. These leads to significant amount of carbon pick-up in uranium. In this study, the attempts are made to develop multilayer coatings comprising of SiC/Y₂O₃ and Mo/Y₂O₃ on graphite to study the feasibility of minimizing U-C interaction. The parameters are optimized to prepare SiC coating of about 70㎛ thickness using CVD technique on graphite coupons and subsequently Y₂O₃ coating of about 250㎛ thickness using plasma spray technique. Molybdenum and Y₂O₃ layers were deposited using plasma spray technique with 70㎛ and 250㎛ thickness, respectively. Interaction studies of the coated graphite with molten uranium at 1450℃ for 20 min revealed that Y₂O₃ coating with SiC interlayer provides physical barrier for uranium-graphite interaction, however, this led to the physical separation of coating layer. Y₂O₃ coating with Mo interlayer provided superior barrier effect showing no degradation and the coatings remained intact after interaction tests. Therefore, the Mo/Y₂O₃ coating was found to be a promising solution for minimizing carbon pick-up during uranium/uranium alloy melting.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.4
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• pp.261-267
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• 2010

The pyrometallurgical nuclear fuel recycle process, called pyroprocessing, has been known as a promising nuclear fuel recycling technology. Pyroprocessing technology is crucial to advanced nuclear systems due to increased nuclear proliferation resistance and economic efficiency. The basic concept of pyroprocessing is group actinide recovery, which enhances the nuclear proliferation resistance significantly. One of the key steps in pyroprocessing is "electrowinning" which recovers group actinides with lanthanide from the spent nuclear fuels. In this study, a vertical cadmium distiller was manufactured. The evaporation rate of pure cadmium in vertical cadmium distiller varied from 12.3 to $40.8g/cm^2/h$ within a temperature range of 773 923 K and pressure below 0.01 torr. Uranium - cadmium alloy was fabricated by electrolysis using liquid cadmium cathode in a high purity argon atmosphere glove box. The distillation behavior of pure cadmium and cadmium in uranium - cadmium alloy was investigated. The distillation behavior of cadmium from this study could be used to develop an actinide recovery process from a liquid cadmium cathode in a cadmium distiller.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.608-609
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• 2018