• Proceedings of the Korean Radioactive Waste Society Conference
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• 2015.10a
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• pp.105-106
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• 2015

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1955-1963
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• 2021

As the IAEA has applied integrated safeguards and a state level approach to member states, the importance of national inspection has increased. However, the requirements for national inspection for some member states are different from the IAEA safeguards. In particular, the national inspection for the ROK requires on-site U concentration analysis due to a domestic notification. This research proposes an on-site U concentration analysis (OUCA) method for UO₂ pellets using gamma spectroscopy to satisfy the domestic notification requirement. The OUCA method calculates the U concentration of UO₂ pellets using the measured net X-ray counts and declared ²³⁵U enrichment. This research demonstrates the feasibility of the OUCA method using both MCNP simulation and experiment. It simulated and measured the net X-ray counts of different UO₂ pellets with different U concentrations and ²³⁵U enrichments. The simulated and measured net X-ray counts were fitted to polynomials as a function of U concentration and ²³⁵U enrichment. The goodness-of-fit results of both simulation and experiment demonstrated the feasibility of the OUCA method.

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1300-1309
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• 2023

Simulated debris was synthesized using UO₂, Zr, and stainless steel and a heat treatment method under inert or oxidizing conditions. The primary U solid phase of the debris synthesized at 1473 K under inert conditions was UO₂, whereas a (U, Zr)O₂ solid solution formed at 1873 K. Under oxidizing conditions, a mixture of U₃O₈ and (Fe, Cr)UO₄ phases formed at 1473 K, whereas a (U, Zr)O_2+x solid solution formed at 1873 K. The leaching behavior of the fission products from the simulated debris was evaluated using two methods: the irradiation method, for which fission products were produced via neutron irradiation, and the doping method, for which trace amounts of non-radioactive elements were doped into the debris. The dissolution behavior of U depended on the properties of the debris and aqueous solution for immersion. Cs, Sr, and Ba leached out regardless of the primary solid phases. The leaching of high-valence Eu and Ru ions was suppressed, possibly owing to their solid-solution reaction with or incorporation into the uranium compounds of the simulated debris.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.1036-1042
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• 2020

Fast-Neutron Multiplicity Counter based on Liquid Scintillator Detector can directly measure the fast neutron multiplicity emitted by UO₂ rod. HPGe gamma spectrometer; which has superior energy resolution; is routinely used for the gamma energy spectrum measurement. Combing Fast-Neutron Multiplicity Counter and HPGe γ-spectrometer, the n/γ joint measurement system is developed. The fast neutron multiplicity and gamma energy spectrum of UO₂ rod assemblies under different conditions are measured by the n/γ joint measurement system. The induced fission rate and the ²³⁵U abundance, thereby the ²³⁵U mass; can be obtained for UO₂ rod assemblies. The ²³⁵U mass deviation of the measured value from the reference value is less than 5%. The results show that the n/γ joint measurement system is effective and applicable in the measurement of the ²³⁵U mass in samples.

• Applied Chemistry for Engineering
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• v.9 no.3
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• pp.388-393
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• 1998

The effects of surface area of $UO_2$ pellets($s:0.034{\sim}0.282cm^2/g-UO_2$), concentration of nitric acid(1.5~10N) and temperature($40{\sim}105^{\circ}C$) on the dissolution rate were investigated in this experiment. The results showed that dissolution rate of $UO_2$ pellets was in direct proportion to effective surface area(s) and well fit to linear plot in high surface area. The dissolution rate was expressed as dc/dt = 10.6s at $90^{\circ}C$, [$HNO_3$]=8N, and proportioned to the power 1.42 of $HNO_3$ concentration. And also, the results showed that dissolution rate was sharply increased according to temperature increase in temperature below $90^{\circ}C$, but decreased in temperature above $90^{\circ}C$. Activation energy(E) was evaluated to be 36.3KJ/mol.

• Journal of the Korean Ceramic Society
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• v.42 no.9 s.280
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• pp.618-623
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• 2005

HTGR (High Temperature Gas-Cooled Reactor) is highlighted to next generation power plant for producing the clean hydrogen gas. In this study, the spherical $UO_2$ kernel via $UO_3$ gel particles was prepared by the sol-gel process. Raw material of slightly Acid Deficient Uranyl Nitrate (ADUN) solution, which has pH = 1.10 and $[NO_3]/[U]$ mole ratio = 1.93, was obtained from dissolution of $U_3O_8$ powder with conc.-$HNO_3$. The surface of these spherical $UO_3$ gel particles, which was prepared from the broth solution, consisted of 1 M-uranium, 1 M-HMTA, and urea, were covered with the fine crystallite aggregates, and these particles were so hard that crushed well. But the other $UO_3$ gel particles prepared with the broth solution, consisted of 2 M-uranium, 2 M-HMTA, and urea, have soft surface characteristics and an amorphous phase. This type of $UO_3$ gel particles is some chance of doing possibility of high density from the compaction. The amorphous $UO_3$ gel particles was converted to $U_3O_8$ and then $UO_2$ by calcination at $600^{\circ}C\;in\;4\%\;-\;H_2\;+\;N2$ atmosphere.

• Journal of the Korean Ceramic Society
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• v.47 no.6
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• pp.590-597
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• 2010

HTGR using a TRISO coated particles as nuclear raw fuel material can be used to produce clean hydrogen gas and process heat for a next-generation energy source. For these purposes, a TRISO coated particle was prepared with 3 pyro-carbon (buffer, IPyC, and OPyC) layers and 1 silicone carbide (SiC) layer using a CVD technique on a spherical $UO_2$ kernel surface as a fissile material. In this study, a spherical $UO_2$ particle was prepared using a modified sol-gel method with a vibrating nozzle system, and TRISO coating fabrication was carried out using a fluidized bed reactor with coating gases, such as acetylene, propylene, and methyltrichlorosilane (MTS). As the results of this study, a spherical $UO_2$ kernel with a sphericity of 1+0.06 was obtained, and the main process parameters in the $UO_2$ kernel preparation were the well-formed nature of the spherical ADU liquid droplets and the suitable temperature control in the thermal treatment of intermediate compounds in the ADU, $UO_3$, and $UO_2$ conversions. Also, the important parameters for the TRISO coating procedure were the coating temperature and feed rate of the feeding gas in the PyC layer coating, the coating temperature, and the volume fraction of the reactant and inert gases in the SiC deposition.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1156-1163
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• 2020

Fuel temperature coefficient (FTC) of PuO₂+ZrO₂ (ROX) fueled LWR cell is analyzed neutronically with reactor- and weapons-grade plutonium fuels in comparison with a U-free PuO₂+ThO₂ (TOX), and a conventional MOX fuel cells. The FTC value of a ROX fueled LWR is smaller compared to a TOX or a MOX fueled LWRs and becomes extremely positive especially, at EOL. This is because when fuel temperature is increased, thermal neutron spectrum is shifted to harder, which is extreme at EOL in ROX fuel than that in TOX and MOX fuels. Consequently at EOL, ²³⁹Pu and ²⁴¹Pu contributes to positive fuel temperature reactivity (FTR) in ROX fuel, while they have negative contribution in TOX and MOX fuels. The FTC problem of ROX fuel is mitigated by additive ThO₂, UO₂ or Er₂O₃. In ROX-additive fuel, the atomic density of fissile Pu becomes more than additive free ROX fuel especially at EOL, which is the main cause to improve the FTC problem. The density of fissile Pu is more effective to decrease the thermal spectrum shifts with increase the fuel temperature than additive ThO₂, UO₂ or Er₂O₃ in ROX fuel.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2011.05a
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• pp.111-112
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• 2011

• Economic and Environmental Geology
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• v.56 no.5
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• pp.603-618
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• 2023

This study focused on evaluating the suitability of the WRK (waste repository Korea) bentonite as a buffer material in the SNF (spent nuclear fuel) repository. The U (uranium) adsorption/desorption characteristics and the adsorption mechanisms of the WRK bentonite were presented through various analyses, adsorption/desorption experiments, and kinetic adsorption modeling at various pH conditions. Mineralogical and structural analyses supported that the major mineral of the WRK bentonite is the Ca-montmorillonite having the great possibility for the U adsorption. From results of the U adsorption/desorption experiments (intial U concentration: 1 mg/L) for the WRK bentonite, despite the low ratio of the WRK bentonite/U (2 g/L), high U adsorption efficiency (>74%) and low U desorption rate (<14%) were acquired at pH 5, 6, 10, and 11 in solution, supporting that the WRK bentonite can be used as the buffer material preventing the U migration in the SNF repository. Relatively low U adsorption efficiency (<45%) for the WRK bentonite was acquired at pH 3 and 7 because the U exists as various species in solution depending on pH and thus its U adsorption mechanisms are different due to the U speciation. Based on experimental results and previous studies, the main U adsorption mechanisms of the WRK bentonite were understood in viewpoint of the chemical adsorption. At the acid conditions (＜pH 3), the U is apt to adsorb as forms of UO₂²⁺, mainly due to the ionic bond with Si-O or Al-O(OH) present on the WRK bentonite rather than the ion exchange with Ca²⁺ among layers of the WRK bentonite, showing the relatively low U adsorption efficiency. At the alkaline conditions (>pH 7), the U could be adsorbed in the form of anionic U-hydroxy complexes (UO₂(OH)₃^-, UO₂(OH)₄^2-, (UO₂)₃(OH)₇^-, etc.), mainly by bonding with oxygen (O^-) from Si-O or Al-O(OH) on the WRK bentonite or by co-precipitation in the form of hydroxide, showing the high U adsorption. At pH 7, the relatively low U adsorption efficiency (42%) was acquired in this study and it was due to the existence of the U-carbonates in solution, having relatively high solubility than other U species. The U adsorption efficiency of the WRK bentonite can be increased by maintaining a neutral or highly alkaline condition because of the formation of U-hydroxyl complexes rather than the uranyl ion (UO₂²⁺) in solution,and by restraining the formation of U-carbonate complexes in solution.