• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.674-680
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• 2020

We used MCNP6 computer code to model HTR-10 core reactor. We used two types of fuel; UO₂ and (Th+Pu)O₂ mixture. We determined the critical height at which the reactor approached criticality in both two cases. The neutronic and burnup parameters were investigated. The results indicated that the core fueled with mixed (Th+Pu)O₂, achieved about 24% higher fuel cycle length than the UO₂ case. It also enhanced safeguard security by burning Pu isotopes. The results were compared with previously published papers and good agreements were found.

• Korean Chemical Engineering Research
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• v.57 no.3
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• pp.368-371
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• 2019

Using first principles calculations we investigated the thermomechanical stability of spent nuclear fuels (SNF), especially how mechanical properties of $UO_2$, such as, bulk, shear and Young's moduli and Poisson's ratio vary through alpha-decay of U into Th with generation of He gas. Our results indicate that substitution of U by Th through alpha decay ($U_{1-x}Th_xO_2$) does not significantly affect the stability of the grain in a fuel matrix. In addition, we studied the transport properties of He in and boundaries of the $U_{1-x}Th_xO_2$ grain. Helium preferentially resides at the grain boundaries through diffusion. Our study can contribute to substantial reduction of environmentally risk and enhancement of our sustainability by safe control of radioactive materials.

• Journal of the Korean Chemical Society
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• v.31 no.4
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• pp.315-321
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• 1987

The ligand, 2,10-dibenzyl-4,6,8-trioxo-3,9-diaza undecane dioic acid(DTDA) for the extraction of uranium was synthesized under dry nitrogen from phenylalanine and 3-oxoglutaric acid. Extraction was performed by stirring a solution of DTDA in dichloromethane for 1 hour with an aqueous solution of $UO_2(ClO_4)_2{\cdot}6H_2O$ at various pH values and at different $DTDA/UO_2{^{2+}}$ molar ratios. Extraction efficiency reaches a maximum when the pH of the aqueous phase was ca 8.0. The extraction percentage was affected by concentration of DTDA and increases with the $DTDA/UO_2{^{2+}}$ molar ratio to complete extraction with a 4 fold excess of DTDA. The high selectivity of the DTDA for uranium was ascertained by competition experiments with other cations. The bound uranyl ion was quantitatively liberated within few minutes from the organic phase by treatment with an aqueous 1M HCI solution and DTDA was recovered very satisfactorily from the organic phase. The values of the over-all formation constants of the complex between uranyl ion and DTDA were determined to be the following : ${\beta}_1=1.20{\times}10^5\;,\;{\beta}_2=1.01{\times}10^8$.

• Journal of the Korean Ceramic Society
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• v.45 no.11
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• pp.725-733
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• 2008

In this study the preparation method of the spherical ADU droplets, intermediate compound of a HTGR nuclear fuel, was detailed-reviewed and then, the characteristics on an ageing and a washing steps among the wet process and the thermal treatment process on the died-ADU${\rightarrow}UO_3$ conversion with the high temperature furnaces were studied. The key parameters for spherical droplets forming are a precise control of feed rate and a suitable viscosity value selection of a broth solution. Also, a harmony of vibrating frequency and amplitude of a vibration dropping system are important factor. In our case, an uranium concentration is $0.5{\sim}0.7mol/l$, viscosity is $50{\sim}80$ centi-Poise, vibration frequency is about 100Hz. In thermal treatment for no crack spherical $UO_3$ particle, the heating rate in the calcination must be operated below $2^{\circ}C$/min, in air atmosphere.

• Nuclear Engineering and Technology
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• v.46 no.1
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• pp.117-124
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• 2014

The effect of various process variables on the powder properties of recycled $U_3O_8$ from MnO-$Al_2O_3$ doped large grain $UO_2$ pellets and the effect of those recycled $U_3O_8$ powders on the sintered density and grain size of MnO-$Al_2O_3$ doped large grain $UO_2$ pellets have been investigated. The evolution of morphology, size, and BET surface area of the recycled $U_3O_8$ powders according to the respective variation of the thermo-mechanical treatment variables of oxidation temperature, powder milling, and sequential cyclic heat treatment of oxidation and then reduction was examined. The correlation between the BET surface area of recycled $U_3O_8$ powder and the sintered pellet properties of MnO-$Al_2O_3$ doped pellets showed that the pellet density and grain size of doped pellets were increased and then saturated by increasing the BET surface area of the recycled $U_3O_8$ powder. The density and grain size of the pellets were maximized when the BET surface area of the recycled $U_3O_8$ powder was in the vicinity of $3m^2/g$. Among the process variables applied in this study, the cyclic heat treatment followed by low temperature oxidation was a potential process combination to obtain the sinter-active $U_3O_8$ powder.

• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1711-1716
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• 2017

Uranium tetrafluoride ($UF_4$) is the most used nuclear material for producing metallic uranium by reduction with Ca or Mg. Metallic uranium is a raw material for the manufacture of uranium silicide, $U_3Si_2$, which is the most suitable uranium compound for use as nuclear fuel for research reactors. By contrast, ammonium uranyl carbonate is a traditional uranium compound used for manufacturing uranium dioxide $UO_2$ fuel for nuclear power reactors or $U_3O_8-Al$ dispersion fuel for nuclear research reactors. This work describes a procedure for recovering uranium and ammonium fluoride ($NH_4F$) from a liquid residue generated during the production routine of ammonium uranyl carbonate, ending with $UF_4$ as a final product. The residue, consisting of a solution containing high concentrations of ammonium ($NH_4^+$), fluoride ($F^-$), and carbonate ($CO_3^{2-}$), has significant concentrations of uranium as $UO_2^{2+}$. From this residue, the proposed procedure consists of precipitating ammonium peroxide fluorouranate (APOFU) and $NH_4F$, while recovering the major part of uranium. Further, the remaining solution is concentrated by heating, and ammonium bifluoride ($NH_4HF_2$) is precipitated. As a final step, $NH_4HF_2$ is added to $UO_2$, inducing fluoridation and decomposition, resulting in $UF_4$ with adequate properties for metallic uranium manufacture.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.1061-1062
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.240.2-240
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.305.2-305
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• 2002

• 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