• 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.29 no.2
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• pp.93-98
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• 1997

Air-oxidation experiments on unirradiated and irradiated UO$_2$ were performed at temperature from 150 to 375$^{\circ}C$ for investigating the long-term storage behavior of spent PWR fuel. The rate of oxidation was monitored by a thermogravimetric analyzer(TGA) and an X-ray diffraction(XRD). The correlation between the onset-time for U$_3$O$_{8}$ formation and temperature was given as follows, logt(hr) = -12.89+7650/T(K), 423$_2$ pellet, the oxidation rate of irradiated UO$_2$ increase more rapidly at the initial stage and shows a lower saturation point at the later Stage. The Oxidation rate of high bumup UO$_2$ and gadolinia-doped UO$_2$(Gd$_2$O$_3$-UO$_2$) were observed to be much slower than that of unirradiated UO$_2$ pellets.s.

• Journal of the Korean Ceramic Society
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• v.29 no.8
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• pp.609-616
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• 1992

This study has been investigated on the milling of Aluminium Distearate (ADS) powder and characteristics of the ADS-doped UO2 pellets. As-received ADS powder of the agglomerated particles has not shown any milling effect because of heat generated during planetary milling. But the use of coolant to effectively remove heat generated during milling has been found an effective way in breaking up the agglomerates of ADS powder. The green density of the UO2 pellet decreases with the amount of ADS powder doped. Therefore, in order to get the sintered density of 95% pellet decreases with the amount of ADS powder doped. Therefore, in order to get the sintered density of 95% theoretical density, the 200 ppm ADS-doped UO2 pellet has to be pressed under higher compacting pressure of 3500~4000 kgf/$\textrm{cm}^2$ compared with the ADS-undoped UO2 pellet pressed under around 3000 kgf/$\textrm{cm}^2$. The ADS-dpoed UO2 pellet with even relatively low sintered density of 10.27 g/㎤ exhibits open porosity of 1% while open porosity of the ADS-undoped UO2 pellet is reduced to around 1% only after its sintered density increases to 10.43g/㎤. It is, therefore, concluded that doping of ADS powder significantly contributes to the decrease in open porosity of the UO2 pellet. The dilatometry of the ADS doped UO2 pellet shows the sintering rate curve with the bimodal mode, which could be attributed to a phase reaction between UO2 and ADS. The X-ray diffraction analysis indicates that there occurs not any new phase formed but the shift of the peaks. It would be expected that a phase reaction resulting in solid solution would happen in the temperature range of 130$0^{\circ}C$ to 150$0^{\circ}C$ between UO2 and ADS.

• Journal of Energy Engineering
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• v.15 no.1 s.45
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• pp.8-13
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• 2006

The oxidation behavior of $UO_2$, pellet was studied using an XRD and a thermogravimetric analyzer in the temperature range from 573 to 873 K and in the density range from 94.64 to 99.10% of theoretical density in air. It was found from the XRD study that $UO_2$ was completely converted to $U_3O_8$ in this experimental temperature range. The formation of $U_3O_8$ displays sigmoidal reaction kinetics. The oxidation rate was reduced with density. Induction time for the oxidation of $UO_2$ was delayed with density because of open pore formed in surface of $UO_2$ pellet. The activation energy for oxidation of $UO_2$ was determined to be 89.54 kJ/mol and 34.40 kJ/mol in the temperature range from 573 to 723 K and from 723 to 873 K, respectively.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.90-99
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• 1994

One of the major phenomena occurring in defective fuel rods is the oxidation of UO$_2$ fuel pellets from UO$_2$ to UO$_{2+}$x/ by the oxygen Produced from the dissociation of the steam in the Pellet-to-clad gap, which leads to the enhancement of fission gas release. In this paper, the oxidation kinetics of defective fuel rods was analyzed on the basis of operating conditions of the reactor and defective fuel rod itself. Oxidation kinetics of the fuel pellet was determined under the assumption that the gap is filled with the saturated steam of 150 atm and an enhancement factor for fission gas release was introduced to take into account the effect of fuel oxidation on fission gas release. Comparison with experimental data shows that the enhancement factor predicts well the increased fission gas release due to the oxidation of UO$_2$fuel pellets.

• Journal of the Korean Ceramic Society
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• v.39 no.2
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• pp.204-209
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• 2002

In this study, $UO_4{\cdot}2NH_4F$, the precipitates which has low solubility, was obtained by chemical precipitation method to recover and reuse the trace uranium from the liquid waste producing in AUC process and for this compound it was characterized by means of chemical analysis, TG-DTA, XRD and FT-IR analyses. This compound was analyzed as $UO_4{\cdot}2NH_4F$ and shape of this precipitate was hexagonal type, having the size of 2∼3 ${\mu}m$. Also, the intermediates were obtained as $UO_4F,\;UO_4,\;UO_3,\;and\;U_3O_8$ by the thermal decomposition over the temperature of 220, 310, 515 and 640$^{\circ}C$, respectively. It is concluded that under the condition of a constant heating rate of 5$^{\circ}C$/min in air atmosphere range of between room temperature and 800$^{\circ}C$, thermal decomposition reaction mechanism of $UO_4{\cdot}2NH_4F$ is as follow; $UO_4{\cdot}2NH_4F{\rightarrow}UO_4F{\rightarrow}UO_4{\rightarrow}UO_3{\rightarrow}U_3O_8$.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.391-392
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• 2005

• 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 Ceramic Society
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• v.35 no.6
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• pp.559-568
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• 1998

Studies of preparation condition and characteristics of AUC(ammonium uranyl carbonate) were carried out to optimize AUC process with different reactor sizes and precipitation methos. As results four types of precipitates with different chemical compositions and morphologies were obtained from the reaction of {{{{ {(NH }_{4 }) { }_{2 } {CO }_{3 } }} with {{{{ {UO }_{2 }( {NO }_{3 }) { }_{2 } }} solution. A phase diagram has been made and crystal structure and chemical composition of each phase have been characterized by using SEM X-ray IR and thermal analysis. It was found that ammonium uranyl carbonate {{{{ {(NH }_{4 }) { }_{4 } {UO }_{2 } {(CO }_{3 }) { }_{3 } }} with monoclinic crystal morphology could be syn-thesized when the mole ratio of in {{{{ {(NH }_{4 }) { }_{2 } {CO }_{3 }/ {UO }_{2 } {(NO }_{3 }) { }_{2 } }} in the solution was higher than 5 Also a mechanism and a precipitating condition on rounding of the AUC particle were examined in the course of the AUC pre-cipitation. The rounding of the AUC particle was possible only by external circulation using pump not by internal circulation using agitator.

• Journal of the Korean Electrochemical Society
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• v.3 no.3
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• pp.141-145
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• 2000

Anodically Polymerized conducting Polypyrrole film electrode was employed to Pick up uranyl ion with the type of Gr/ppy, xylenol orange modified electrode. To have Porous and oriented ppy film, NBR was applied as precoating agent. The rate constant of polymerization was $3.22\times10^{-3}s^{-1}$ which was 1.6 times smaller value than bare graphite surface. The deposited amount of uranyl iou on $1.70Ccm^{-2}$ of ppy was $1.55\times10^{-4}g$. The matrix effect in artificial seawater was $6.8\%$. The polymer film electrode has a diffusion controlled process in conduction, but the modified Gr/ppy, $X.O^{4-}UO^+$ type was influenced on the ion doping and electronic conduction of film itself owing to increasing of impedance. The capacitance of electrical double layer was respectively enhanced to 56 and 130 times in Gr/ppy, $X.O.^{4-}$ and Gr/ppy, $X.O^{4-}UO^+$ than Grippy type electrode.