• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.405-406
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.171-172
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• 2005

• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.863-868
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• 2014

The evaluation of thermal properties of actinide oxide fuels is a problem of high importance for the development of new generation reactors. In the present study, an expression obtained for n-dimensional Debye functions is used to derive a simple analytical expression for the specific heat capacity of nuclear fuels. To test the validity and reliability of this expression, the analytical expression is applied to $UO_2$, $NpO_2$, $ThO_2$, and $PuO_2$. It is seen that the formula was in agreement with the experimental and theoretical results reported in the literature.

• Journal of the Korean Crystal Growth and Crystal Technology
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• v.17 no.2
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• pp.75-81
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• 2007

Densification and grain growth have been investigated in 5 wt% $U_3O_8$ seeded $UO_2$ and compared with those of the common $UO_2$ pellet. $UO_2$ compacts and $U_3O_8$ seeded $UO_2$ compacts were sintered at $1300{\sim}1700^{\circ}C$ for 0 h to 4 h. Density and grain size of the sintered pellets were measured by the water immersion method and the image analyzer, respectively. The seeded pellet has a slightly lower density during the intermediate sintering stage. However, the difference of density between two pellets decreases to about 0.5%TD with increasing the sintering temperature. The grain size of the two pellets is similar until $1600^{\circ}C$ but that of the seeded pellet rapidly increases with increasing the sintering temperature.

• Journal of the Korean Ceramic Society
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• v.41 no.8
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• pp.588-592
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• 2004

Effect of the UO$_2$ powder characteristic changes by dynamic milling method on the density was investigated. particle size decreased and its shape was changed from irregular to round form with increasing milling time (0∼8 h), while its specific surface area and O/U ratio increased. It was shown that the sintered density decreased, while green density increased with these powder characteristic changes. It could be considered that this decrement was affected by increased O/U ratio of the UO$_2$ powder during dynamic milling.

• Nuclear Engineering and Technology
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• v.9 no.3
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• pp.125-137
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• 1977

The temperature coefficient has been investigated on the Wolsung nuclear power reactor, in which fuel is natural uranium dioxide and moderator heavy water. The numerical computations are carried out in terms of changes of the effective neutron multiplication factor with respect to fuel, moderator, and coolant temperatures. Those results are compared with the computed values of temperature coefficient based on the LATREP computer code.

• Nuclear Engineering and Technology
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• v.16 no.1
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• pp.18-20
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• 1984

The detection limit of boron has been lowered further in the capture gamma-ray measurement after preconcentration of boron by placing natural lithium brick in front of Ge(Li) detector. The experimental detection limit is found to be 0.30ppm, 0.18ppm, 0.045ppm and 0.090ppm for the samples of aluminum, steel, uranium dioxide and graphite, respectively. An alternate counting technique kas been also used for neglecting the error caused by the fluctuation of neutron flux during counting.

• Proceedings of the Korean Powder Metallurgy Institute Conference
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• 2006.09a
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• pp.187-188
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• 2006

In contrast with the Finite Element Method, the Discrete Element Method (DEM) takes explicitly into account the particulate nature of powders. DEM exhibits some drawbacks and many advantages. Simulations can be computationally expensive and they are only able to represent a volume element. However, these simulations have the great advantage of providing a wealth of information at the microstructural level. Here we demonstrate that the method is well suited for modelling, in coordination with FEM, the compaction of ceramic $UO_2$ particles that have been aggregated. Aggregates of individual ceramic crystallites that are strongly bonded together are represented by porous spheres.

• Nuclear Engineering and Technology
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• v.37 no.4
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• pp.309-316
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• 2005

The contents of a new book currently in preparation are described. The dearth of books in the field of nuclear materials has left both students in nuclear materials classes and professionals in the same field without a resource for the broad fundamentals of this important sub-discipline of nuclear engineering. The new book is devoted entirely to materials problems in the core of light-water reactors, from the pressure vessel into the fuel. Key topics deal with the $UO_2$ fuel, Zircaloy cladding, stainless steel, and of course, water. The restriction to LWR materials does not mean a short monograph; the enormous quantity of experimental and theoretical work over the past 50 years on these materials presents a challenge of culling the most important features and explaining them in the simplest quantitative fashion. Moreover, LWRs will probably be the sole instrument of the return of nuclear energy in electric power production for the next decade or so. By that time, a new book will be needed.

• 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.