• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.45-50
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• 1997

The conventional MOX fuel shows adverse controllability in view of its neutronic characteristics such as decreased soluble boron worth and effective delayed-neutron fraction compared to the UO$_2$ fuel. In order to mitigate these disadvantages, we devised a new concept of the hybrid UO$_2$-MOX fuel pellet with dual structure such that its outer annular section contains. UO$_2$ fuel and its inner cylindrical bar contains MOX fuel. The lattice physics code HELIOS was used to evaluate the neutronic characteristics of three different types of fuel pellets ; UO$_2$ fuel pellet, MOX fuel pellet, and hybrid UO$_2$-MOX fuel pellet. Results show that the hybrid UO$_2$-MOX fuel pellet generally has intermediate neutronic tendency between UO$_2$ fuel and MOX which could diminish the problems arising from the use of the conventional MOX fuel.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2880-2886
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• 2020

A small leak occurring on the surface of a fuel rod due to damage exposes UO₂ to a steam atmosphere. During this time, fission gas trapped inside the fuel rod leaks out, and the gas leakage can be increased due to UO₂ oxidation. Numerous studies have focused on the steam oxidation and its thermodynamic calculation in UO₂. However, the thermodynamic calculation of the UO₂ oxidation in a pressurized water reactor (PWR) environment has not been studied extensively. Moreover, the kinetics of the oxidation of UO₂ pellet also has not been investigated. Therefore, in this study, the thermodynamics of UO₂ oxidation under steam injection due to a damaged fuel rod in a PWR environment is studied. In addition, the diminishing radius of the UO₂ pellet with time in the PWR environment was calculated through an experiment simulating the initial time of steam injection at the puncture.

• Nuclear Engineering and Technology
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• 제50권2호
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• pp.253-258
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• 2018

A microcell $UO_2$ pellet, as an accident-tolerant fuel pellet, is being developed to enhance the accident tolerance of nuclear fuels under accident conditions as well as the fuel performance under normal operation conditions. Improved capture-ability for highly radioactive and corrosive fission product (Cs and I) is the distinct feature of a ceramic microcell $UO_2$ pellet, and the enhanced pellet thermal conductivity is that of a metallic microcell $UO_2$ pellet. The fuel temperature can be effectively decreased by enhanced thermal conductivity. In this study, the material concepts of metallic and ceramic microcell $UO_2$ pellets were designed, and the fabrication process of microcell $UO_2$ pellets embodying the designed concept was developed. We successfully implemented the microcell $UO_2$ pellets and produced microcell $UO_2$ pellets. In addition, an assessment of the out-of-pile properties of a microcell $UO_2$ pellet was performed, and the in-reactor performance and behavior of the developed microcell pellets were evaluated through a Halden irradiation test. According to the expectations, the excellent performance of the microcell $UO_2$ pellets was confirmed by the online measurement data of the Halden irradiation test.

• Nuclear Engineering and Technology
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• 제30권2호
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• pp.128-139
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• 1998

The effect of UO$_2$ powder property and oxygen potential on characteristics of sintered UO$_2$-Gd$_2$O$_3$ fuel pellets has been investigated. Two types of powder, mixture of AUC-UO$_2$ and Gd$_2$O$_3$powders (type I) and mixture of ADU-UO$_2$ and Gd$_2$O$_3$powders (type II), have been prepared, pressed, and sintered at 168$0^{\circ}C$ for 4 hours. Four sintering atmospheres with different mixing ratios of $CO_2$to H$_2$ gas ranging from 0 to 0.3 have been used. UO$_2$-Gd$_2$O$_3$ fuel has lower sintered density than UO$_2$ fuel, and the density drop is larger for powder type I than for powder type II. As the oxygen potential increases, the sintered density of UO$_2$-2wt% Gd$_2$O$_3$pellets increases but that of UO$_2$-10wt% Gd$_2$O$_3$ pellets decreases. It is found that pores are newly formed in UO$_2$-10wt% Gd$_2$O$_3$ pellets in accordance with the decrease in density. The grain size of UO$_2$-Gd$_2$O$_3$ fuel increases and a short range G4 distribution becomes homogeneous as the oxygen potential increases. A long range ed distribution and grain structure are inhomogeneous for powder type II. The lattice parameter of (U,Gd)O$_2$solid solution decreases linearly with Gd$_2$O$_3$ content. The dependence of UO$_2$-Gd$_2$O$_3$fuel characteristics on powder type and sintering atmosphere have been discussed.

• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1099-1109
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• 2020

The performance of gadolinium burnable absorber (GdBA) for reactivity control in UO₂ and (U, Th)O₂ fuels and its impact on spent fuel characteristics was performed. Five fuel assemblies: one without GdBA fuel rod and four each containing 16, 24, 34 and 44 GdBA fuel rods in both fuels were investigated. Reactivity swing in all the FAs with GdBA rods in UO₂ fuel was higher than their counterparts with similar GdBA fuel rods in (U, Th)O₂ fuel. The excess reactivity in all FAs with (U, Th)O₂ fuel was higher than UO₂ fuel. At the end of single discharge burn-up (~ 49.64 GWd/tHM), the excess reactivity of (U, Th) O₂ fuel remained positive (16,000 pcm) while UO₂ fuel shows a negative value (-6,000 pcm), which suggest a longer discharge burn-up in (U, Th)O₂ fuel. The concentration of plutonium isotopes and minor actinides were significantly higher in UO₂ fuel than in (U, Th)O₂ fuel except for ²³⁶Np. However, the concentration of non-actinides (gadolinium and iodine isotopes) except for ¹³⁵Xe were respectively smaller in (U, Th)O₂ fuel than in UO₂ fuel but may be two times higher in (U, Th)O₂ fuel due to its potential longer discharge burn-up.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(2)
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• pp.171-176
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• 1997

UO$_2$-Gd$_2$O$_3$fuel has been sintered to study the effect of powder property and sintering atmospheres on densification and microstructure. Three types of powders have been used; AUC-UO$_2$ powder and ADU-UO$_2$ powder were mixed with Gd$_2$O$_3$ Powder, and co-milled AUC-UO$_2$ and Gd$_2$O$_3$ powder. UO$_2$-(2, 5, 10)wt% Gd$_2$O$_3$pellets have been sintered at 168$0^{\circ}C$ for 4 hours in the mixture of H$_2$ and $CO_2$ gases, of which oxygen potential has been controlled by the ratio of $CO_2$ to H$_2$ gas. Densities of UO$_2$-Gd$_2$O$_3$ fuel pellets are quite dependent on powder types, and UO$_2$-Gd$_2$O$_3$ fuel using co-milled UO$_2$ powder yields the highest density. A long range homogeneity of Gd is determined by powder mixing. As the oxygen potential of sintered atmosphere increases, the sintered densities of UO$_2$-Gd$_2$O$_3$ pellets decrease but grain size increases. In addition, (U, Gd)O$_2$ solid solution becomes more homogeneous. The UO$_2$-Gd$_2$O$_3$fuel having adequate density and homogeneous microstructure can be fabricated by co-milling powder and by high oxygen potential.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2206-2214
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• 2023

The existing wet reconversion processes for the recovery of scraps generated in manufacturing of nuclear fuel are complex and require several unit operation steps. In this study, it is attempted to simplify the recovery process of high-quality fuel-grade UO₂ powder. A novel wet reconversion process for uranyl nitrate hexahydrate solution is suggested by using a newly developed pulsed fluidized bed reactor, and the resultant chemical characteristics are evaluated for the intermediate ammonium uranate hydrate product and subsequently converted UO₂ powder, as well as the compliance with nuclear fuel specifications and advantages over existing wet processes. The UO₂ powder obtained by the suggested process improved fuel pellet properties compared to those derived from the existing wet conversion processes. Powder performance tests revealed that the produced UO₂ powder satisfies all specifications required for fuel pellets, including the sintered density, increase in re-sintered density, and grain size. Therefore, the processes described herein can aid realizing a simplified manufacturing process for nuclear-grade UO₂ powders that can be used for nuclear power generation.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3860-3865
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• 2023

A metallic microcell UO₂ pellet has a microstructure where a metal wall is connected to overcome the low thermal conductivity of the UO₂ fuel pellet. It has been verified that metallic microcell fuel pellets provide an impressive reduction of the fuel centerline temperature through a Halden irradiation test. However, it is difficult to predict the effective thermal conductivity of these pellets and researchers have had to rely on measurement and use of the finite element method. In this study, we designed a unit microcell model using a thermal resistance circuit to calculate the effective thermal conductivity on the basis of the microstructure characteristics by using the aspect ratio and compared the results with those of reported metallic UO₂ microcell pellets. In particular, using the thermal conductivity calculated by our model, the fuel centerline temperature of Cr microcell pellets on the 5th day of the Halden irradiation test was predicted within 6% error from the measured value.

• 한국세라믹학회지
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• 제40권5호
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• pp.410-414
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• 2003

Characteristics of $UO_{2+x}$ powders oxidized at different temperatures were examined. Pellets were fabricated by adding these oxidation powders and their properties were also investigated. Particle size of the $UO_{2+x}$ powders decreased with increasing oxidation temperature while surface area increased. Only the powders oxidized at 35$0^{\circ}C$ enhanced the strength of green pellet. However, 35$0^{\circ}C$ oxidized powders added pellet had many surface defects. The difference of shrinkage rate between the oxidized and UO$_2$ powders was thought to be the cause of them.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.739-744
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• 1996

The U solubility in the Zircaloy melt including the other investigators' result was investigated in a range of reaction temperatures from 2223k to 2473k and for the mole ratios of UO$_2$ crucible/Zircaloy-4 melt(subsequently abbreviated as UO$_2$/Zry) from 2.4 to 18.2, The U solubility in the melt increased with increasing reaction temperature and with decreasing the mole ratio of UO$_2$/Zry. An empirical correlation was obtained as functions of UO$_2$/Zry mole ratio and reaction temperature including other investigators' results. The experimental results with use of internally heated fuel element simulators were analyzed by the empirical correlation from UO$_2$ crucible experiments.