• Nuclear Engineering and Technology
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• v.50 no.2
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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.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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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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• v.52 no.12
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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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• v.55 no.10
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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.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.267-267
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• 1999

• Journal of the Korean Society for Precision Engineering
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• v.30 no.2
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• pp.223-230
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• 2013

When a new nuclear fuel is developed, an irradiation test needs to be carried out in the research reactor to analyze the performance of the new nuclear fuel. In order to check the performance of a nuclear fuel during the irradiation test in the test loop of a research reactor, sensors need to be attached in and out of the fuel rod and connect them with instrumentation cables to the measuring device located outside of the reactor pool. In particular, to check the temporary temperature change at the center of a nuclear fuel during the irradiation test, a thermocouple should be instrumented at the center of the fuel rod. Therefore, a hole needs to be made at the center of fuel pellet to put in the thermocouple. However, because the hardness and the density of a sintered $UO_2$ pellet are very high, it is difficult to make a small fine hole on a sintered $UO_2$ pellet using a simple drilling machine even though we use a diamond drill bit made by electro deposition. In this study, an automated drilling machine using a CVD diamond drill has been developed to make a fine hole in a fuel pellet without changing tools or breakage of workpiece. A sintered alumina ($Al_2O_3$) block which has a higher hardness than a sintered $UO_2$ pellet is used as a test specimen. Then, it is verified that a precise hole can be drilled off without breakage of the drill bit in a short time.

• Journal of the Korean Ceramic Society
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• v.40 no.5
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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.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.268-268
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• 1999

• The Korean Journal of Ceramics
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• v.6 no.3
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• pp.209-213
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• 2000

Microstructural changes of AlOOH doped UO$_2$pellet after annealing up to 216h have been observed and they were compared with those of the standard pellet. Grain and pore size of UO$_2$pellet increased with the addition of AlOOH and its effect was still validated during annealing. Densification rate was reduced by the addition of AlOOH and it was attributed to coarsened pores with spherical shape. Grain and pore growth was stopped and density increase was the least after 144h of annealing. The variation of pore size resulting from annealing has a linear relationship with that of grain size.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.265-270
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• 1996

The temperature distribution in the pellet was obtained from beginning the general heat conduction equation. The thermal conductivity of pellet used the SIMFUEL data that made clear the effect of burnup on the thermal conductivity degradation. Since the pellet rim acts as the thermal barrier to heat flow. the pellet was subdivided into several rings in which the outer ring was adjusted to play almost the same role as the rim. The local burup in each ring except the outer ring was calculated from the power depression factor based on FASER results. whereas the rim burnup at the outer ring was achieved by the pellet averaged burnup based on the empirical relation. The rim changed to the equivalent Xe film so the predicted temperature shooed the thermal jump across the rim. The observed temperature profiles depended on linear heat generation rate. fuel burnup. and power depression factor. The thermal conductivity degradation modelling can be applied to the fuel performance code to high burnup fuel,