• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.109-114
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• 1995

A preliminary study of the heterogeneity effect of spent P% fuel in CANDU was made using a reduced spent PWR fuel data base. The instantaneous core simulation has shown that the refueling ripple in the CANDU reactor is large if the spent PWR fuel is directly used. But the fuel heterogeneity effect can be reduced appreciably by blending spent PWR fuel with a small amount of fresh UO$_2$. The refueling simulation has shown that the operating margins of 6.0% and 8.7% are achievable for the peak channel and bundle powers, respectively, with the blended fuel.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.75-81
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• 1996

COBRA-TF is a two fluid, three field subchannel code. Three fields are continuous vapor, continuous liquid and droplet. Some assessments are conducted to validate the related models and to estimate a code ability through dryout and post-CHF experiment in a tube and DNB test in rod bundles. It turned out form dryout and post-CHF experiment that the predicted dryout locations and wall temperature profiles are in close agreement with the experiments. On the other hand, DNB prediction of COBRA-TF are performed for two kinds of rod bundles along with EPRI CHF correlation. To estimate its performance COBRA-IV of homogeneous model is also run for the same data. The results say that COBRA-TF/EPRI is better in DNB prediction than COBRA-IV/EPRI. In addition the thermal-hydraulic behaviors due to the different two-phase flow models are presented at the condition of CHF.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.295-300
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• 1996

To investigate the effects of the side-orifice on the pressure drop for single-phase flow, a series of experiments have been carried out with 16 different downstream test sections with various combinations of side-orifice shapes, different numbers of side-orifices, and different arrangements of the side-orifice using water as a working fluid. From the measurements of the pressure drop and the flow rate, the pressure loss coefficient of the side-orifice(s) has been evaluated. Based on the total number of 529 present data, an empirical correlation for the pressure loss coefficient has been developed in terms of Reynolds number and geometric parameters, such as area ratio, equivalent diameter, leading edge, and average width of side-orifice.

• Nuclear Engineering and Technology
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• v.28 no.4
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• pp.379-389
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• 1996

To experimentally investigate the several wave patterns for the horizontal countercurrent stratified air-water flow, a series of systematic experimental studies have been performed. The experiments are carried out in a horizontal pipe with 4m in length and 102mm in inner diameter. The oater and air superficial velocities vary from 0.0004 to 0.0204 and from 0 to 6m/s, respectively. The instantaneous water thickness is measured by parallel-wire conductance probes, and the wave field is recorded by high speed video camera. Also, to evaluate the wave effect on interfacial friction factor, the pressure drop is measured. Statistical data anal)sis is accomplished in order to obtain the fundamental wave parameters such as un amplitude, length and velocity, and spatial growth factor. By using these statistical parameters, the wave regime boundaries can be verified.

• Nuclear Engineering and Technology
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• v.29 no.2
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• pp.158-166
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• 1997

Integrated performance test of Chemical and Volume Control System was successfully performed in 1994. However, an extensive effort to correct hardware and software problems in the letdown line was required mainly due to the lack of adequate simulation code to predict the test accurately. Although the LTC computer code was used during the YGN 3'||'&'||'4 NSSS design process, the code can not satisfactorily predict the test due to it insufficient letdown line modeling. This study developed a numerical model to simulate the letdown test by modifying the current LTC code, and then verified the model by comparing with the test data. The comparison shows that the modified LTC computer code can predict the transient behavior of letdown system lese very well. Especially, the model was verified to be able to predict the "Stiction (composition of stick and friction)" phenomena which caused instantaneous fluctuations in the letdown backpressure and flowrate. Therefore, it is concluded that the modified LTC computer code with the ability of calculating the "Stiction" phenomena will be very useful for future plant design and test predictions.predictions.

• Nuclear Engineering and Technology
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• v.31 no.2
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• pp.139-150
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• 1999

The swelling pressure of a potential buffer material was measured and the effect of dry density, bentonite content and initial water content on the swelling pressure was investigated to provide the information for the selection of buffer material in a high-level waste repository. Swelling tests were carried out according to Box-Behnken's experimental design. Measured swelling pressures were in the wide range of 0.7 Kg/$\textrm{cm}^2$ to 190.2 Kg/$\textrm{cm}^2$ under given experimental conditions. Based upon the experimental data, a 3-factor polynomial swelling model was suggested to analyze the effect of dry density, bentonite content and initial water content on the swelling pressure The swelling pressure increased with an increase in the dry density and bentonite content, while it decreased with increasing the initial water content and, beyond about 12 wt.% of the initial water content, levelled off to nearly constant value.

• Nuclear Engineering and Technology
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• v.31 no.3
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• pp.314-327
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• 1999

A subchannel analysis code MATRA applicable to PWRs and ALWRs has been developed to be run on an IBM PC or HP WS based on the existing CDC CYBER mainframe version of COBRA-Rf-1. This MATRA code is a thermal-hydraulic analysis code based on the subchannel approach for calculating the enthalpy and How distribution in fuel assemblies and reactor cores for both steady-state and transient conditions. HATRA has been provided with an improved structure, various functions, and models to give more convenient user environment and to enhance the code accuracy. Among them, the pressure drop model has been improved to be applied to non-square-lattice rod arrays, and the models for the lateral transport between adjacent subchannels have been improved to enhance the accuracy in predicting two-phase flow phenomena. The predictions of MATRA were compared with the experimental data on the flow and enthalpy distribution in some sample rod-bundle cases to evaluate the performance of MATRA. All the results revealed that the predictions of MATRA were better than those of COBRA-IV-I.

• 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,

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.1073-1078
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• 2009

A kinetic model for the oxidation of a $UO_2$ pellet to $U_3O_8$ powder has been suggested by considering the mass transfer and the diffusion of oxygen molecules. The kinetic parameters were estimated by a fitting of the experimental data. The activation energies for the chemical reaction and the product layer diffusion were calculated from the kinetic model. The oxidation conversion of a $UO_2$ pellet was simulated at various operating conditions. The suggested model explains the oxidation behavior of $UO_2$ well.

• Journal of Institute of Control, Robotics and Systems
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• v.22 no.5
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• pp.367-373
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• 2016

In this paper, range measurement performance of LRF (Laser Range Finder) module and image contrast of color CCD camera are evaluated under the aerosol (high temperature steam) environments, which are simulated severe accident conditions of the LWR (Light-Water-Reactor) nuclear power plant. Data of LRF and color CCD camera are key informations, which are needed in the implementation of SLAM (Simultaneous Localization and Mapping) function for emergency response robot system to cope with urgently accidents of the nuclear power plant.