• Proceedings of the Korean Radioactive Waste Society Conference
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• 2007.11a
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• pp.327-328
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• 2007

The deposition conditions for plating polonium have been optimized with deposition parameters such as pH, volume and temperature of the deposition and deposition time. In the tap water, the chemical yields of polonium forthe deposition solution adjusted to pH 0 were higher than those for the deposition solution adjusted to pH 2. This modified auto-deposition method made it possible to obtain reliable data of activity concentration of Po-210.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.655-676
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• 2009

The first-ever integral effect test for simulating a guillotine break of a DVI (Direct Vessel Injection) line of the APR1400 was carried out with the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) from the same prototypic pressure and temperature conditions as those of the APR1400. The major thermal hydraulic behaviors during a DVI line break accident were identified and investigated experimentally. A method for estimating the break flow based on a balance between the change in RCS inventory and the injection flow is proposed to overcome a direct break low measurement deficiency. A post-test calculation was performed with a best-estimate safety analysis code MARS 3.1 to examine its prediction capability and to identify any code deficiencies for the thermal hydraulic phenomena occurring during the DVI line break accidents. On the whole, the prediction of the MARS code shows a good agreement with the measured data. However, the code predicted a higher core level than did the data just before a loop seal clearing occurs, leading to no increase in the peak cladding temperature. The code also produced a more rapid decrease in the downcomer water level than was predicted by the data. These observable disagreements are thought to be caused by uncertainties in predicting countercurrent flow or condensation phenomena in a downcomer region. The present integral effect test data will be used to support the present conservative safety analysis methodology and to develop a new best-estimate safety analysis methodology for DVI line break accidents of the APR1400.

• Journal of Radiation Protection and Research
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• v.28 no.3
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• pp.239-245
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• 2003

Every nuclear power plant measured concentrations of tritium in groundwater and surface water around the plants periodically. It was not easy to predict the tritium concentration only with these measurement data in case of various release scenarios. KAERI developed a new approach to find the relationship between the tritium release rate and tritium concentration in the environment. The approach was based upon a dynamic compartment model. In this paper the dynamic compartment model was modified to predict the tritium behavior more accurately. The mechanisms considered for the transfer of tritium between the compartments were evaporation, groundwater flow, infiltration, runoff, and hydrodynamic dispersion. Time dependent source terms of the compartment model were introduced to refine the release scenarios. Also, transfer coefficients between the compartments were obtained using realistic geographical data. In order to illustrate the model various release scenarios were developed, and the change of tritium concentration in groundwater and surface water around the nuclear power plants was estimated.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.117-122
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• 1995

An Integrated Test Facility(ITF) is a human factors experimental environment to evaluate an advanced man machine interface(MMI) design. The ITF includes a human machine simulator(HMS) comprised of a nuclear power plant function simulator, man-machine interface, experiment control station for the experiment control and design, human behavioural data measurement system, and data analysis and experiment evaluation supporting system(DAEXESS). The most important features of ITF is to secure the flexibility and expandibility of Man Machine Interlace(MMI) design to change easily the environment of experiments to accomplish the experiment's objects In this paper, we describe a development scope and characteristics of the ITF such as, hardware and software development scope and characteristics, system thermohydraulic modelling characteristics, and experiment station characteristics for the experiment variables design and control, to be used as an experiment environment for the evaluation of VDU-based control room.

• Proceedings of the KIEE Conference
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• 2006.04a
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• pp.138-140
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• 2006

A new diagnosis model was constructed by combining atomic force microscopy (AFM), wavelet, and neural network. Plasma faults were characterized by filtering AFM-measured etch surface roughness with wavelet. The presented technique was evaluated with the data collected during the etching of silicon oxynitride thin film. A total of 17 etch experiments were conducted. Applying wavelet to AFM, surface roughness was detailed into vertical, horizon%at, and diagonal components. For each component, neural network recognition models were constructed and evaluated. Comparisons revealed that the vertical component-based model yielded about 30% improvement in the recognition accuracy over others. The presented technique was evaluated with the data collected during the etching of silicon oxynitride thin film. A total of 17 etch experiments were conducted

• Nuclear Engineering and Technology
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• v.12 no.3
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• pp.180-193
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• 1980

Subroutines to enable fast and accurate generation of water properties-enthalpy, specific volume, viscosity, thermal conductivity and saturation entropy-which are usually basic requirements for nuclear calculation of LWR, have been developed. The sources of data were quoted from “ASME Steam Tables (1967)” and their Revision (1975). It is ensured that the obtained values from this routine fall within 0.2% difference compared with the reference data, in the ranges of temperature and pressure for LWR nuclear calculation.

• Nuclear Engineering and Technology
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• v.24 no.1
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• pp.1-13
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• 1992

A one-group cross section data base of the ORIGEN2 computer code was developed for research reactor applications. For this, ENDF/B-IV and -V data were processed using the NJOY code system into 69-group data. The burnup-dependent weighting spectra for KMRR were calculated with the WIMS-KAERI computer code, and then the 69-group data were collapsed to one-group using the spectra. The ORIGEN2-predicted burnup-dependent acti-nide compositions of KMRR spent fuel using the newly developed data base show a good agreement with the results of detailed multigroup transport calculation. In addition, the burnup characteristics of KMRR spent fuel was analyzed with the new data base.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1796-1802
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• 2024

The production cross sections of ^55,56,57Co, ^52gFe, ^52g,54Mn, ⁵¹Cr, and ⁴⁸V from the ^natFe (p,x) reactions were measured using a proton energy of 57 MeV at the Korea Multi-purpose Accelerator Complex (KOMAC) in Gyeongju, Korea. The conventional stacked-foil activation method and offline γ-ray spectroscopy were used to determine the excitation functions of proton induced nuclear reactions on iron. The measured excitation functions were compared with experimental data in literature and theoretical data from the TENDL-2021 library. The present data show generally good agreement with other experimental data, but discrepancies were found between the present data and the excitation functions of the TENDL-2021 library in the investigated energy range, except for ^56,57Co and ⁵⁴Mn.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.363-364
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.61.1-61.1
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• 2003