• Journal of KSNVE
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• v.10 no.1
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• pp.129-137
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• 2000

This paper presents the vibration characteristics of reactor internals of Ulchin-1 nuclear power plant, which are identified by using the conventional and the phase separated spectral analysis of the pressure vessel acceleration and ex-core neutron signals. These identified vibration characteristics show excellent agreement with those of Tricastin-1 nuclear power plant that is the prototype of Ulchin-1. And the trend of ex-core neutron signals has been observed during one reactor cycle. These results can be used as basic data for fault diagnosis of reactor internals.

• Journal of the Korea Safety Management & Science
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• v.7 no.4
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• pp.85-99
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• 2005

The main objective of this paper is to search whether containment vessel's best pressure may increase until how long when loss of coolant accident (LOCA) happened in containment vessel of Ulchin nuclear power plant 1 and 2. Another goal of this research is to find the influential factors that increase containment vessel pressure. Model for this research is Ulchin nuclear power plant 1 with 10 cycles. Data were collected by simulator of Ulchin nuclear power plant 1 and design of experiment was used for data analysis. For the experiment, seven factors that are going to influence in containment vessel pressure were chosen. It was found that fatter which influences in early rise of containment vessel pressure after LOCA is only explosion size. Also, containment vessel's best pressure (3.74 bar.a) was much lower than limit (4.86 bar.a) of FSAR (Final Safety Analysis Report).

• Journal of Korean Society of Coastal and Ocean Engineers
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• v.14 no.1
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• pp.1-7
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• 2002

In the past, safety assessment on the site of Ulchin Nuclear Power Plants against tsunamis was carried out with probable maximum earthquake magnitude and related tsunamigenic fault parameters. Recently, however, based on the seismic gap theory, some seismologists warned about earthquakes of larger magnitudes than had been expected. In this study, we revaluated tsunami risk with a finite difference model based on linear and nonlinear shallow water equations. Firstly, we simulated the\`83 tsunami and compared the calculated water surface profile with the observed wave heights. Secondly, we evaluated the rise and drop of sea water level at the site of Ulchin Nuclear Power Plant with fault parameters of the past '83, '93 tsunamis and some dangerous faults. Finally, we showed that the cooling water intake facility of Ulchin Nuclear Power Plants would be safely operated in disastrous tsunamis.

• Korean Journal of Environmental Biology
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• v.22 no.3
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• pp.426-437
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• 2004

To understand the phytoplankton community in adjacent waters of Ulchin nuclear power plant (UNPP), abundance and the size fractionated $chl-\alpha$ concentrations were evaluated through seasonal interval sampling from April 2003 to February 2004. A total of 211 different phytoplankton species was observed and mean abundance of phytoplankton in each study period ranged from 244,286 to 1,221,779 cells $L^{-1}$. The contributions of microplankton $(>20\mu{m})$ to total phytoplankton abundance ranged from 42.5 to 83.6% (average 66.1%) and those of nanoplankton $(>20\mu{m})$ ranged from 16.4 to 57.5% (average 33.9%). Total chl-$\alpha$ concentrations of phytoplankton ranged from 0.52 to $2.26\mu{g}\;L^{-1}$. The contribution of chl-$\alpha$ concentrations of microplankton was higher than that of nano- and picoplankton through the study period with exception of July 2008. The results of abundances and $chl-\alpha$ concentrations suggest that microplankton has an important role in adjacent waters of UNPP. The diminution of abundances and $chl-\alpha$ concentrations of phytoplankton was observed after passage through the cooling water system, but it was gradually recovered by mixing with the ambient waters. Our results suggested that the influence of thermal discharges on phytoplankton should be restricted within narrow limits around outlet area of thermal effluents.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.417-422
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• 1998

he Reactor Coolant Gas Vent System(RCGVS) design for Ulchin Nuclear Power Plant Units 3&4(UCN 3&4) has been improved from the Yonggwang Nuclear Power Plant Units 3&4(YGN 3&4) based on the evaluation results for depressurization capability tests performed at YGN 3&4. There has been a series of plant safety analyses for Natural Circulation Cooldown(NCC) event and thermo-dynamic analyses with RELAP5 code for the steam blowdown Phenomena in order to optimize the orifice size of UCN 3&4 RCGVS. Baesd on these analyses results, the RCGVS orifice size for UCN 3&4 has been reduced to 9/32 inch from the l1/32 inch for YGN 3&4. The depressurization capability tests, which were performed at UCN 3 in order to verify the FSAR NCC analysis results, show that the RCGVS depressurization rates are being within the acceptable ranges. Therefore, it is concluded that the orificed flow path of UCN 3&4 RCGVS is adequately designed, and can provide the safety-grade depressurization capability required for a safe plant operation.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.76.1-76
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.488.1-488
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• 2001

• Nuclear Engineering and Technology
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• v.28 no.3
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• pp.290-298
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• 1996

Direct Containment Heating, DCH, is supposed to occur during a core melt-down accident if the primary system pressure is still high at the time of vessel breach in a Nuclear Power Plant (NPP). In this case, DCH is considered to be one of very important severe phenomena during postulated severe accident scenario because of the fast heat transfer rate to atmosphere and the sharp pressure increase in a containment. To reduce the effect of this DCH phenomena, the capture volume wes designed at Ulchin NPP units 3 and 4. But, the effect of this has not been studied extensively. This work consists of experimental and numerical analyses of the effects of capture volume in the cavity on DCH phenomena. The experimental model is a 1/30 scaled-down model of Ulchin NPP units 3 and 4. We used three types of capture volumes to investigate the effect of size. Numerical analysis using CONTAIN 1.2 is performed with the correlation for the dispersed fraction of molten corium from the cavity into the containment derived from the experimental data to examine the effect of capture volume on DCH phenomena in full scale of Ulchin NPP units 3 and 4.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.2
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• pp.14-20
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• 2012

Full load rejection capability of nuclear power plant depends primarily on steam dump capacity (SDCAP) and steam generator level control capability. Recently, Ulchin Units 1&2 have performed stretched power uprate (SPU) and replacement steam generator (RSG) projects, which increase the power by 4.5 percent. They change major design or operating parameters and especially reduces steam dump capacity at full power due to increase of the steam flow. The reduction of SDC after SPU results in degradation of heat removal capability in full load rejection transients. Therefore, we should perform evaluation to determine whether reactor trips occur in large load rejection transients. Uchin Units 1&2 have experienced full load rejection (FLR) three times from 2004 to 2010. Operating data from the plant occurrence of FLR at Ulchin Units 1&2 showed that steam generator (SG) level transients were limiting in point of reactor trip. However the plant had never reached reactor trip in the FLR and successfully continued in house load operation. The parameters and setpoints for the SG will be changed if the SG is replaced. Therefore, we evaluated the appropriateness of steam dump, main feedwater and steam generator water level control system preventing the plant from reactor trip in case of FLR by the parameter sensitivity study whether SG water level operated smoothly after SPU and RSG projects.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.429-434
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• 1996

K-REM［1］, which is under development as a realistic evaluation model of large break LOCA, is applied to the analysis of cold leg guillotine break of Ulchin 3&4. Fuel parameters on which statistical analysis of their effects on the peak cladding temperature (PCT) are made and system parameters on which the concept of limiting value approach (LVA) are applied, are determined from the single parameter sensitivity study. 3 parameters of fuel gap conductance, fuel thermal conductivity and power peaking factor are selected as fuel related ones and 4 parameters of axial power shape, reactor power, decay heat and the gas pressure of safety injection tank (SIT) are selected as plant system related ones. Response surface of PCT is generated from the plant calculation results and on which Monte Carlo sampling is made to get plant application uncertainty which is statistically combined with code uncertainty to produce the 95th percentile PCT. From the break spectrum analysis, blowdown PCT of 1350.23 K and reflood PCT of 1195.56 K are obtained for break discharge coefficients of 0.8 and 0.5, respectively.