• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.1087-1088
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.420-421
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• 2005

• Journal of Energy Engineering
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• v.21 no.1
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• pp.75-80
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• 2012

In this paper a realistic bounding method for flooding analysis following rupture of large size of thanks and piping is defined. Mass and energy release during main feedwater line break accident is analyzed with RETRAN code. It is modeled that the main feed water control valve is closed in 5.0 seconds after reactor trip. In result of the analysis, largest mass and energy is discharged at 70% reactor power. The flood sources for main feedwater room are calculated when piping failure occurs in the high energy line and medium energy line. Based on the result of flood level (1.43m), it is investigated that all of the safety-related environmental qualification equipments are well located above the flood level.

• Proceedings of the KIEE Conference
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• 2006.07d
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• pp.1885-1886
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• 2006

The thermal energy from nuclear fission is transferred to the steam generator which is a kind of a large heat exchanger. After the feedwater is injected into the steam generator and absorbs the thermal energy, it is converted into the steam. This steam goes into the turbine. The balance between the generated energy and the consumed energy is required for the nuclear power plant to be stable. For the purpose of which, the feed water, a parameter for energy transfer, should be controlled in stability. Usually, the nuclear power plants are operated in base load in the view of power system for the stability of fission system. Therefore, though there will be almost no unbalance, there can be some instability from unbalance in case of startup/shutdown or disturbance. In this case, the controllability of feedwater pump is very important for the quick recover of stability.

• Transactions of the Korean Society of Mechanical Engineers
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• v.16 no.4
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• pp.753-764
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• 1992

The steam generator water level is difficult to control at low power due to its reversed responses to the feedwater flow, which are well known as the shrink and swell phenomena. With regard to this problem a new control scheme has been studied by which the level transients could be kept within permissible ranges at low power. The relations between the various input conditions to steam generator and the level transients have been examined to be expressed in the form of process transfer functions. Analog filters have been incorporated to be expressed in the process with proper control constants. This control scheme allows the prediction of level variation together with the corresponding feedwater rate and results in mider transients with good stabilites.

• Nuclear Engineering and Technology
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• v.41 no.3
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• pp.287-294
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• 2009

Due to thermal hydraulics phenomena, such as thermal stratification, various events occur to the parts of a nuclear power plant during their lifetimes: e.g. cracked and dislocated pipes and thermally fatigued, bent, and damaged supports. Due to the operational characteristics of the parts of the steam generator feedwater inlet horizontal pipe, thermal stratification takes place particularly frequently. However, the thermal stress due to thermal stratification at the steam generator feedwater inlet horizontal pipe was not reflected in the design stage of old plants(Kori Unit No.1, 2, 3 and 4, Yeonggwang Unit No. 1 and 2, and Uljin Unit No. 1 and 2; referred to as old-style power plants hereinafter). Accordingly, a verification experiment was performed for thermal stratification in the horizontal inlet nozzle steam generator of old-style plants. If thermal stratification occurred in the horizontal pipe of an old-style power plant, numerical analysis of the temperature distribution of the pipes and fluids was conducted. The temperature distributions were compared at the curved part of the pipe and the horizontal pipe before and after the installation of the improved thermal sleeves designed to alleviate thermal stress due to thermal stratification. The thermal stress reduction measure was proven effective at the steam generator inlet horizontal pipe and the curved part of the pipe.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2004.05a
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• pp.490-494
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• 2004

The purpose of this paper was to identify the mechanism that caused abnormal vibration and noise on the piping system connected to discharge flow of BFP(Boiler Feed water Pump) in a coal fired power plant, and to develop the device that can reduce the level of abnormal vibration and noise. Major results of this project can be summarized as follows: First, we analyzed the acoustic mode for the discharge piping of BFP to trace a path of the noise, and assumed that noise and vibration on the piping system can be related with length of pipe. Second, a minimized model of the piping system was set up to simulate abnormal vibration and noise within the specific range of operating frequencies, and as a result we confirmed that the acoustic mode affected the piping system considerably. Finally the test device which can reduce the level of abnormal noise and vibration was built to verify validity applying for the piping system. Then we concluded that the noise and vibration generated from the piping system was attributed to the acoustic resonance in piping system, and so developed new device which can reduce the level of noise and vibration under 40%. Put Abstract here.

• Journal of the Korean Society of Safety
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• v.31 no.6
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• pp.129-134
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• 2016

Standard design of APR+(advanced power reactor plus) was certified at 2014 by Korea regulatory body. Based on the experience gained from OPR1000 and APR1400, the APR1400 was being developed as a 1,500MWe class reactor using Korean technologies for design code, reactor coolant pump, and man-machine interface system. APR+ has been basically designed to have the seismic design basis of safe shutdown earthquake (SSE) 0.3g, a 4-train safety concept based on N+2 design philosophy, and a passive auxiliary feedwater system (PAFS). Also, safety issues on the Fukushima-type accidents have been extensively reviewed and applied to enhance APR+ safety. APR+ provides higher reliability and safety against tsunami and earthquake. The purpose of this paper is to implement probabilistic safety assessment considering these design features and to analyze sensitivity of core damage frequency for large loss of coolant accident of APR+.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.1
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• pp.61-68
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• 2015

The function of Feed Water Ring is to provide the flow path from Feedwater Nozzle to inside of SG(steam generator). Significant amounts of general FAC on the outside of the Feed Water Ring are not likely due to the low flow velocities in this area. However, on the interior of the Feed Water Ring, there may be areas of local higher flow velocity which could lead to higher FAC rates. These may include the inlet tee from the Feedwater Nozzle into the Feed Water Ring, the areas where the Feed Water Ring changes diameter, and especially the entrance area to the J-Nozzles. In this paper, the results of root cause analysis of through-wall hole observed at domestic WH 51F SG Feed Water Ring and its effect on the integrity and performance of SG are described. And, the maintenance strategy for WH 51F SG Feed Water Ring and the monitoring strategy for Downcomer Feed Water Ring of CE System 80 SG are presented.

• Analytical Science and Technology
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• v.8 no.4
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• pp.427-434
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• 1995

Inductively coupled plasma mass spectrometry combined with the isotope dilution method is used for the determination of lithium. The isotope dilution method is based on the addition of a known amount of enriched isotope (spike) to a sample. The analyte concentration is obtained by measuring the altered isotope ratio. The spike solution is calibrated through so called reverse isotope dilution with a primary standard. The spike calibration is an important step to minimize error in the determined concentration. It has been found essential to add spike to a sample and the primary standard so that the two isotope ratios should be as dose as possible. Since lithium is neither corrosive nor toxic, lithium is used as a chemical tracer in the nuclear power plants to measure feedwater flow rate. 99.9% $^7Li$ was injected into a feedwater line of an experimental system and sample were taken downstream to be spiked with 95% $^6Li$ for the isotope dilution measurements. Effects of uncertainties in the spike enrichment and isotope ratio measurement error at various spike-to-sample ratios are presented together with the flow rate measurement results in comparison with a vortex flow meter.