• Nuclear Engineering and Technology
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• v.47 no.6
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• pp.669-677
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• 2015

A mechanistic model for void fraction prediction with improved interfacial friction factor in nearly horizontal tubes has been proposed in connection with the development of a condensation model package for the passive auxiliary feedwater system of the Korean Advanced Power Reactor Plus. The model is based on two-phase momentum balance equations to cover various types of fluids, flow conditions, and inclination angles of the flow channel in a separated flow. The void fraction is calculated without any discontinuity at flow regime transitions by considering continuous changes of the interfacial geometric characteristics and interfacial friction factors across three typical separated flows, namely stratified-smooth, stratified-wavy, and annular flows. An evaluation of the proposed model against available experimental data covering various types of fluids and flow regimes showed a satisfactory agreement.

• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.522-523
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.389-394
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• 1996

The conceptual design of Korea Next Generation Reactor (KNGR), which is 3914 MWt PWR, includes the safety depressurization system (SDS) to comply with U.S. NRC's severe accident policy. In this analysis, it is assumed that three Monobloc Sebim valves are adopted for the SDS bleed valves of KNGR. The characteristic of Monobloc Sebim are modeled in the CE-FLASH-4AS/REM code for this analysis. The various feed and bleed (F&B) procedures with Sebim valves are investigated for total loss of feedwater (TLOFW) event. It is found that if operators open two out of three Sebim valves in conjunction with four HPSI pumps before hot leg temperature reaches saturation condition, the decay heat removal and core inventory make-up function can be successfully accomplished. Therefore, this F&B procedure can be used for mitigating the TLOFW event of the KNGR. This result also demonstrates the feasibility of adopting the Monobloc Sebim valves for the SDS of KNGR.

• 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.

• 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.15 no.1
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• pp.33-39
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• 2019

An 1,500MW advanced power reactor required the standard design approval by a Korean regulatory body in 2014. The reactor has been designed to have a 4-train independent safety concept and a passive auxiliary feedwater system (PAFS). The full power risk or core damage frequency (CDF) of 1,500MW advanced power reactor has been reduced more than that of APR1400. However, the risk during the low power and shutdown (LPSD) operation should be reduced because CDF of LPSD is about 4.7 times higher than that of internal full power. The purpose of paper is to analysis design alternatives to reduce risk during the LPSD. This paper suggests design alternatives to reduce risk and presents sensitivity analysis results.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.157-157
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• 2003

• Nuclear Engineering and Technology
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• v.32 no.2
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• pp.157-168
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• 2000

The nuclear steam generator level control system is designed by robust control methods. The feedwater controller is designed by three methods of the H$\infty$, the mixed weight sensitivity and the structured singular value. Then the controller located on the feedback loop of the level control system is designed. For the system performance, the controller of simple PID whose coefficients vary with the power is selected. The simulations show that the system has a good performance with proper stability margins.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.4 no.2
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• pp.50-56
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• 2008

The tubes in heat exchanger are typically made of copper alloy, stainless steel, carbon steel, titanium alloy material. Type-439 ferritic stainless steel is ferromagnetic material, and furnish higher heat transfer rates than austenitic stainless steels and higher resistance to corrosion-induced flaws. Ferritic stainless steel can be found in low-pressure(LP) feedwater heaters and moisture separator reheaters(MSRs) in turbine system. LP feedwater heaters generally utilize thin wall Type-439 stainless steel tubing, whereas MSRs typically employ a heavier wall tubing with integral fins. Service-induced damage can occur on the O.D(outside diameter) surface of Type-439 ferritic stainless steel tubing which is employed for MSRs tubing, and the most typical damage mechanism is vibration-induced tube-to-TSP(tube support plate) wear and fatigue cracking. The wear has been reported that occurs mainly on the OD surface. Accordingly, in this study, we have evaluated the flaw sizing capability of magnetic saturation eddy current technique using magnetic saturation probe and flawed specimen.