• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2173-2187
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• 2022

A 3D thermo-elasto-hydrodynamic model is developed to analyze the sealing performance of a notched mechanical seal applied in the reactor coolant pump. In the model, the generalized Reynolds equation, the energy equation coupled with notch heat balance equation, the heat conduction equations, and the deformation equations of the sealing rings are iteratively solved by the finite element method. The film pressure and temperature distribution are obtained, and the deformation of the sealing rings is revealed to study the mechanism of the notched mechanical seals. A parameterized study is conducted to analyze the sealing performance under different operating conditions. As a comparison, the sealing performance of non-notched seals is also studied. The results show that the hydrostatic effect is dominant in the load-carrying capacity of the fluid film due to the radial mechanical and thermal deformations. The notch can cool the fluid film and influence the thermal deformation of seal rings. The sealing performance is sensitive to the pressure difference, ambient temperature, and rotational speed. It is suggested to set the notches on the softer sealing rings to acquire the greater hydrodynamic effect. Compared with the non-notched, the notched end face holds a better lubrication performance, especially under lower rotational speed.

• 한국소음진동공학회논문집
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• 제25권9호
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• pp.599-605
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• 2015

U.S. Nuclear Regulatory Commission(NRC), Regulatory Guide(RG) 1.20(Rev.3, 2007년)은 원자로 냉각재계통, 주증기, 주급수 및 복수시스템의 주요 배관 및 기기에 대하여 압력변동 및 진동에 의한 잠재적 유해효과에 대한 평가를 요구한다. 그러나 증기발생기와 연결된 주증기, 주급수 및 복수시스템의 주요 배관 전체에 대하여 상세 해석하는 것은 매우 복잡하여 한계가 있다. 이 논문은 APR1400 원전의 종합진동평가(comprehensive vibration assessment program, CVAP)를 수행하기 위하여 증기발생기에 연결된 2차측 주요 배관의 음향공진과 펌프유발진동을 위한 간이평가 방법에 관한 것이다. 이 논문에서는 이러한 배관시스템의 잠재적 진동 원인이 무엇인지, 음향공진과 펌프유발진동의 가능성을 예방하기 위한 간이평가 방법은 무엇인지를 고찰하고자 한다. 이 논문은 APR1400 원전 증기발생기와 연결된 주증기 및 주급수 배관의 유동유발진동 간이평가를 위해 사용될 것으로 판단된다.

• Journal of Electrical Engineering and Technology
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• 제1권3호
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• pp.359-365
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• 2006

Since the start of the KSTAR (Korea Superconducting Tokamak Advanced Research) project, Instrumentation and Control (I&C) of the Neutral Beam Test Facility (NB-TF) has been striving to answer diverse requests arising from various facets during the project's development and construction phases. Hard-wired electrical circuits have been designed, tested, fabricated, and finally installed to the relevant parts of the system. In relation to the vacuum system I&C, controlling functions for the rotary pumps, a Roots pump, two turbomolecular pumps, and four cryosorption pumps have been constructed. I&C for the ion source operation are the temperature and flow rate signal monitoring, Langmuir probe signal measurements, gradient grid current measurements, and arc detector circuit. For the huge power system to be monitored or safely operated, many temperature measurement functions have also been implemented for the beam line components like the neutralizer, bending magnet, ion dump, and calorimeter. Nearly all of the control and probe signals between the NB test stand and the control room were made to be transmitted through the optical cables. Failures of coolant flow or beam line vacuum pressure were made to be safely blocked from influencing the system by an appropriate interlock circuit that will shut down the extraction voltage application to the system or prevent damages to the vacuum components. Preliminary estimation of the beam power through the calorimetric measurement shows that 87.9% of the total power of the 60kV/18A beam with 200 seconds duration is absorbed by the calorimeter surface. Most of these I&C results would be highly appropriate for the construction of the main NBI facility for the KSTAR national fusion research project.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1955-1962
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• 2020

In a sudden shutdown of primary pump or coolant loss accident in a marine nuclear power plant, the primary flow decreases rapidly in a transition process from forced circulation (FC) to natural circulation (NC), and the lower flow enters the steam generator (SG) causing reverse flow in the U-tube. This can significantly compromise the safety of nuclear power plants. Based on the marine natural circulation steam generator (NCSG), an experimental loop is constructed to study the characteristics of reverse flow under middle-temperature and middle-pressure conditions. The transition from FC to NC is simulated experimentally, and the characteristics of SG reverse flow are studied. On this basis, the experimental loop is numerically modeled using RELAP5/MOD3.3 code for system analysis, and the accuracy of the model is verified according to the experimental data. The influence of the flow variation rate on the reverse flow phenomenon and flow distribution is investigated. The experimental and numerical results show that in comparison with the case of adjusting the mass flow discontinuously, the number of reverse flow tubes increases significantly during the transition from FC to NC, and the reverse flow has a more severe impact on the operating characteristics of the SG. With the increase of flow variation rate, the reverse flow is less likely to occur. The mass flow in the reverse flow U-tubes increases at first and then decreases. When the system is approximately stable, the reverse flow is slightly lower than obverse flow in the same U-tube, while the flow in the obverse flow U-tube increases.

• 한국IT서비스학회지
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• 제8권2호
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• pp.147-156
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• 2009

As the result of the rapid development of IT technology, an on-line diagnostic system using the field bus communication network coupled with a smart sensor module will be widely used at the nuclear power plant in the near future. The smart sensor system is very useful for the prompt understanding of abnormal state of the key equipments installed in the nuclear power plant. In this paper, it is assumed that a smart sensor system based on the fieldbus communication network for the surveillance and diagnostics of safety-critical equipments will be installed in the harsh-environment of the nuclear power plant. It means that the key components of fieldbus communication system including microprocessor, FPGA, and ASIC devices, are to be installed in the RPV (reactor pressure vessel) and the RCS (reactor coolant system) area, which is the area of a high dose-rate gamma irradiation fields. Gamma radiation constraints for the DBA (design basis accident) qualification of the RTD sensor installed in the harsh environment of nuclear power plant, are typically on the order of 4 kGy/h. In order to use a field bus communication network as an ad-hoc diagnostics sensor network in the vicinity of the RCS pump area of the nuclear power plant, the robust survivability of IT-based micro-electronic components in such intense gamma-radiation fields therefore should be verified. An intelligent CCD camera system, which are composed of advanced micro-electronics devices based on IT technology, have been gamma irradiated at the dose rate of about 4.2kGy/h during an hour UP to a total dose of 4kGy. The degradation performance of the gamma irradiated CCD camera system is explained.

• 한국재료학회지
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• 제17권7호
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• pp.366-370
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• 2007

Silicon nitride ($Si_3N_4$) ceramics have been considered for various components of nuclear power plants such as the mechanical seal of a reactor coolant pump (RCP), the guide roller for a control rod drive mechanism (CRDM), and a seal support, etc. Corrosion behavior of $Si_3N_4$ ceramics in a high-temperature and high-pressure water must be elucidated before they can be considered as components for nuclear power plants. In this study, the corrosion behaviors of $Si_3N_4$ ceramics containing MgO and $Al_2O_3$ as sintering aids were investigated at a hydrothermal condition ($300^{\circ}C$, 9.0 MPa) in pure water and 35 ppm LiOH solution. The corrosion reactions were controlled by a diffusion of the reactive species and/or products through the corroded layer. The grain-boundary phase was preferentially corroded in pure water whereas the $Si_3N_4$ grain seemed to be corroded at a similar rate to the grain-boundary phase in LiOH solution. Flexural strengths of the $Si_3N_4$ ceramics were significantly degraded due to the corrosion reaction. Results of this study imply that a variation of the sintering aids and/or a control (e.g., crystallization) of the grain-boundary phase are necessary to increase the corrosion resistance of $Si_3N_4$ ceramics in a high-temperature water.

• 한국자동차공학회논문집
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• 제15권3호
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• pp.133-140
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• 2007

The purpose of this study was to evaluate the effects of gasoline fuel to the LPI engine. Firing test bench was used in order to assess the effect on gasoline-injected LPI engine. Gasoline fuel was supplied into the reverse direction(3-4-2-1 cylinder) at 3.0 bar with commercial gasoline fuel pump. Engine test was performed using the firing test mode at end of line. The deviations of excess air ratio of each cylinder and maximum combustion pressure using gasoline fuel were within 0.1 and $1{\sim}2\;bar$. Engine start time was measured with changing coolant temperature at $20^{\circ}C,\;40^{\circ}C,\;80^{\circ}C$, respectively. Residual gasoline volume in the fuel line was measured about 32 cc after firing test and it was less than 2 cc within 10 seconds purging. To simulate the end of line, the residual gasoline in the fuel line was purged during 5 and 10 seconds. Start time of LPI engine with LPG fuel were 0.61 and 0.58 seconds. This work showed that severe problems such as misfiring and liner scuffing were not occurred applying gasoline fuel to LPI engine.

• 에너지공학
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• 제20권4호
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• pp.259-266
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• 2011

원자로 정지냉각계통은 원자로 정지 시 핵연료 잔열 제거를 위하여 냉각수가 충분히 공급하고 원자로기기들을 보호할 수 있는 냉각율을 유지할 수 있도록 설계되어야 한다. 경수로 정지냉각계통을 분석하기 위한 KDESCENT코드를 중수로 정지냉각계통에 적용하여 보았으며 기존의 중수로형 해석코드인 SOPHT, SDCS 코드 결과와 비교분석하였다. 정지냉각펌프 모드와 열수송펌프 모드에서 정상냉각 운전상태는 계통의 설계 요건을 만족시켰으며 정지냉각 열교환기를 열제거원으로 사용하였을 때 냉각률은 설계요건에서 규정하고 있는 제한치인 $2.8^{\circ}C/min$ 이하의 값을 얻었다. 전반적인 냉각능력 분석 결과 월성 2, 3, 4호기 정지냉각계통은 핵연료로부터 핵분열 생성물의 방출을 충분히 제한하고 핵연료채널의 건전성을 유지시키기 위한 충분한 냉각을 핵연료에 제공하였다.

• 대한설비공학회:학술대회논문집
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• 대한설비공학회 2005년도 동계학술발표대회 논문집
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• pp.353-358
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• 2005

The heat transfer and pressure drop of supercritical $CO_2$ cooled in a helically coiled tube was investigated experimentally. The experiments were conducted without oil in the refrigerant loop. The experimental apparatus of the refrigerant loop consist of receiver, a variable speed pump, a mass flowmeter, a pre-heater, a gas cooler(test section) and an isothermal tank. The test section is a helically coiled tube in tube counter flow heat exchanger with $CO_2$ flowed inside the inner tube and coolant( water) flowed along the outside annular passage, It was made of it copper tube with the inner diameter of 4.55[mm]. the outer diameter of 6.35 [mm] and length of 10000 [mm]. The refrigerant mass fluxes were $200^{\sim}600$ [kg/m2s] and the inlet pressure of gas cooler varied from 7.5 [MPa] to 10.0 [MPa]. The main results are summarized as follows : The heat transfer coefficient of supercritical $CO_2$ increases, as the cooling pressure of gas cooler decreases. And the heat transfer coefficient increases with the increase of the refrigerant mass flux. The pressure drop decreases in increase of the gas cooler pressure and increases with increase the refrigerant mass flux.

• 한국수소및신에너지학회논문집
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• 제23권4호
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• pp.323-329
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• 2012

A thermal management system of a proton exchange membrane fuel cell is taken charge of controlling the temperature of fuel cell stack by rejection of electrochemically reacted heat. Two major components of thermal management system are heat exchanger and pump which determines required amount of heat. Since the performance and durability of PEMFC system is sensitive to the operating temperature and temperature distribution inside the stack, it is necessary to control the thermal management system properly under guidance of operating strategy. The control study of the thermal management system is able to be boosted up with hardware in the loop simulation which directly connects the plant simulation with real hardware components. In this study, the plant simulation of fuel cell stack has been developed and the simulation model is connected with virtual data acquisition system. And HIL simulator has been developed to control the coolant supply system for the study of PEMFC thermal management system. The virtual data acquisition system and the HIL simulator are developed under LabVIEWTM Platform and the Simulation interface toolkit integrates the fuel cell plant simulator with the virtual DAQ display and HIL simulator.