• Journal of Mechanical Science and Technology
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• 제19권5호
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• pp.1206-1215
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• 2005

The thermal stratification phenomena, frequently occurring in the component of nuclear power plant system such as pressurizer surge line, steam generator inlet nozzle, safety injection system (SIS), and chemical and volume control system (CVCS), can cause through-wall cracks, thermal fatigue, unexpected piping displacement and dislocation, and pipe support damage. The phenomenon is one of the unaccounted load in the design stage. However, the load have been found to be serious as nuclear power plant operation experience accumulates. In particular, the thermal stratification by the turbulent penetration or valve leak in the SIS and SCS pipe line can lead these safety systems to failure by the thermal fatigue. Therefore in this study an 1/10 scaledowned experimental rig had been designed and installed. And a series of experimental works had been executed to measure the temperature distribution (thermal stratification) in these systems by the turbulent penetration, valve leak, and heat transfer through valve. The results provide very valuable informations such as turbulent penetration depth, the possibility of thermal stratification by the heat transfer through valve, etc. Also the results are expected to be useful to understand the thermal stratification in these systems, establish the thermal strati­fication criteria and validate the calculation results by CFD Codes such as Fluent, Phenix, CFX.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.571-576
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• 1996

Currently most nuclear power plants(NPPs) are adopted the mid-loop operation to minimize the overhaul period and save the operating cost. For mid-loop operation it is essential to install nozzle dam between RCS pipe and steam generator(SG). Because SG remains more highly contaminated with radioactive material than any other parts of the NPPs, the repairmen are very reluctant to carry out installing nozzle dam inside the SG. Until now, unfortunately, it appears that no practically applicable device was developed to provide the longstanding demand. Also the accidents have been reported by licenser event report during this operation mode due to loss of residual heat removal(RHR). The purpose of this paper is to conduct remotely blocking and disintegration of nozzle of a SG which has the highest radiation exposure during the maintenance in NPPs. The remote nozzle blocking device of a SG includes three bladders, hubs, air controller provisions to supply and contact air pressure into the bladders. This remote nozzle block device will give the larger operation margin to prevent the loss of RHR and minimize the radiation exposure dose to the repairman and shorten the overhaul periods.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3635-3642
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• 2021

Natural circulation systems (NCSs) are extensively applied in nuclear power plants because of their simplicity and inherent safety features. For some passive natural circulation systems in floating nuclear power plants (FNPPs), the ocean is commonly used as the heat sink. Condensation induced water hammer (CIWH) events may appear as the steam directly contacts the subcooled seawater, which seriously threatens the safe operation and integrity of the NCSs. Nevertheless, the research on the formation mechanisms of CIWH is insufficient, especially in NCSs. In this paper, the characteristics of flow rate and fluid temperature are emphatically analyzed. Then the formation types of CIWH are identified by visualization method. The experimental results reveal that due to the different size and formation periods of steam slugs, the flow rate presents continuous and irregular oscillation. The fluid in the horizontal hot pipe section near the water tank is always subcooled due to the reverse flow phenomenon. Moreover, the transition from stratified flow to slug flow can cause CIWH and enhance flow instability. Three types of formation mechanisms of CIWH, including the Kelvin-Helmholtz instability, the interaction of solitary wave and interface wave, and the pressure wave induced by CIWH, are obtained by identifying 67 CIWH events.

• 한국수소및신에너지학회논문집
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• 제10권1호
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• pp.41-48
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• 1999

산업단지에서 손실되는 막대한 폐열을 효율적으로 회수하고 이를 인근의 배후 도시에서 활용하기 위해서는 이에 적합한 열수송기술이 필요하다. 현재 온수나 증기에 의한 열수송은 배관을 통하여 열손실 및 마찰손실 등이 발생하므로 수송거리는 3 내지 5km가 한계이다. 그러나 대부분의 공단이 도시지역에서 10km 이상 떨어져 있으므로 이들 지역에서 발생되는 폐열을 적절히 활용하기 위해서는 새로운 열수송시스템이 개발되어야 한다. 본 연구에서는 수소저장합금이 수소를 흡수 또는 방출하면서 발열반응과 흡열반응을 일으키는 특성을 이용하여 산업공단지역의 폐열로부터 수소저장합금의 수소를 방출시키고, 이 수소를 인근 도시지역에 파이프라인으로 수송한 후 필요시 또 다른 수소저장합금과 반응시켜 열을 얻을 수 있는 열수송시스템에 대하여 고찰하였다. 이 시스템에서는 난방의 목적 외에도 수소의 흡수 방출온도가 낮은 합금을 이용하여 냉열을 얻을 수도 있으며, 폐열의 저장수단으로, 또한 수소를 수송함으로서 열수송의 수단으로 활용할 수 있다. 이에 따라 수소저장합금을 이용한 열수송기술의 문제점과 열수송시스템의 구성기술에 대하여도 검토하였다.

• Corrosion Science and Technology
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• 제12권3호
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• pp.125-131
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• 2013

The most common pipe wall thinning degradation mechanisms that can occur in the steam and feedwater systems are FAC (Flow Acceleration Corrosion), cavitation, flashing, and LDIE (Liquid Droplet Impingement Erosion). Among those degradation mechanisms, FAC has been investigated by many laboratories and industries. Cavitation and flashing are also protected on the piping design phase. LDIE has mainly investigated in aviation industry and turbine blade manufactures. On the other hand, LDIE has been little studied in NPP (Nuclear Power Plant) industry. This paper presents the development of prediction system for pipe wall thinning caused by LDIE in terms of erosion rate based on air-water ratio and material. Experiment is conducted in 3 cases of air-water ratio 0.79, 1.00, and 1.72 using the three types of the materials of A106B, SS400, and A6061. The main control parameter is the air-water ratio which is defined as the volumetric ratio of water to air (0.79, 1.00, 1.72). The experiments were performed for 15 days, and the surface morphology and hardness of the materials were examined for every 5 days. Since the spraying velocity (v) of liquid droplets and their contact area ($A_c$) on specimens are changed according to the air-water ratio, we analyzed the behavior of LDIE for the materials. Finally, the prediction equations(i.e. erosion rate) for LDIE of the materials were determined in the range of the air-water ratio from 0 to 2%.

• 비파괴검사학회지
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• 제23권4호
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• pp.364-371
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• 2003

In this paper, a review of the current status, on the use of long range cylindrically guided wave modes, and their interaction with cracks and corrosion damage in pipe-like structures will be discussed. Applications of cylindrically guided ultrasonic wave modes have been developed for inspection of corrosion damage in pipelines at chemical plants, flow-accelerated corrosion damage (wall thinning) in feedwater piping, and circumferential stress corrosion cracks in PWR steam generator tubes. It has been demonstrated that this inspection technique can be employed on a variety of piping geometries (diameters from 1 in. to 3 ft, and wall thickness from 0.1 to 6 in.) and a propagation distance of 100 meters or more is sometimes feasible. This technique can also be used in the inspection of inaccessible or buried regions of pipes and tubes.

• 한국유체기계학회 논문집
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• 제6권1호
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• pp.57-65
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• 2003

The measurement uncertainties of an ultrasonic flowmeter were analyzed to evaluate its applicability to the measurement of the steam generator feedwater flow-rate in a nuclear power plant. The analyses of measurement uncertainties of a reactor power were also performed with the analyses of feedwater flow measurement uncertainties. Two ultrasonic flowmeters based on a cross-correlation technique and a transit time method were used in this study. The ultrasonic flowmeters were installed on a feedwater pipe line of a typical 1000 MWe Korea-standardized nuclear power plant to take the necessary data. The results have shown that the measurement uncertainties of the ultrasonic flowmeters are adequately smaller than those or a venturi meter. The research has also indicated that the measurement uncertainties of the reactor power based on the ultrasonic flowmeter uncertainties are sufficiently bounded by the uncertainty range usually assumed in nuclear safety analyses.

• 한국신재생에너지학회:학술대회논문집
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• 한국신재생에너지학회 2009년도 춘계학술대회 논문집
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• pp.813-816
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• 2009

A commercial-scale RDF boiler that its burning capacity is 400 kg-RDF/hr and steam production capacity is 2 ton/hr. It has a chain type stoker and waste heat recovery system. Heat exchanger is vertical water-pipe so that soot blowing and removal is convenient during operation. Dry scrubber, bag filter and activated carbon tower have been installed for the reduction of air pollutant gases and dust. Analysing data of pollutants from stack such as $SO_x$. $NO_x$ and dioxin shows so good results that the boiler system could comply the regulated emission limits.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2003년도 춘계학술대회 논문집
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• pp.1458-1462
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• 2003

High temperature and high pressure heavy water flows through the pipes in atomic power plants. The curved parts of pipes are critical parts in that they change the direction of steam flow, and these parts are especially affected by severe wear. Therefore, most pipes in atomic power plants are tested by non-destructive examination by workers who use ultrasonic sensors to measure the wall thickness of pipes. But not only are these pipes located in a very dangerous environment, but the space is also very limited. For the safety of workers, it is necessary to design a device that uses a mobile robot that can inspect curved pipes. This paper presents the design and construction of a small device that can generate the necessary contact forces between ultrasonic sensors and pipe walls in a limited space. And a mobile robot is used in place ortho worker for successful non-destructive examination.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집A
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• pp.730-735
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• 2000

The finite element analysis by the computer program SYSWELD in consideration of phase transformation was carried out to simulate the multi-pass welding process of SA106 Gr. C which is used for the main steam pipe in nuclear power plant. All the numerical results such as temperatures, the size of heat affected zone and the residual stresses were compared to the experimental results.