• Journal of the Computational Structural Engineering Institute of Korea
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• v.18 no.4 s.70
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• pp.347-360
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• 2005

The fuel rod instability can be occurred because of the axial and cross flow due to the flow anomaly and/or flow redistribution in the lower core plate region of the pressurized water reactor. The fuel rod vibration due to the hydraulic instability is one of the root causes of fuel failure. The verification on the fuel rod vibration and instability is needed for the new fuel assembly design to verify the fuel rod instability. In this study, the fuel rod vibration and stability analyses were performed to investigate the effect of the grid height, fuel rod support condition, and span adjustment on the fuel rod vibration characteristics for the advanced $16{\times}16$ fuel assembly design. Based on the analysis results, the grid height and grid axial elevation of the advanced $16{\times}16$ fuel assembly design were proposed.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2011.10a
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• pp.40-41
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• 2011

• Clean Technology
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• v.14 no.1
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• pp.61-68
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• 2008

There have been great attentions on polymer electrolyte membrane fuel cell (PEMFC) due to their advantageous characteristics such as zero emission of hazardous pollutant and high energy density. In this work, we evaluated degradation phenomena and stability of single cell performance via one dimensional single cell modeling. Here, CO poisoning on anode on anode was considered for cell performance degradation. Modeling results showed that the performance and stability were highly degraded with CO concentration in fuel gas. In addition, cell performance was reduced by slow oxygen reduction on cathode in long term operation. In order to overcome, it is required to increase ratio o#hydrogen in the fuel gas of anode and high Pt loading contained in the cathodic catalyst layer.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.16 no.11 s.116
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• pp.1115-1123
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• 2006

Nuclear fuel rods are exposed to axial flow in a reactor, and flow-induced-vibration due to the flow usually causes damage in the fuel rods. Thus a prior knowledge about dynamic behavior of a fuel rod exposed to the flow condition should be provided. This paper shows that dynamic characteristics of a nuclear fuel rod depend on axial flow velocity. Assuming small lateral displacement, the effects of uniform axial flow are investigated. The analytic results show that axial flow generally reduces fuel rod stiffness and raises its damping in normal condition. Also, the critical axial velocities which make the fuel rod behavior unstable were found. That is, solving generalized eigenvalue equation of the fuel rod dynamic system, the eigenvalues with positive real part are detected. Based on the simulation results, on the other hand, it turns out that the ordinary axial flow in nuclear reactors does not affect to stability of a nuclear fuel rod even in the conservative condition.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.40 no.5
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• pp.437-447
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• 2016

In a nuclear power plant, the fuel assembly, which is composed of fuel rods, burns, and the high temperature can generate power. The fuel rod consists of pellets and a cladding that covers the pellets. It is important to understand the pellet-cladding mechanical interaction with regard to nuclear safety. This paper proposes simulation of the PCMI. The gap between the pellets and the cladding, and the contact pressure are very important for conducting thermal analysis. Since the gap conductance is not known, it has to be determined by a suitable method. This paper suggests a solution. In this study, finite element (FE) contact analysis is conducted considering thermal expansion of the pellets. As the contact causes plastic deformation, this aspect is considered in the analysis. A 3D FE module is developed to analyze the PCMI using FORTRAN 90. The plastic deformation due to the contact between the pellets and the cladding is the major physical phenomenon. The simple analytical solution of a cylinder is proposed and compared with the fuel rod performance code results.

• Proceedings of the Korean Vacuum Society Conference
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• 2012.08a
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• pp.243-243
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• 2012

화석연료를 대체할 새로운 청정 에너지원의 요구가 높아지고 있는 현 시점에서 고효율, 무공해, 무소음 등의 장점으로 인해 친환경적 에너지원으로 연료전지의 수요가 증가하고 있다. 연료전지 분리판으로 고밀도 흑연을 종래에 가공하여 제작하였는데, 가공이 어렵고, 비용이 크게 들며, 대량생산이 어렵다는 등의 문제로 스테인리스강을 위주로 한 금속 분리판 개발이 이루어지고 있다. 본 연구에서는, 낮은 가격, 고속 증착, 우수한 가공성, 높은 기계적 강도 및 전기전도도, 화학적 안정성 및 내식성을 충족시키기 위하여 스테인리스 강박(0.1 mm이하)에 보호막으로 CrN을 선택하였다. 저가격화를 위하여 새로운 증착원인 스퍼터-승화형 소스의 가능성을 유도 결합 플라즈마에 Cr 봉을 직류 바이어스 함으로써 시도하였다. 10 mTorr의 Ar 유도 결합 플라즈마를 2.4 MHz, 400 W로 유지하면서 직류 바이어스 전력을 120 W (615 V, 0.19 A) 인가하였을 때 10분 동안의 증발양이 0.35 gr으로 측정이 되어 그 가능성을 확인할 수 있었다. 또한 OES(Optical emission spectrometer)를 이용하여 RPS로 방전시킨 고밀도 ICP를 측정한 결과 842.4 nm, 811.4 nm, 772.3 nm 등의 파장에서 높은 intensity를 갖는 peak을 찾을 수 있었고, 이 peak 들은 Ar 중성의 peak임을 확인할 수 있었다. ICP+DC bias로 Cr rod를 가열하는 공정에서의 plasma를 OES로 측정한 결과 Ar 중성의 peak은 감소하고, 520.5 nm, 425.5 nm, 357.7 nm 등의 파장에서 높은 intensity를 갖는 peak을 찾을 수 있었으며, 이 peak들은 Cr 중성의 peak임을 확인할 수 있었다. OES 측정 data를 토대로 공정 중의 rod type Cr target의 교체 주기를 예측할 수 있고 공정 중 실시간 감시가 가능할 것으로 기대된다.

• 한국전산유체공학회:학술대회논문집
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• 2008.03b
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• pp.264-267
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• 2008

Dual-cooled fuel with inner and outer flow channel was proposed for high burup, next generation nuclear fuel design. The annular cylinder of dual cooled fuel has higher structural strength compared to the conventional one, but also have concerns about flow induced vibration due to an additional flow of inner channel and the difference of flow velocity in between inner and outer channel. In this study, the dynamic stability of flexible, annular cylinder was evaluated according to the flow variation and compared to the that of the conventional PWR fuel rod. Centrifugal and Coriolis force by the additional flow in the inner channel were added in the dynamic equation of flexible beam in uniform, external, and axial flow. Complex eigenfrequency was calculated by the finite element method. Stability margin of annular cylinder compared to the solid cylinder and change of the dynamic characteristic are presented and discussed as a analysis results.

• Journal of Korean Society of Environmental Engineers
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• v.34 no.10
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• pp.671-677
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• 2012

In this study, the characteristics of emulsified fuel and engine emissions were studied with an engine dynamometer. In the results of physical property analysis, the margin of error of net calorific value and gross calorific value was ${\pm}0.5%$, were almost same theoretical calculation and results of physical property analysis test. In emulsified fuel, density and viscosity increased with increasing water contents. Emulsified fuel which is composed of water and Bunker-A was manufactured by using homogenizer and ultrasonic generator in $80^{\circ}C$. Phase separation did not take place in $20^{\circ}C$ and $50^{\circ}C$. In the results of engine dynamometer test, NOx concentration and smoke density were reduced with increasing water contents in using emulsified fuel. Total NOx could be reduced by about 41%, 10%, 32% and 28% at 1,000 rpm, 1,200 rpm, 1,500 rpm and 2,500 rpm respectively. Total smoke density was reduced to 42%, 65%, 70%, 62%, and 82% at 1,000 rpm, 1,200 rpm, 1,500 rpm, 2,000 rpm, and 2,500 rpm respectively.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.20 no.3
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• pp.215-222
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• 2010

Long and slender body, like a fuel rod, oscillating in axial flow can be unstabilized even by the small cross flow which can be activated by the flow mixer or turbulent generator. It is important to include these effects of flow disturbance in dynamic stability analysis of nuclear fuel rod. This work shows how eigen frequency of a multi-span fuel rod can be changed by the swirl flow, which is discretely generated by a flow mixer. By solving a state-space form of the eigenvalue equation for a multi-span fuel rod system, the critical velocity at which a fuel rod becomes unstable was calculated. Based on the simulation results, we evaluated how stability of a multi-spanned nuclear fuel rod with mixing vanes can be affected by the coolant flow in an operating reactor core.