• 한국지진공학회논문집
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• 제22권4호
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• pp.245-251
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• 2018

In this paper, we address some issues in existing seismic hazard closed-form equations and present a novel seismic hazard equation form to overcome these issues. The presented equation form is based on higher-order polynomials, which can well describe the seismic hazard information with relatively high non-linearity. The accuracy of the proposed form is illustrated not only in the seismic hazard data itself but also in estimating the annual probability of failure (APF) of the structural systems. For this purpose, the information on seismic hazard is used in representative areas of the United States (West : Los Angeles, Central : Memphis and Kansas, East : Charleston). Examples regarding the APF estimation are the analyses of existing platform structure and nuclear power plant problems. As a result of the numerical example analyses, it is confirmed that the higher-order-polynomial-based hazard form presented in this paper could predict the APF values of the two example structure systems as well as the given seismic hazard data relatively accurately compared with the existing closed-form hazard equations. Therefore, in the future, it is expected that we can derive a new improved APF function by combining the proposed hazard formula with the existing fragility equation.

• 대한기계학회논문집A
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• 제25권3호
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• pp.381-389
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• 2001

The objective of this paper is to evaluate the fracture resistance characteristics of SA508 Cl.1a to SA508 Cl.3 welds manufactured for the reactor coolant loop piping system of nuclear power plants. The effect of the crack plane orientation to the welding process orientation and the preheat temperature on the fracture resistance characteristics were discussed. Results of the fracture resistance test showed that the effect of the crack plane orientation to the welding process orientation of the fracture toughness is significant, while that of preheat temperature on the fracture toughness is negligible. The micro Vickers hardness test, the metallographic observation and the fractography analysis were conducted to analyse the crack jump phenomenon on the L-R crack plane orientation in the multi-pass welding zone. As these results, it is shown that the crack jump phenomenon was produced because of the inhomogeneity between welding beads and the crack plane orientation must be considered for the safety of the welding zone in the piping system.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 1990년도 추계학술대회논문집; 한양대학교, 서울; 24 Nov. 1990
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• pp.113-124
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• 1990

Tap lines are small branch piping generally less than two inches in diameter. They typically branch off of header piping having a much larger diameter. An example of a common tap line is a 3/4 inch size high point vent or low point drain. Most tap lines have at least one valve near the header tap connection to provide isolation. Two valves are often required for double isolation. A light water reactor(LWR) nuclear power plant will have several hundred tap lines. These lines come in many sizes and shapes and serve numerous functions. A single process piping valve may have three different tap lines associated with it (figure 1). Table 1 delineates the different categories of tap lines. Vibration failures of tap lines are a common occurrence in all industrial plants including nuclear and fossil power plants. These types of failures constitute a significant percentage of all piping related failures. An unscheduled plant shutdown or outage resulting from the failure of a tap line decreases plant reliability and may have a detrimental effect on plant safety. Most tap line vibration failures can be avoided through the use of appropriate routing and support techniques. Standardized designs can be developed for use in a myriad of applications. These designs will not only minimize failures but will also reduce the necessary analysis and installation efforts.

• 공업화학
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• 제24권4호
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• pp.337-343
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• 2013

멤브레인 반응기는 멤브레인과 반응기를 결합하여 반응과 분리의 단위공정을 하나로 결합함으로써 전체공정을 단순화하고 반응효율을 높이고자 하는 혁신 기술로써, 멤브레인을 이용한 생성물의 선택적 제거를 통해 열역학적 평형을 뛰어넘는 전환율, 부반응물 생성 억제에 의한 반응 효율 및 선택성을 향상시킬 수 있다. 특히 이온전도성 세라믹을 이용한 멤브레인 반응기는 연료전지의 개발, 고순도 산소/수소의 분리/정제, 이산화탄소의 전환 및 다양한 화학제품제조에 까지 응용될 수 있기 때문에 시장의 확대와 더불어 크게 발전할 수 있을 것으로 기대된다. 본 총설에서는 수소이온 전도성 세라믹 멤브레인 반응기에 대한 연구동향과 다양한 응용분야 및 향후 전망 등에 고찰해 보고자 한다.

• Nuclear Engineering and Technology
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• 제20권1호
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• pp.35-46
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• 1988

원자로의 설계나 안전성 분석을 위해서는 핵연료 집합체 내의 유동 구조와 열전달에 대한 지식이 매우 중요하다. 따라서 핵연료 집합체 내의 유체 온도 분포를 정확히 계산하기 위해서는 냉각재 유로 내에서의 속도분포를 정확히 알아야 하는데 이것은 복잡한 난류 현상 때문에 예측하기가 매우 어렵다. 본 연구는 비등방성을 고려한 2-방정식 모형을 사용하여 속도분포를 구하고 핵연료 표면에서의 균일열속을 가정하므로써 유로내에서의 속도 분포를 예측하였다. 수치해는 Galerkin유한 요소법에 의해 핵연료봉 표면까지 구하여졌다. 수치 결과는 알려진 실험치 및 계산치와 비교되어 잘 일치하고 있고, 또한 난류 비등방성이 유로 내의 평균속도와 온도분포에 영향을 미치고 있음을 보았다. 그리고 조밀한 삼각 배열 핵연료 집합체(P/D=1.05-1.3) 내에서 나트륨 냉각재를 사용한 경우의 Nu-P/D관계식을 수립하였다.

• Nuclear Engineering and Technology
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• 제47권2호
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• pp.165-175
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• 2015

The application of neutron noise analysis (NNA) to the ex-core neutron detector signal for monitoring the vibration characteristics of a reactor core support barrel (CSB) was investigated. Ex-core flux data were generated by using a nonanalog Monte Carlo neutron transport method in a simulated CSB model where the implicit capture and Russian roulette technique were utilized. First and third order beam and shell modes of CSB vibration were modeled based on parallel processing simulation. A NNA module was developed to analyze the ex-core flux data based on its time variation, normalized power spectral density, normalized cross-power spectral density, coherence, and phase differences. The data were then analyzed with a fuzzy logic module to determine the vibration characteristics. The ex-core neutron signal fluctuation was directly proportional to the CSB's vibration observed at 8Hz and15Hzin the beam mode vibration, and at 8Hz in the shell mode vibration. The coherence result between flux pairs was unity at the vibration peak frequencies. A distinct pattern of phase differences was observed for each of the vibration models. The developed fuzzy logic module demonstrated successful recognition of the vibration frequencies, modes, orders, directions, and phase differences within 0.4 ms for the beam and shell mode vibrations.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.650-659
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• 2016

As a type of accident-tolerant fuel, fully ceramic microencapsulated (FCM) fuel was proposed after the Fukushima accident in Japan. The FCM fuel consists of tristructural isotropic particles randomly dispersed in a silicon carbide (SiC) matrix. For a fuel element with such high heterogeneity, we have proposed a two-temperature homogenized model using the particle transport Monte Carlo method for the heat conduction problem. This model distinguishes between fuel-kernel and SiC matrix temperatures. Moreover, the obtained temperature profiles are more realistic than those of other models. In Part I of the paper, homogenized parameters for the FCM fuel in which tristructural isotropic particles are randomly dispersed in the fine lattice stochastic structure are obtained by (1) matching steady-state analytic solutions of the model with the results of particle transport Monte Carlo method for heat conduction problems, and (2) preserving total enthalpies in fuel kernels and SiC matrix. The homogenized parameters have two desirable properties: (1) they are insensitive to boundary conditions such as coolant bulk temperatures and thickness of cladding, and (2) they are independent of operating power density. By performing the Monte Carlo calculations with the temperature-dependent thermal properties of the constituent materials of the FCM fuel, temperature-dependent homogenized parameters are obtained.

• 한국가시화정보학회지
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• 제14권2호
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• pp.52-60
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• 2016

The Passive Auxiliary Feedwater System(PAFS) is one of the advanced safety concepts adopted in the Advanced Power Reactor Plus(APR+). To validate the operational performance of the PAFS, detailed understanding of a boiling heat transfer on horizontal tube outside is of great importance. Especially, in the mechanistic boiling heat transfer model, it is important to visualize the phenomena but there are some limitations with conventional experimental approaches. In the present study, we devised a heater based on the Flexible Printed Circuit Board (FPCB) for a more comprehensive visualization and subsequently, a digital image processing technique for the bubble motion measurement was established. Using the measurement technique, important parameters of the nucleate boiling are analyzed.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집A
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• pp.843-848
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• 2000

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw is found during the periodic inspection from the pressure tubes, the integrity evaluation must be carried out, and the safety requirements must be satisfied for continued service. In order to complete the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the entire integrity evaluation process. For this reason, an integrity evaluation system, which provides efficient way of evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL, and covers the delayed hydride cracking(DHC). Various analysis methods are provided for the integrity evaluation of pressure tube. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 추계학술대회
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• pp.568-573
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• 2003

When a safety valve equipped in a nuclear power plant opens in an instant by an accident, a moving shock wave propagates downstream the valve, inducing a complicated unsteady flow field. The moving shock wave may exert severe load to the structure. So, to reduce the load acting on the wall of POSRV, a gradual opening of POSRV is adopted in general. In theses connections, a numerical work is performed to investigate the effect of valve opening time on the unsteady flow fields downstream of the valve. Compressible, two-dimensional Navier-Stokes equations are used with the finite volume method. The obtained results show that sharp pressure rise through moving shock tor the case of instant opening is attenuated by employing the gradual opening of valve. It is turned that the flows for the two cases of gradual valve opening time show the similar to that of highly under-expanded one in jet structure having expansion and compression waves and Mach stem. Also, comparing with the results for the two cases of opening time, the shorter the valve opening is, the pressure gradient at the downstream of the valve becomes softly.