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Accurate Circuit Model of Patterned Vertical Alignment (PVA) Liquid Crystal Displays
Chansoo Park,Youngmin Cho,Joon-Chul Goh,Chong Chul Chai,Seung-Woo Lee 대한전자공학회 2010 ICEIC:International Conference on Electronics, Inf Vol.1 No.1
We propose an accurate circuit model of patterned vertical alignment (PVA) liquid crystal (LC). The model adopts a first-order macro model described by Verilog-A. To improve accuracy of the transient response of PVA LCD panel, we propose T-V curve approximation method. We report a delay at the first frame of gray-to-gray transition, which results in large deviation. We achieve highly accurate simulation results by taking on account of the delay in PVA panel.
Estimation of the scale parameter of the Rayleigh distribution under general progressive censoring
Chansoo Kim,Keunhee Han 한국통계학회 2009 Journal of the Korean Statistical Society Vol.38 No.3
Based on a general progressively type II censored sample, the maximum likelihood estimator (MLE), Bayes estimator under squared error loss and credible intervals for the scale parameter and the reliability function of the Rayleigh distribution are derived. Also, the Bayes predictive estimator and highest posterior density (HPD) prediction interval for future observation are considered. Comparisons among estimators are investigated through Monte Carlo simulations. An illustrative example with real data concerning 23 ball bearings in a life test is presented.
Chansoo Lee,Youho Lee 한국방사성폐기물학회 2023 한국방사성폐기물학회 학술논문요약집 Vol.21 No.2
Dry storage of nuclear fuel is compromised by threats to the cladding integrity, such as creep and hydride reorientation. To predict these phenomena, spent fuel simulation codes have been developed. In spent fuel simulation, temperature information is the most influential factor for creep and hydride formation. Traditional fuel simulation codes required a user-defined temperature history input which is given by separate thermal analysis. Moreover, geometric changes in nuclear fuel, such as creep, can alter the cask’s internal subchannels, thereby changing the thermal analysis. This necessitates the development of a coupled thermal and nuclear fuel analysis code. In this study, we integrated the 2D FDM nuclear fuel code GIFT developed at SNU with COBRA -SFS. Using this, we analyzed spent nuclear stored in TN-24P dry storage cask over several decades and identified conditions posing threats due to phenomena like creep and hydrogen reorientation, represented by the burnup and peak cladding temperature at the start of dry storage. We also investigated the safety zone of spent nuclear fuel based on burnup and wet storage duration using decay heat.