• Journal of the Korea Society for Simulation
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• v.13 no.2
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• pp.13-21
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• 2004

The thermal-hydraulic model ARTS which was based on the RETRAN-3D code adopted in the domestic full-scope power plant simulator which was provided in 1998 by KEPRI. Since ARTS is a generalized code to model the components with control volumes, the smaller time-step size should be used even if converged solution could not get in a single volume. Therefore, dedicated models which do not force to reduce the time-step size are sometimes more suitable in terms of a real-time calculation and robustness. In the case of PRT(Pressurizer Relief Tank) model, it is consist of subcooled water in bottom and non-condensable gas in top. The sparger merged under subcooled water enhances condensation. The complicated thermal-hydraulic phenomena such as condensation, phase separation with existence of non-condensable gas makes difficult to simulate. Therefore, the PRT volume can limit the time-step size if we model it with a general control volume. To prevent the time-step size reduction due to convergence failure for simulating this component, we developed a dedicated model for PRT. The dedicated model was expected to provide substantially more accurate predictions in the analysis of the system transients. The results were resonable in terms of accuracy, real-time simulation, robustness and education of operators, complying with the ANSI/ANS-3.5-1998 simulator software performance criteria and RETRAN-3D results.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.2
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• pp.69-76
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• 2018

The paper presents preliminary investigation results for the effect of the baseline correction in the acceleration excitation method on finite element seismic analysis results (such as accumulated equivalent plastic strain, equivalent plastic strain considering cyclic plasticity, von Mises effective stress, etc) of nuclear safety Class I components. For investigation, finite element elastic-plastic time-history seismic analysis is performed for a surge line including a pressurizer lower head, a pressurizer surge nozzle, a surge piping, and a hot leg surge nozzle using the Chaboche hardening model. Analysis is performed for various seismic loading methods such as acceleration excitation methods with and without the baseline correction, and a displacement excitation method. Comparing finite element analysis results, the effect of the baseline correction is investigated. As a result of the investigation, it is identified that finite element analysis results using the three methods do not show significant difference.

• Nuclear Engineering and Technology
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• v.39 no.2
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• pp.133-142
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• 2007

The Electric Power Research Institute (EPRI) demonstrated a method for monitoring the performance of instrument channels in Topical Report (TR) 104965, 'On-Line Monitoring of Instrument Channel Performance.' This paper presents the results of several models originally developed by EPRI to monitor three nuclear plant sensor sets: Pressurizer Level, Reactor Protection System (RPS) Loop A, and Reactor Coolant System (RCS) Loop A Steam Generator (SG) Level. The sensor sets investigated include one redundant sensor model and two non-redundant sensor models. Each model employs an Auto-Associative Kernel Regression (AAKR) model architecture to predict correct sensor behavior. Performance of each of the developed models is evaluated using four metrics: accuracy, auto-sensitivity, cross-sensitivity, and newly developed Error Uncertainty Limit Monitoring (EULM) detectability. The uncertainty estimate for each model is also calculated through two methods: analytic formulas and Monte Carlo estimation. The uncertainty estimates are verified by calculating confidence interval coverages to assure that 95% of the measured data fall within the confidence intervals. The model performance evaluation identified the Pressurizer Level model as acceptable for on-line monitoring (OLM) implementation. The other two models, RPS Loop A and RCS Loop A SG Level, highlight two common problems that occur in model development and evaluation, namely faulty data and poor signal selection

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.3 no.5
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• pp.337-349
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• 1991

Heat and mass transfer rates to spray water droplets for spray transients in a high pressure vessel have been predicted by two different droplet models: the complete mixing model and the non-mixing model. In this process, the ambient fluid surrounding the droplets is a real-gas mixture composed of saturated steam and noncondensable hydrogen gas at high pressure. The physical properties of the mixture are estimated by applying the concept of compressibility factor and using appropriate correlations. A computer program, DROPHMT, to calculate the heat and mass transfer rates for two different droplet models has been developed. As an illustrative application of the computer program to engineering practices, heat and mass transfer rates to spray water droplets for spray transients in a Pressurized Water Reactor (PWR) pressurizer have been calculated, and the typical results have been provided.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.24 no.2 s.173
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• pp.386-393
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• 2000

Fracture mechanics analysis (FMA) is an essential work for integrity evaluation of nuclear power plant. The flaws inspected by In-Service Inspection(ISI) should be confirmed by FMA for the decision of the operation status of stop or continuance. The basic data for FMA are the stress of the interested area. The purpose of this research is to develop a system which can obtain stress data efficiently based on various database. Mesh generation program generates mesh using MSC/PATRAN and provides input file for finite element analysis according to the databases (shape, dimension, transient and material). The stress data from the finite element analysis are stored to be stress database so that it can be applied to FMA. As an example, the system developed by this study is applied to pressurizer nozzle and confirmed to be a useful tool for efficient FMA.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2003.11a
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• pp.247-252
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.107-107
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.135-135
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.178-178
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.532-533
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• 2005