• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2642-2649
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• 2023

A real-time unmeasured dynamic response prediction process for the nuclear power plant pressure pipeline is proposed and its performance is tested in the test-loop system (KAERI). The aim of the process is to predict unmeasurable or unreachable dynamic responses such as acceleration, velocity, and displacement by using a limited amount of directly measured physical responses. It is achieved by combining a well-constructed finite element model and robust inverse force identification algorithm. The pressure pipeline system is described by using the displacement-pressure vibro-acoustic formulation to consider fully filled liquid effect inside the pipeline structure. A robust multiphysics modal projection technique is employed for the real-time sensor synchronized prediction. The inverse force identification method is also derived and employed by using Bathe's time integration method to identify the full-field responses of the target system from the modal domain computation. To validate the performance of the proposed process, an experimental test is extensively performed on the nuclear power plant pressure pipeline test-loop under operation conditions. The results show that the proposed identification process could well estimate the unmeasured acceleration in both frequency and time domain faster than 32,768 samples per sec.

• Journal of the Korean Applied Science and Technology
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• v.32 no.1
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• pp.62-71
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• 2015

In order to deal with the globally increasing electric power demand and reduce $CO_2$ emission, complex thermoelectric power plants are being constructed in densely populated downtown areas. As the environmental regulations are continuously strengthened, various facilities like low NOx burner and SCR are being installed to reduce NOx emission. This study is applied using the TMS emission of $NO_2$ from combined heat and power plant located in Goyang-si Gyeonggi-do. Applying data to the computational fluid dynamics(CFD), and compared with the actual measurement results. It is judged that even though there might be differences between actual measurements and CFD results due to the instant changes of wind direction and wind speed according to measurement time during measurement period, modeling results and actual measurement results showed similar concentration at most forecasting areas and therefore, the forecasting concentration could be deducted which is close to actual measurement by calculating the contribution concentration considering the surrounding concentration in the future.

• Journal of Electrical Engineering and Technology
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• v.9 no.5
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• pp.1511-1519
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• 2014

The objective of this paper is to propose a modelling of a small compressed air energy storage system, which drives an induction generator based on a field-oriented control (FOC) principle for a renewable power generation. The proposed system is a hybrid technology of energy storage and electrification, which is developed to use as a small scale of renewable energy power plant. The energy will be transferred from the renewable energy resource to the compressed air energy by reciprocating air compressor to be stored in a pressurized vessel. The energy storage system uses a small compressed air energy storage system, developed as a small unit and installed above ground to avoid site limitation as same as the conventional CAES does. Therefore, it is suitable to be placed at any location. The system is operated in low pressure not more than 15 bar, so, it easy to available component in country and inexpensive. The power generation uses a variable speed induction generator (IG). The relationship of pressure and air flow of the compressed air, which varies continuously during the discharge of compressed air to drive the generator, is considered as a control command. As a result, the generator generates power in wide speed range. Unlike the conventional CAES that used gas turbine, this system does not have any combustion units. Thus, the system does not burn fuel and exhaust pollution. This paper expresses the modelling, thermodynamic analysis simulation and experiment to obtain the characteristic and performance of a new concept of a small compressed air energy storage power plant, which can be helpful in system designing of renewable energy electrification. The system was tested under a range of expansion pressure ratios in order to determine its characteristics and performance. The efficiency of expansion air of 49.34% is calculated, while the efficiency of generator of 60.85% is examined. The overall efficiency of system of approximately 30% is also investigated.

• Plant Journal
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• v.17 no.4
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• pp.35-40
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• 2021

In this study, a technical characteristic test was conducted on the gas turbine generator system of Seoincheon Combined cycle no.6 to derive and verify the model constants. As a result of the generator maximum/minimum reactive power limit test, the maximum reactive power limit is 80 MVar and the minimum is -30 MVar. The generator uses the GENROU model, the field time constant (T'do) is 4.077 s, and the inertial constant (H) is 5.461 P.U. Excitation system used ESST4B model to derive and verify model constants by simulating no-load 2% AVR step test, PSS modeling derived from PSS2A model constants, and simulated and compared measurement data measured when PSS off/on Did. The GGOV1 model was used for the governor-turbine, and the numerical stability of the determined governor-turbine model constant was verified by simulating a 10% governor step test through the PSS/E simulation program

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.113-121
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• 2014

In this work, an integrated model including molecular dynamics and chemical rate theory was implemented to calculate the growth of point defect clusters(PDC) and copper-rich precipitates(CRP) which could change the mechanical properties of reactor pressure vessel(RPV) steels in a nuclear power plant. A number of time-dependent differential equations were established and numerically integrated to estimate the evolution of irradiation defects. The calculation showed that the concentration of the vacancies was higher than that of the self-interstitial atoms. The higher concentration of vacancies induced a formation of the CRPs in the later stage. The size of the CRPs was used to estimate the mechanical property changes in RPV steels, as is the same case with the PDCs. The calculation results were compared with the measured values of yield strength change and Charpy V-notch transition temperature shift, which were obtained from the surveillance test data of Korean light water reactors(LWRs). The estimated values were in fair agreement with the experimental results in spite of the uncertainty of the modeling parameters.

• Nuclear Engineering and Technology
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• v.33 no.2
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• pp.241-253
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• 2001

Under severe accidents, the pressure and temperature response has an important role for the integrity of a nuclear power plant containment. The history of the pressure and temperature is characterized by the amount and state of steam/air mixture in a containment. Recently, the heat transfer rate to the structure surface is supposed to be increased by the wavy interface formed on condensate film. However, in the calculation by using CONTAIN code, the condensation heat transfer on a containment wall is calculated by assuming the smooth interface and has a tendency to be underestimated for safety. In order to obtain the best- estimate heat transfer calculation, we investigated the condensation heat transfer model in CONTAIN 1.2 code and adopted the new forced convection correlation which is considering wavy interface. By using the film tracking model in CONTAIN 1.2 code, the condensate film is treated to consider the effect of wavy interface. And also, it was carried out to investigate the effect of the different cell modelings - 5-cell and 10-cell modeling - for KNGR(Korean Next Generation Reactor) containment phenomena during a severe accident. The effect of wavy interface on condensate film appears to cause the decrease of peak temperature and pressure response . In order to obtain more adequate results, the proper cell modeling was required to consider the proper flow of steam/air mixture.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.119-130
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• 2023

Among the various elements of probabilistic safety assessment (PSA), human failure events (HFEs) and their dependencies are major contributors to the quantification of risk of a nuclear power plant. Currently, the dependency among HFEs is reflected using a post-processing method in PSA, wherein several drawbacks, such as limited propagation of minimal cutsets through the fault tree and improper truncation of minimal cutsets exist. In this paper, we propose a method to model the HFE dependency directly in a fault tree using the if-then-else logic. The proposed method proved to be equivalent to the conventional post-processing method while addressing the drawbacks of the latter. We also developed a software tool to facilitate the implementation of the proposed method considering the need for modeling the dependency between multiple HFEs. We applied the proposed method to a specific case to demonstrate the drawbacks of the conventional post-processing method and the advantages of the proposed method. When applied appropriately under specific conditions, the direct fault-tree modeling of HFE dependency enhances the accuracy of the risk quantification and facilitates the analysis of minimal cutsets.

• Journal of Korean Society for Atmospheric Environment
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• v.19 no.6
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• pp.789-803
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• 2003

The purpose of this study is to compare the usefulness between Gaussian dispersion model and receptor model with the experimental result of the dispersion tracing of the particulate pollutants from Taean coal-fired power plants. For this purpose, the component analysis of the collected PM 10 samples was performed. In order to trace the pollution sources, factor analysis was done with the result of the component analysis. As a result of the correlativity analysis of the fifteen power plants' profiles offered by US EPA, the correlativity of No.11202 source profile showed highest rate up to 84.5%. Thus it was adopted as proper one and the contribution rate by each pollution source was calculated by Chemical Mass Balance (CMB)-8 model. The contribution rate, which was the effect rate of the power plants on each measuring point, were calculated with a range of 24∼52％ and the standard error was below 0.9 $\mu\textrm{g}$/㎥. This indicates the selection of the source profile was appropriate. Also, the concentrations of each point were calculated by the ISCST3 which is suggested by US EPA as one of the regulatory Gaussian dispersion model. The calculation result showed that the predicted concentration was 50∼58 $\mu\textrm{g}$/㎥, comparing with the measured result of 9∼65 $\mu\textrm{g}$/㎥. It was found that the concentration calculated by ISCST3 was underpredicted. It was thought that the receptor model was more favorable than the Gaussian dispersion model in estimating the effect of the particulate matter on a certain receptive point.

• Fire Science and Engineering
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• v.20 no.4 s.64
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• pp.13-18
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• 2006

The fire protection regulation for the nuclear power plants is based on the qualitative fire hazard assessment and the quantitative fire risk analysis, and the fire risk is managed by the fire protection plan with the appropriate balance among the fire prevention, fire suppression and the minimization of the fire effect. In these days, the zone model or the field model is generally used for the detail evaluation for the fire risk. At this paper, with consideration of the present trend, we evaluate whether the quantitative fire risk analysis and the assessment of fire result for fire areas at nuclear power plants can be possible by use of Fire Dynamics Simulator (FDS) that is the state-of-the-art fire modeling tool. Consequently, it is expected that the quantitative fire risk evaluation propelled by the fire modeling can be available as an applicable tool to improve the core damage frequency as well as the quantitative fire risk analysis.

• Journal of the Korean Society of Systems Engineering
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• v.14 no.2
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• pp.24-32
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• 2018

System Complexity one of the most common cause failure of the projects, it leads to a lack of understanding about the functions of the system. Hence, the model is developed for communication and furthermore modeling help analysis, design, and understanding of the system. On the other hand, the text-based specification is useful and easy to develop but is difficult to visualize the physical composition, structure, and behaviour or data exchange of the system. Therefore, it is necessary to transform system description into a diagram which clearly depicts the behaviour of the system as well as the interaction between components. According to the International Atomic Energy Agency (IAEA) Safety Glossary, The safety system is a system important to safety, provided to ensure the safe shutdown of the reactor or the residual heat removal from the reactor core, or to limit the consequences of anticipated operational occurrences and design basis accidents. Core Protection Calculator System (CPCS) in Advanced Power Reactor 1400 (APR 1400) Nuclear Power Plant is a safety critical system. CPCS was developed using systems engineering method focusing on Departure from Nuclear Boiling Ratio (DNBR) calculation. Due to the complexity of the system, many diagrams are needed to minimize the risk of ambiguities and lack of understanding. Using Model-Based Systems Engineering (MBSE) software for modeling the DNBR algorithm were used. These diagrams then serve as the baseline of the reverse engineering process and speeding up the development process. In addition, the use of MBSE ensures that any additional information obtained from auxiliary sources can then be input into the system model, ensuring data consistency.