• Nuclear Engineering and Technology
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• 제55권7호
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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.

• 한국응용과학기술학회지
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• 제32권1호
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• pp.62-71
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• 2015

세계적으로 급증하는 전력수요에 대처하고, $CO_2$ 배출을 줄이고자 인구가 밀집되어 있는 도심지에 복합화력 발전소가 건설되고 있다. 환경규제가 계속적으로 강화됨에 따라 NOx 배출량을 줄이고자 저 NOx 버너, SCR 등 여러 가지 설비들을 설치하고 있다. 본 연구는 경기도 고양시 소재의 일산열병합발전소 1개소에서 배출되는 질소산화물을 TMS를 이용하여 배출계수를 산정하여 이를 전산유체동역학(CFD)에 적용하여 질소산화물의 거동을 살펴보고, 현장 실측 결과와 비교 검토하였다. 실측 기간 중 측정 시간에 따른 주 풍향 풍속의 순간적인 변화로 인해 실측 결과와 CFD 모델링 결과의 차이가 나타날 수 있으나, 모델링 결과와 실측 결과는 대부분 예측지점에서 유사한 농도로 나타났다. 향후 주변농도를 고려한 기여농도를 산출하여 실측농도에 가까운 예측농도 도출이 가능 할 것으로 판단된다.

• Journal of Electrical Engineering and Technology
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• 제9권5호
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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.

• 플랜트 저널
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• 제17권4호
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• pp.35-40
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• 2021

본 연구에서는 서인천복합 6호기 가스터빈 발전기계통에 대하여 기술특성시험을 실시하여 모델정수를 도출하고 검증하였다. 발전기 최대/최소 무효전력 한계시험 결과 최대 무효전력 한계는 80 MVar이고, 최소 무효전력 한계는 -30 MVar이다. 발전기는 GENROU 모델을 사용하였고, 계자시정수(T'do)는 4.077 s, 관성정수(H)는 5.461 P.U로 결정하였다. 여자시스템은 ESST4B 모델을 사용, 무부하 2% AVR 스텝시험을 모의하는 방식으로 모델정수를 도출하고 검증하였으며, PSS 모델링은 PSS2A 모델정수로 도출하였고, PSS Off/On일때 측정된 계측 데이터를 모의, 비교하여 검증하였다. 조속기-터빈는 GGOV1 모델을 사용하여 모델정수를 도출하고 검증하였다. PSS/E 시뮬레이션 프로그램을 통해 10% 조속기 스텝시험을 모의하여 결정된 조속기-터빈 모델정수의 수치 안정성을 확인하였다.

• 한국압력기기공학회 논문집
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• 제10권1호
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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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• 제33권2호
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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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• 제55권1호
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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.

• 한국대기환경학회지
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• 제19권6호
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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.

• 한국화재소방학회논문지
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• 제20권4호
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• pp.13-18
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• 2006

원자력발전소의 화재방호규정은 정성적인 화재위험성평가와 정량적인 화재위험도분석에 기반을 두며, 화재위험은 심층화재방어개념인 화재 예방, 화재 진압, 및 피해 최소화의 3가지 요소에 균형을 유지하면서 화재방호계획에 의해 관리되고 있다. 최근 화재위험 상세평가는 일반적으로 존모델 또는 필드모델을 이용하고 있다. 본 논문에서는 이런 추세에 따라 최신 화재모델링 도구인 FDS를 이용하여 원자력 발전소의 방화지역에 대한 정량적 화재위험분석 및 화재영향 평가가 가능한지 그 여부를 확인하였다. 이의 결과 화재모델링을 이용한 정량적 위험분석은 원자력발전소의 방화지역에 대한 정량적 위험도 분석뿐만 아니라 화재로 인한 원자로 노심 손상빈도를 개선할 수 있는 응용 도구로 활용될 수 있을 것으로 기대된다.

• 시스템엔지니어링학술지
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• 제14권2호
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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.