• The Journal of the Institute of Internet, Broadcasting and Communication
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• v.24 no.5
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• pp.195-202
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• 2024

Since the permanent shutdown of the Kori No. 1 reactor, research on nuclear power plant decommissioning has been actively conducted. The core facilities (reactor pressure vessel, steam generator, reactor coolant pump, and pressurizer) of a nuclear power plant have the highest radioactivity among the structures of a nuclear power plant, and the reactor pressure vessel (RPV) is the most radioactive object other than the nuclear fuel. In order to dismantle them, accurate preliminary information (2D, 3D models, etc.) and radiological characterization of the dismantling object are required, as well as feasibility studies of dismantling equipment and dismantling processes. However, it is impossible to review the dismantling process with only prior information and radiological characterization, and when using physical mock-ups, simulation and training in a virtual environment are necessary due to the difficulty of applying various dismantling equipment. In this paper, we developed a remote decommissioning training system that can improve the remote decommissioning technology of the nuclear power plant decommissioning process and the decommissioning skills of decommissioning workers by applying virtual reality and haptic technology.

• Nuclear Engineering and Technology
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• v.56 no.11
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• pp.4880-4886
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• 2024

The approach for determining the break size of recirculation loops in a multiple-loop power plant in the event of a loss of coolant accident (LOCA) is presented in this study. In this study, the MAAP5 simulation program was used. An approach to measuring the size of a crack or break in the cooling system is the temperature difference between the recirculation loops. This method does not require any additional facilities; it compares the temperatures of the cooling loops to determine which one has a rupture. The best data source was the loop monitoring system, which sends temperature data for analysis to the main control room. A real operating power reactor training simulator and the FSAR are applied to evaluate MAAP5, the methodology's engine. The results of the MAAP5 simulation code were consistent with those of the power plant simulator. Therefore, MAAP5 could produce enough analytical data to create the relationship diagram between temperature difference and break size. The study hypothesized that there exists a maximum value of temperature difference corresponding to each break size and suggested that applying the absolute maximum temperature difference can aid in identifying the break size. This approach proposes an assistive method for determining the size of a fracture or break in the recirculation system by leveraging the temperature difference between each loop. This approach eliminates the need for additional facilities, as temperature data from the recirculation loops can be transmitted to the main control room. After the reactor scram, operators can monitor the maximum temperature differences at the inlet to estimate the break size. Although the fitting curve used to preliminary estimate the Large Break Loss of Coolant Accident break size may overestimate the break size, it still provides valuable insights. This novel tool offers a rapid and comprehensive method for detecting LOCA events in the recirculation loops.

• Nuclear Engineering and Technology
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• v.50 no.4
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• pp.582-588
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• 2018

Accident diagnosis is one of the complex tasks for nuclear power plant (NPP) operators. In abnormal or emergency situations, the diagnostic activity of the NPP states is burdensome though necessary. Numerous computer-based methods and operator support systems have been suggested to address this problem. Among them, the recurrent neural network (RNN) has performed well at analyzing time series data. This study proposes an algorithm for accident diagnosis using long short-term memory (LSTM), which is a kind of RNN, which improves the limitation for time reflection. The algorithm consists of preprocessing, the LSTM network, and postprocessing. In the LSTM-based algorithm, preprocessed input variables are calculated to output the accident diagnosis results. The outputs are also postprocessed using softmax to determine the ranking of accident diagnosis results with probabilities. This algorithm was trained using a compact nuclear simulator for several accidents: a loss of coolant accident, a steam generator tube rupture, and a main steam line break. The trained algorithm was also tested to demonstrate the feasibility of diagnosing NPP accidents.

• Proceedings of the Korea Society for Simulation Conference
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• 2005.11a
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• pp.153-158
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• 2005

영광 원자력발전소 1호기가 16주기로 운전됨에 따라 훈련용 시뮬레이터의 입력자료 또한 16주기가 반영되어야 한다. 시뮬레이터의 여러 모델 중 노심모델(REMARK)에 필요한 입력자료는 Westinghouse의 핵 설계 코드체계인 APA 시스템의 Output에서 얻을 수 있으나 그 양이 방대하기 때문에 수작업을 통한 입력자료 생산은 큰 어려움을 갖는다. 따라서 이러한 작업을 수행할 프로그램 개발이 필수적이며 개발된 프로그램을 매 교체주기마다 적용하여 노심모델에 대한 원활한 입력상수 생산을 가능하게 할 수 있다.

• Proceedings of the Korea Society for Simulation Conference
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• 2005.11a
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• pp.165-170
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• 2005

원전 시뮬레이터의 노심모델(REMARK)은 원자로에서 발생하는 1차원 열원을 정확히 모사하고, 정상상태 및 과도상태의 반응도 변화에 따른 중성자속의 거동을 실제 원자로와 유사하게 모사할 수 있어야 한다. 모사의 정확성을 높이기 위해 Tuning 작업이 필수적이나 그 작업 단계가 매우 복잡하여 많은 시간과 노력이 필요하기 때문에 이를 간소화하면서 정확성을 높일 수 있는 Tuning 지원 프로그램을 개발하였다. 개발된 프로그램의 사용결과 신속하고 정확한 Tuning이 가능하였다.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.129-143
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• 2016

Sometimes, we need or try to figure out somebody's thoughts from his or her behaviors such as eye movement, facial expression, gestures, and motions. In safety-critical and complex systems such as nuclear power plants, the inference of operators' thoughts (understanding or diagnosis of a current situation) might provide a lot of opportunities for useful applications, such as development of an improved operator training program, a new type of operator support system, and human performance measures for human factor validation. In this experimental study, a novel method for inference of an operator's thoughts from his or her eye movement data is proposed and evaluated with a nuclear power plant simulator. In the experiments, about 80% of operators' thoughts can be inferred correctly using the proposed method.

• Nuclear Engineering and Technology
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• v.37 no.1
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• pp.118-126
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• 2005

While emergency operating procedures (EOPs) occupy an important role in the management of various abnormal situations in nuclear power plants (NPPs), current technology for the validation of EOPs still largely depends on manual review. A validation method for EOPs of NPPs is thus proposed based on dynamic multi-level flow modeling (MFM). The MFM modeling procedure and the EOP validation procedure are developed and provided in the paper. Application of the proposed method to EOPs of an actual NPP shows that the proposed method provides an efficient means of validating EOPs. It is also found that the information on state transitions in MFM models during the management of abnormal situations is also useful for further analysis on EOPs including their optimization.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2783-2791
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• 2022

Molecular dynamics simulations were performed to predict the behavior of graphite atoms under neutron irradiation using large-scale atomic/molecular massively parallel simulator (LAMMPS) package with adaptive intermolecular reactive empirical bond order (AIREBOM) potential. Defect structures of graphite were compared with results from previous studies by means of density functional theory (DFT) calculations. The quantitative relation between primary knock-on atom (PKA) energy and irradiation damage on graphite was calculated. and the effect of PKA direction on the amount of defects is estimated by counting displaced atoms. Defects are classified into four groups: structural defects, energy defects, vacancies, and near-defect structures, where a structural defect is further subdivided into six types by decision tree method which is one of the supervised machine learning techniques.

• Journal of the Korea Society for Simulation
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• v.3 no.1
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• pp.135-150
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• 1994

일반 산업체와 화력발전소 계측제어시스템에 널리 사용되고 있는 분산제어시스템(Distributed Control System)을 원자력발전소에 사용하기 위해서는 분산제어시스템의 안전성과 신뢰성의 입증이 선결과제이다. 따라서 새로운 시스템을 시뮬레이션에 의해 구현하고 검증하기 위한 시뮬레이터가 필요하게 되었다. 발전소 전범위 시뮬레이터(full Scope Simulator)를 제작 하기에 앞서 발전소 계통 중에서 소규모계통을 대상으로 부분범위 시뮬레이터(Compact Simulator)를 제작하였다. 개발된 DCS 검증용 시뮬레이터의 시스템은 발전소 제어반을 모의한 소프트 패널, 발전소 프로세스을 모이한 계통 모델링 소프트웨어, 그리고 현재 발전소의 아날로그 제어계통을 대신한 DCS 제어 계통등의 세 개의 계통으로 구성하였다. 개발 제작된 시뮬레이터를 이용하여 원자력 발전소 계측제어시스템에 분산제어시스템 적용을 시뮬레이션을 통햐여 구현하였으며 분산제어시스템의 적용 검증작업은 물론 적용을 위한 설계업부에도 DCS 검증용 시뮬레이터가 효과적으로 사용될 수 있음을 알았다.

• Proceedings of the KIEE Conference
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• 2002.07a
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• pp.265-267
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• 2002

This paper deals with an I/O simulator design for performance evaluation of reactor protection systems in nuclear power plants. The I/O simulator provides input signals for the reactor protection system, and acquires output signals from the initiation circuits. The simulator is based on VMEbus system, and all VMEbus boards are developed within the country.