• International conference on construction engineering and project management
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• 2013.01a
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• pp.451-458
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• 2013

This research documents the initial findings and recommendations for developing a risk management tool to assess and quantify the risks associated with the construction of the next generation of nuclear power plants. The proposed tool builds upon the Construction Industry Institute's International Project Risk Assessment (IPRA) Best Practice. This paper provides an overview of the investigation to assess the unique risk elements pertaining to nuclear power plant construction and documents the preliminary findings from historical project performance data to better understand the function and use of the IPRA's Relative Impact value.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.144-152
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• 2016

Human capabilities, such as technical/nontechnical skills, have begun to be recognized as crucial factors for nuclear safety. One of the most common ways to improve human capabilities in general is training. The nuclear industry has constantly developed and used training as a tool to increase plant efficiency and safety. An integrated training framework was suggested for one of those efforts, especially during simulation training sessions of nuclear power plant operation teams. The developed training evaluation methods are based on measuring the levels of situation awareness of teams in terms of the level of shared confidence and consensus as well as the accuracy of team situation awareness. Verification of the developed methods was conducted by analyzing the training data of real nuclear power plant operation teams. The teams that achieved higher level of shared confidence showed better performance in solving problem situations when coupled with high consensus index values. The accuracy of nuclear power plant operation teams' situation awareness was approximately the same or showed a similar trend as that of senior reactor operators' situation awareness calculated by a situation awareness accuracy index (SAAI). Teams that had higher SAAI values performed better and faster than those that had lower SAAI values.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.59 no.8
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• pp.1449-1455
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• 2010

The excitation control system of an emergency diesel generator is classified as a kind of safety-related system. Compared with other control systems in a power plant, this system is required to be more reliable and have better performance. In this paper, the digital multi-redundant excitation system for a diesel generator was proposed. The signal processing system of the proposed system makes high speed signal processing and arithmetic in excitation control possible. The improved soft start algorithm and multiple PI parameters adaptation considering the diesel generator characteristics were implemented in the proposed system. The developed system was applied to a nuclear power plant successfully.

• Transactions of the Korean Society of Mechanical Engineers
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• v.19 no.1
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• pp.111-121
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• 1995

A fuzzy control algorithm of bell-type membership functions and 9 rules is constructed for narrow range level control of steam generators in nuclear power plants. It is implemented at a field digital distributed controller, a Westinghouse-made controller called Westinghouse Distributed Processing Family(WDPF). Performance for level control of the developed fuzzy controller is compared with that of conventional controller, both at the field controller. For these comparisons, both the fuzzy control algorithm and the conventional PI control algorithm were carefully tuned. Also the sampling time for optimal performance was investigated. The results show that the fuzzy control algorithm is not only better in performance than the conventional algorithm but also much easier to be tuned by operators in the field.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.929-930
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• 2005

• The Transactions of The Korean Institute of Electrical Engineers
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• v.63 no.12
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• pp.1637-1639
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• 2014

In the past, the analog protection relays have been widely used for the safety-related systems in the nuclear power plants due to their stability and reliability. Meanwhile, as the high performance digital system has been developed, the digital systems have been adopted in the non-safety systems. However, since the digital systems currently used in the non-safety systems were not developed according to Q-class standard, Commercial Grade Item Dedication (CGID) procedure should be performed in order to apply them to the safety-related system. The purpose of this paper is to describe the CGID procedure including the analysis of the hardware architecture as well as the software embedded in protective relay to apply to the emergency diesel generator in the nuclear power plant. The entire CGID procedure was performed strictly according to the international standard and regulations.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4422-4430
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• 2022

Studying the effects of gamma radiation on the instrumentation and control (I&C) system of a nuclear power plant is critical to the successful and reliable operation of the plant. In the accidental scenario, the adverse environment of ionizing radiation affects the performance of the I&C system and it leads to inaccurate and incomprehensible results. This paper reports the effects of gamma radiation on the AT89C51RD2, a commercial-off-the-shelf 8-bit high-performance flash microcontroller. The microcontroller, selected for the device under test for this study is used in the remote terminal unit for a nuclear power plant. The custom circuits were made to test the microcontroller under different gamma doses using a ⁶⁰Co gamma source in both ex-situ and in-situ modes. The device was exposed to a maximum dose of 1.5 kGy. Under this hostile environment, the performance of the microcontroller was studied in terms of device current and voltage changes. It was observed that the microcontroller device can operate up to a total absorbed dose of approximately 0.6 kGy without any failure or degradation in its performance.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.146-154
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• 2019

An improved principal component analysis (PCA) method is applied for sensor fault detection and isolation (FDI) in a nuclear power plant (NPP) in this paper. Data pre-processing and false alarm reducing methods are combined with general PCA method to improve the model performance in practice. In data pre-processing, singular points and random fluctuations in the original data are eliminated with various techniques respectively. In fault detecting, a statistics-based method is proposed to reduce the false alarms of $T^2$ and Q statistics. Finally, the effects of the proposed data pre-processing and false alarm reducing techniques are evaluated with sensor measurements from a real NPP. They are proved to be greatly beneficial to the improvement on the reliability and stability of PCA model. Meanwhile various sensor faults are imposed to normal measurements to test the FDI ability of the PCA model. Simulation results show that the proposed PCA model presents favorable performance on the FDI of sensors no matter with major or small failures.

• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.463-463
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• 2001