• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.627-638
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• 2018

This article describes an integrated Level 1-Level 2 probabilistic risk assessment (PRA) methodology to evaluate the radiological risk during postulated accident scenarios initiated during the decommissioning phase of a typical Mark I containment boiling water reactor. The fuel damage scenarios include those initiated while the reactor is permanently shut down, defueled, and the spent fuel is located into the spent fuel storage pool. This article focuses on the integrated Level 1-Level 2 PRA aspects of the analysis, from the beginning of the accident to the radiological release into the environment. The integrated Level 1-Level 2 decommissioning PRA uses event trees and fault trees that assess the accident progression until and after fuel damage. Detailed deterministic severe accident analyses are performed to support the fault tree/event tree development and to provide source term information for the various pieces of the Level 1-Level 2 model. Source terms information is collected from accidents occurring in both the reactor pressure vessel and the spent fuel pool, including simultaneous accidents. The Level 1-Level 2 PRA model evaluates the temporal and physical changes in plant conditions including consideration of major uncertainties. The goal of this article is to provide a methodology framework to perform a decommissioning Probabilistic Risk Assessment (PRA), and an application to a real case study is provided to show the use of the methodology. Results will be derived from the integrated Level 1-Level 2 decommissioning PSA event tree in terms of fuel damage frequency, large release frequency, and large early release frequency, including uncertainties.

• Nuclear Engineering and Technology
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• v.43 no.3
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• pp.271-278
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• 2011

When performing fire PSA in a nuclear power plant, an event mapping method, using an internal event PSA model, is widely used to reduce the resources used by fire PSA model development. Feasible initiating events and component failure events due to fire are identified to transform the fault tree (FT) for an internal event PSA into one for a fire PSA using the event mapping method. A surrogate event or damage term method is used to condition the FT of the internal PSA. The surrogate event or the damage term plays the role of flagging whether the system/component in a fire compartment is damaged or not, depending on the fire being initiated from a specified compartment. These methods usually require explicit states of all compartments to be modeled in a fire area. Fire event scenarios, when using explicit identification, such as surrogate or damage terms, have two problems: (1) there is no consideration of multiple fire propagation beyond a single propagation to an adjacent compartment, and (2) there is no consideration of simultaneous fire propagations in which an initiating fire event is propagated to multiple paths simultaneously. The present paper suggests a fire propagation equation to identify all possible fire event scenarios for an explicitly treated fire event scenario in the fire PSA. Also, a method for separating fire events was developed to make all fire events a set of mutually exclusive events, which can facilitate arithmetic summation in fire risk quantification. A simple example is given to confirm the applicability of the present method for a $2{\times}3$ rectangular fire area. Also, a feasible asymptotic approach is discussed to reduce the computational burden for fire risk quantification.

• Journal of Korea Society of Digital Industry and Information Management
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• v.10 no.1
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• pp.39-45
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• 2014

This paper proposes an architecture of a sensor network for gathering data under a powerful beam cluster tree architecture. This architecture is used when there is a need to gather data from sensor node where there is no sink node connected to an existing network, or it is required to get a series of data specific to an event or time. The transmit distance of the beam signal is longer than that of the usual sensor node. The nodes of the network make a tree network when receiving a beam message transmitting from the powerful root node. All sensor nodes in a sink tree network synchronize to the superframe and know exactly the sequence value of the current superframe. When there is data to send to the sink node, the sensor node sends data at the corresponding allocated channel. Data sending schemes under the guaranteed time slot are tested and the delay and jitter performance is explained.

• Journal of the Korean Society of Safety
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• v.32 no.2
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• pp.27-33
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• 2017

This study is intended to identify issues on the basis of investigating the actual state of laboratory environment and outlet circuit, and derive end states by expressing sequences from the initiating event of disaster to accident in leakage current, poor contact and overload through ETA(event tree analysis). To this end, this study investigated the actual state of electric equipment of laboratory at universities in all parts of country. And it is shown that most of them are failure in electric work and user negligence in the investigation of actual state. It is found that there is earth fault and defect in wire diameter in the failure of electric work and the problem of partial disconnection due to wire bundling and poor contact in user negligence. Outlet-related component, failure rate and initiating events are composed of a total of 41 initiating events, i.e., 30 internal initiating events and 11 external initiating events. And end states are composed of a total of 15 parts, i.e., 3 electric power parts and 12 safety parts. Earthing class 3 is the most important safety device against leakage current (initiating event). And in case of poor contact, it is necessary for manager to check thoroughly because there is no safety device. In case of overload/overcurrent, when high-capacity equipment is connected, a molded case circuit breaker, safety device, worked. However, in most cases, it is verified that this doesn't work. This study can be utilized as electric equipment safety guide for laboratory safety manager and managers.

• Journal of Korea Multimedia Society
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• v.9 no.7
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• pp.927-937
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• 2006

Auto-playing programs are often used in behalf of human players in MMORPG(Massively Multi-player Online Role Playing Game). By playing automatically and continuously, it helps to speed up the game character's level-up process. However, the auto-playing programs, either software or hardware, do harm to games servers in various ways including abuse of resources. In this paper, we propose a way of detecting the auto programs by analyzing the window event sequences produced by the game players. In our proposed method, the event sequences are transformed into a set of attributes, and the Decision Tree learning is applied to classify the data represented by the set of attribute values into human or auto player. The results from experiments with several MMORPG show that the Decision Tree learning with proposed method can identify the auto-playing programs with high accuracy.

• Fire Science and Engineering
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• v.8 no.2
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• pp.58-63
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• 1994

독일 원자력 발전소에서는 포괄적인 방화 연구의 한 부분으로써 방화에 관한 분석과 그것을 최적화 할 수 있는 확률론적 방법을 개발하였다. 그 일반적인 흐름을 살펴보면, 미국의 화재 위험성 분석의 방법을 따랐으며, 세밀한 부분에서는 약간의 수정을 한 것이다. 먼저, 선정된 공장지역에서의 화재 사건 경로(fire event tree)는 화재가 발생했을 때, 방화 조치와 안전시스템을 능 수동적으로 고려해서 설정된다. 방화 조치와 안전 시스템에 있어서의 실패 모델(failure model)은 발화 후 시간과 화재 영향과 같은 일상적인 변수와 관련해서 생긴다. 이러한 관련성은 일차(first-order) 시스템의 신뢰성 이론을 적절히 이용해서 화재 사건 경로를 분석할 때 알 수 있다. 더불어 화재가 발생했을 떠 방화 시스템의 실패 빈도, event paths의 상대적인 비중, 이러한 path내에서의 방화 조치 그리고 실패모델의 변수 등은 모두 시간 함수로 계산된다. 이러한 자료에 근거를 두고, 방화의 최적화는 주로 event path, 방화조치와 비중이 가장 큰 변수를 수정함으로써 가능하게 된다. 이것은 독일의 1300 MW PWR reference plant를 예를 들어서 증명될 것이다. 또한 충고를 받아들여서 수정을 하는 것은 발전소 직원과 화재 피해의 위험성을 줄일 수 있다는 것을 보여주고 있다.

• Nuclear Engineering and Technology
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• v.40 no.7
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• pp.571-580
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• 2008

A Binary Decision Diagram (BDD) is a graph-based data structure that calculates an exact top event probability (TEP). It has been a very difficult task to develop an efficient BDD algorithm that can solve a large problem since it is highly memory consuming. In order to solve a large reliability problem within limited computational resources, many attempts have been made, such as static and dynamic variable ordering schemes, to minimize BDD size. Additional effort was the development of a ZBDD (Zero-suppressed BDD) algorithm to calculate an approximate TEP. The present method is the first successful application of a BDD truncation. The new method is an efficient method to maintain a small BDD size by a BDD truncation during a BDD calculation. The benchmark tests demonstrate the efficiency of the developed method. The TEP rapidly converges to an exact value according to a lowered truncation limit.

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.386-403
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• 2016

The safety of nuclear power plants is analyzed by a probabilistic risk assessment, and the fault tree analysis is the most widely used method for a risk assessment with the event tree analysis. One of the well-known disadvantages of the fault tree is that drawing a fault tree for a complex system is a very cumbersome task. Thus, several graphical modeling methods have been proposed for the convenient and intuitive modeling of complex systems. In this paper, the reliability graph with general gates (RGGG) method, one of the intuitive graphical modeling methods based on Bayesian networks, is improved for the reliability analyses of dynamic systems that have various operation modes with time. A reliability matrix is proposed and it is explained how to utilize the reliability matrix in the RGGG for various cases of operation mode changes. The proposed RGGG with a reliability matrix provides a convenient and intuitive modeling of various operation modes of complex systems, and can also be utilized with dynamic nodes that analyze the failure sequences of subcomponents. The combinatorial use of a reliability matrix with dynamic nodes is illustrated through an application to a shutdown cooling system in a nuclear power plant.

• Proceedings of the KIEE Conference
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• 2005.05a
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• pp.81-83
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• 2005

The Reactor Protection System (RPS) is a very important system in a nuclear power plant because the system shuts down the reactor to maintain the reactor core integrity and the reactor coolant system pressure boundary if the plant conditions approach the specified safety limits. This paper describes the unavailability assessment of a digital reactor protection system using the fault tree analysis technique. The fault tree technique can be expressed in terms of combinations of the basic event failures. In this paper, a prediction method of the hardware failure rate is suggested for a digital reactor protection system. and applied to the reactor protection system being developed in Korea.

• International Journal of Control, Automation, and Systems
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• v.4 no.1
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• pp.105-112
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• 2006

The main function of a reactor protection system is to maintain the reactor core integrity and the reactor coolant system pressure boundary. Generally, the reactor protection system adopts the 2-out-of-m redundant architecture to assure a reliable operation. This paper describes the safety assessment of a digital reactor protection system using the fault tree analysis technique. The fault tree technique can be expressed in terms of combinations of the basic event failures such as the random hardware failures, common cause failures, operator errors, and the fault tolerance mechanisms implemented in the reactor protection system. In this paper, a prediction method of the hardware failure rate is suggested for a digital reactor protection system, and applied to the reactor protection system being developed in Korea to identify design weak points from a safety point of view.