• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1983-1989
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• 2020

The human reliability analysis is a method by which, in general terms, the human impact to the safety and risk of a nuclear power plant operation can be modelled, quantified and analysed. It is an indispensable element of the PSA process within the nuclear industry nowadays. The paper herein presents a sensitivity study of the human reliability analysis performed on a real nuclear power plant-specific probabilistic safety assessment model. The analysis is performed on a pre-selected set of post-initiator operator actions. The purpose of the study is to investigate the impact of these operator actions on the plant risk by altering their corresponding human error probabilities in a wide spectrum. The results direct the fact that the future effort should be focused on maintaining the current human reliability level, i.e. not letting it worsen, rather than improving it.

• International Journal of Reliability and Applications
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• 제2권1호
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• pp.37-48
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• 2001

This paper presents a new dynamic human reliability analysis method and its application for quantifying the human error probabilities in implementing management action. For comparisons of current HRA methods with the new method, the characteristics of THERP, HCR, and SLIM-MAUD, which are most frequency used method in PSAs, are discussed. The action associated with implementation of the cavity flooding during a station blackout sequence is considered for its application. This method is based on the concepts of the quantified correlation between the performance requirement and performance achievement. The MAAP 3.0B code and Latin Hypercube sampling technique are used to determine the uncertainty of the performance achievement parameter. Meanwhile, the value of the performance requirement parameter is obtained from interviews. Based on these stochastic obtained, human error probabilities are calculated with respect to the various means and variances of the things. It is shown that this method is very flexible in that it can be applied to any kind of the operator actions, including the actions associated with the implementation of accident management strategies.

• Nuclear Engineering and Technology
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• 제29권1호
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• pp.1-6
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• 1997

This paper presents a new dynamic approach for assessing feasibility associated with the implementation of accident management strategies by the operators. This approach includes the combined use of both the concept of reliability physics and a dynamic event tree generation scheme. The reliability physics is based on the concept of a comparison between two competing variables, i.e., the requirement and the achievement parameter, while the dynamic event tree generation scheme on the continuous generation of the possible event sequences at every branch point up to the desired solution. This approach is applied to a cavity flooding strategy in a reference plant, which is to supply water into the reactor cavity using emergency fire systems in the station blackout sequence. The MAAP code and Latin Hypercube sampling technique are used to determine the uncertainty of the requirement parameter. It has been demonstrated that this combined methodology may contribute to assessing the success likelihood of the operator actions required during accidents and therefore to developing the accident management procedures.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2453-2466
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• 2022

Nuclear power plant operators in the main control room are exposed to stressful conditions in emergency situations as immediate and appropriate mitigations are required. While emergency operating procedures (EOPs) provide operators with the appropriate tasks and diagnostic guidelines, EOPs have static properties that make it difficult to reflect the dynamic changes of the plant. Due to this static nature, operator workloads increase because unrelated information must be screened out and numerous displays must be checked to obtain the plant status. Generally, excessive workloads should be reduced because they can lead to human errors that may adversely affect nuclear power plant safety. This paper presents a framework for an operator support system that can substitute the initial responses of the EOPs, or in other words the immediate actions and diagnostic procedures, in the early stages of an emergency. The system assists operators in emergency operations as follows: performing the monitoring tasks in parallel, identifying current risk and latent risk causality, diagnosing the accident, and displaying all information intuitively with a master logic diagram. The risk causalities are analyzed with a functional modeling methodology called multilevel flow modeling. This system is expected to reduce workloads and the time for performing initial emergency response procedures.

• 한국안전학회지
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• 제17권3호
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• pp.123-130
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• 2002

본 논문은 사고관리방안 수행에 있어서 발생되는 인적오류의 정량적 평가방법을 개발하여 공동범람 사고관리방안의 예제문제에 적용한 연구결과를 기술하고있다. PSA에서 사용되었던 기존의 인간오류평가 방법론인 THERP, HCR, SLIM-MAUD 방법의 특징을 검토하여 장단점을 기술하였다. 본 연구에서 제시하는 인간오류평가 방법론은 신뢰도물리모델을 이용하는 새로운 HRA 분석방법이다. 불확실성 분석을 위하여 MAAP 코드와 LHS 코드가 사용되었다. 본 연구를 통하여 제안하는 방법은 매우 유연하여 중대사고관리방안과 관련한 다양한 인간오류행위에 대한 평가에 사용될 수 있음을 보여주었다.

• Nuclear Engineering and Technology
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• 제44권8호
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• pp.911-918
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• 2012

Steam generator level control at low power is difficult due to its adverse thermal hydraulic properties, and is usually conducted by an operator. The basic model predictive control (MPC) is similar to the action of an operator in that the operator knows the desired reference trajectory for a finite period of time and takes the necessary control actions needed to ensure the desired trajectory. An MPC is based on a model; the performance as well as the efficiency of the MPC depends heavily on the exactness of the model. In this study, steam generator models that can describe in detail its thermal hydraulic behaviors, particularly at low power, are used in the MPC design. The design scope is divided into two parts. First, the MPC feedwater controller of the feedwater station is determined, and then the MPC level controller for the overall system is designed. Because the dynamic properties of a steam generator change with the power levels, a realistic situation is simulated by changing the transfer functions of the steam generator at every time step. The resulting MPC controller shows good performance.

• Nuclear Engineering and Technology
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• 제49권2호
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• pp.306-312
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• 2017

The purpose of this research is to introduce the technical standard of accident sequence precursor (ASP) analysis, and to propose a case study using the dynamic-probabilistic safety assessment (D-PSA) approach. The D-PSA approach can aid in the determination of high-risk/low-frequency accident scenarios from all potential scenarios. It can also be used to investigate the dynamic interaction between the physical state and the actions of the operator in an accident situation for risk quantification. This approach lends significant potential for safety analysis. Furthermore, the D-PSA approach provides a more realistic risk assessment by minimizing assumptions used in the conventional PSA model so-called the static-PSA model, which are relatively static in comparison. We performed risk quantification of a steam generator tube rupture (SGTR) accident using the dynamic event tree (DET) methodology, which is the most widely used methodology in D-PSA. The risk quantification results of D-PSA and S-PSA are compared and evaluated. Suggestions and recommendations for using D-PSA are described in order to provide a technical perspective.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.268-273
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• 2016

An innovative safety concept for a light water reactor has been developed at the Korea Atomic Energy Research Institute. It is a unique concept that adopts both a fast heat transfer mechanism for a small containment and a changing mechanism of the cooling geometry to take advantage of the potential, thermal, and dynamic energies of the cold water in the containment. It can bring about rapid cooling of the containment and long-term cooling of the decay heat. By virtue of this innovative concept, nuclear fuel damage events can be prevented. The ultimate heat transfer mechanism contributes to minimization of the heat exchanger size and containment volume. A small containment can ensure the underground construction, which can use river or seawater as an ultimate heat sink. The changing mechanism of the cooling geometry simplifies several safety systems and unifies diverse functions. Simplicity of the present safety system does not require any operator actions during events or accidents. Therefore, the unique safety concept of PX can realize both economic competitiveness and inherent safety.

• Nuclear Engineering and Technology
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• 제27권3호
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• pp.292-300
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• 1995

이 논문은 새로운 동적 인간신뢰도 분석방법을 제시하였고, 사고관리 방안의 수행시 인간오류확률의 계산에 이 방법을 적용하였다. 기존의 다른 방법과 비교하기 위하여 PSA의 HRA수행시 가장 많이 사용되는 THERP, HCR, 및 SLIM-MAUD 방법론들의 특징을 논의하였다. 정전사고시 공동범람시키는 방안을 예제로 사용하였다. 이 방법은 Requirement와 Achievement의 연관개념에 기초하고 있다. Achievement 변수의 불확정성은 MAAP 전산코드와 Latin Hypercube Sampling 방법을 이용하여 결정하였고 Requirement 변수값은 운전원과의 면담을 통하여 얻었다. 이렇게 얻어진 변수들의 분포를 가지고 여러가지 시간값의 평균과 분산에 대하여 인간오류 확률값을 계산하였다. 이 방법은 매우 유연하여 사고관리 전략수행과 관련한 행위를 포함한 어떤 종류의 운전원 행위에도 적용가능 함을 보여주었다.

• 한국신뢰성학회:학술대회논문집
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• 한국신뢰성학회 2000년도 춘계학술대회 발표논문집
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• pp.239-246
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• 2000

이 논문은 중대사고가 발생하였을때 운전원에 의하여 사고관리 방안을 수행하는 경우 그 실현성(Feasibility)을, 평가하는데 사용할 수 있는 새로운 시간의존적 신뢰도 분석방법을 제시하였다. 이 방법은 성능요구 (Performance Requirement)와 성능성취 (Performance Achievement)의 상관관계의 개념을 이용하는 신뢰도물리(Reliability Physics)와 모든 시간의존적 사고경위를 도출하는 동적사건수목 생성방법에 기초하고있다. 신뢰도물리는 성능요구변수와 성능성취변수의 비교를 이용한 신뢰도분석방법인 반면 동적사건수목 생성방법은 바람직한 해를 얻을 때까지 모든 가능한 사고경위를 도출해 내는 방법이다. 이 방법론을 정전사고시 참조원전의 공동에 비상화재시스템을 이용하여 물을 공급하는 공동범람사고관리 방안에 적용시켰다. Latin Hypercube Sampling 방법은 성능요구변수의 불확실성을 평가하는데 사용되었다. 제시된 방법론은 사고시 필요한 운전원의 방안수행 성공가능성을 평가하는데 사용될 수 있을 뿐만 아니라 궁극적으로 사고관리 절차서 개발에 도움이 될 수 있음을 보여주었다.