• 한국경영과학회지
• /
• 제30권2호
• /
• pp.63-71
• /
• 2005

In the field of nuclear reactor safety study, common cause failures (CCFs) became significant contributors to system failure probability and core damage frequency in most Probabilistic risk assessments. However, it is hard to estimate the reliability of such a system, because of the dependency of components caused by CCFs. In order to analyze the system, we propose an analytic method that can find the parameters with lack of raw data. This study adopts the shock model in which the failure probability increases as the shock is cumulated. We use two-step Expectation and Maximization (EM) algorithm to find the unknown parameters. In order to verify the analysis result, we perform the simulation under same environment. This approach might be helpful to build the defensive strategy for the CCFs.

• 산업공학
• /
• 제25권4호
• /
• pp.405-415
• /
• 2012

The reliability performance measures for low and high or continuous demand modes of operation of safety instrumented systems(SISs) are examined and compared by analyzing the official definitions in IEC 61508 standard. This paper also presents a status of common cause factor(CCF) models used in IEC 61508 and problems relating CCF modelling are discussed and ideas to solve these ones are suggested. An example with mixed M-out-of-N architecture is carried out to illustrate the proposed methods.

• Nuclear Engineering and Technology
• /
• 제45권6호
• /
• pp.795-802
• /
• 2013

For the improvement of APR1400 Diverse Protection System (DPS) design, the Advanced DPS (ADPS) has recently been developed to enhance the fault tolerance capability of the system. Major fault masking features of the ADPS compared with the APR1400 DPS are the changes to the channel configuration and reactor trip actuation equipment. To minimize the fault occurrences within the ADPS, and to mitigate the consequences of common-cause failures (CCF) within the safety I&C systems, several fault avoidance design features have been applied in the ADPS. The fault avoidance design features include the changes to the system software classification, communication methods, equipment platform, MMI equipment, etc. In addition, the fault detection, location, containment, and recovery processes have been incorporated in the ADPS design. Therefore, it is expected that the ADPS can provide an enhanced fault tolerance capability against the possible faults within the system and its input/output equipment, and the CCF of safety systems.

• International Journal of Control, Automation, and Systems
• /
• 제4권1호
• /
• pp.105-112
• /
• 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.

• 품질경영학회지
• /
• 제39권2호
• /
• pp.320-328
• /
• 2011

This research proposes comprehensive models for analyzing common cause failures (CCF) due to cumulative shocks and to assess system reliability under the CCF. The proposed cumulative shock models are based on the binomial failure rate (BFR) model. Six kinds of models are proposed so as to explain diverse cumulative shock phenomena. The models are composed of the initial failure probability, shape parameter, and the total shock number. Some parameters of the proposed models can not be explicitly estimated, so we adopt the Expectation-maximization (EM) algorithm in order to obtain the maximum likelihood estimator (MLE) for the parameters. By estimating the parameters for the cumulative shock models, the system reliability with CCF can be assessed sequentially according to the number of cumulative shocks. The result can be utilizes in dynamic probabilistic safety assessment (PSA), aging studies, or risk management for nuclear power plants. Replacement or maintenance policies can also be developed based on the proposed model.

• 한국항공우주학회지
• /
• 제50권4호
• /
• pp.259-267
• /
• 2022

기존의 국내 개발 회전익 항공기 시스템의 고장-안전 설계 요구사항 도출 방법은 최신 통합형 항공전자 시스템에 적용 시 단일 항목의 고장으로 인하여 치명적인 시스템 기능 고장을 발생시키는 요인을 누락할 수 있다. 그 원인은 고장-안전 설계 대상을 선정함에 있어 단일 품목의 체계 기능 고장 영향성을 그 기준으로 함에 있다. 본 연구에서는 이를 해결하기 위하여 민수 항공기 개발 국제 표준인 SAE ARP4754A의 기능적 위험요소 평가 및 개발보증수준 할당 절차를 활용하여, 시스템 구조의 고장-안전 설계 요구사항을 도출하기 위한 체계적인 분석 절차를 제시한다. 또한 본 연구에서 제시한 절차가 앞서 제시한 문제점을 해결할 수 있는지를 확인하기 위하여 치명적인 기능 고장을 발생시킬 수 있는 단일 요인을 내포한 시스템 구조를 가정하여 교차 검증을 수행하였다. 그 결과 기존 연구 방법으로는 누락되었던 치명적인 공통 요인을 식별할 수 있었고 이를 통제하기 위한 고장-안전 설계 요구사항이 도출됨을 확인하였다.

• 한국항공운항학회지
• /
• 제4권1호
• /
• pp.7-24
• /
• 1996

A methodology has been developed for predicting aircraft reliability incorporating both C.C.F.s(Common-Cause Failures), and phased missions. Failure behaviour of an aircraft, or it's systems are predicted. Both independent failures, and C.C.F.s, are modelled by the Markov process, and simulated using Monte Carlo sampling with the robust variance reduction method. Prediction of safety and reliability is made through discrete-event simulation of aircraft operations. A case study is described for investigating the safety and reliability of the propulsion system of two-, three- and four-engined aircraft. This is particularly important for the design of ETOPS(Extended Range of Two-Engined Aircraft Operations) and results are presented for the cases with, and without the effect of C.C.F.s.

• 한국화재소방학회논문지
• /
• 제17권4호
• /
• pp.76-85
• /
• 2003

본 논문에서는 공통원인고장의 이론적 고찰로서 정의와 그 원인, 분석방법을 기술하고 대표적인 소방설비의 하나인 소방펌프에 대해 공통원인고장을 고려한 신뢰도분석에 적용함으로써 공통원인고장의 중요성과 그 한계성을 규명하고자 한다. 공통원인고장을 고려한 소방수 공급설비에 대한 신뢰도분석 결과 알 수 있듯이 펌프의 운전고장이 계통실패의 가장 큰 요인으로 나타났으며 특히 두 펌프의 공통원인고장이 지배적이다. 다시 말하면 공통원인고장을 고려하지 않을 경우에 계통신뢰도를 실제보다 2배 이상 초과하여 평가할 수 있다는 것이다. 이로서 계통 신뢰도분석에서 공통원인고장의 중요성을 인식할 수 있으며 분석결과는 공통원인고장의 변수인자의 값에 크게 의존하는 것을 알 수 있다. 그리고 소방수 공급설비설계에 계통설계 시 다중성을 반영하면 신뢰도가 증가하는 것은 사실이나 공통원인고장 요인 때문에 다중기기 설치대수에 비례하는 정도의 신뢰도 향상을 얻지 못할 수도 있다. 또한 공통원인고장의 한계성으로는 분석모델의 차이로 인한 차이는 미미한 수준이었으나 각각 다른 데이터 원을 사용했을 경우 그 결과는 큰 차이를 나타내었다. 따라서 공통원인고장 분석에 사용되는 모델보다는 이용 가능한 경험데이터의 품질이 그 분석결과의 신뢰성에 큰 영향을 미친다는 것을 알 수 있었다. 결과적으로 다중기기의 공통원인고장을 방지하기 위한 기본적이고 공학적인 방안으로는 설계시 요구되는 적정 신뢰도를 유지하는 것이므로 적어도 소방펌프에 요구되는 신뢰도수준으로 설계되어야 한다. 즉 SIS(Safety Instrumented system)에 요구되는 신뢰도수준인 안전건전성수준(SIL; Safety integrity level)에 적합한가의 유무를 PFD를 활용하여 정량적으로 파악하는 것이다. 공통원인고장을 고려한 소방수 공급설비에 대한 신뢰도분석 결과 계통작동요구시 실패확률(PFD: Probability of failure on demand), 즉 계통 이용 불능도는 3.80E-3이므로 규정목표인 SIL5의 범주 안에 들어있지 않아 안전건전성수준으로 설계되어 있지 않다고 판단되며, 만일 공통원인고장을 고려하지 않았을 경우인 계통 이용불능도 또한 1.82E-3으로 계산되는데, 이 또한 SIL5의 범주 안에 들어있지 않으므로 단전건전성수준으로 설계되어 있지 않다고 판단된다.

• Nuclear Engineering and Technology
• /
• 제44권5호
• /
• pp.471-482
• /
• 2012

To assess the risk of nuclear power plant operation and to determine the risk impact of digital systems, there is a need to quantitatively assess the reliability of the digital systems in a justifiable manner. The Probabilistic Risk Analysis (PRA) is a tool which can reveal shortcomings of the NPP design in general and PRA analysts have not had sufficient guiding principles in modelling particular digital components malfunctions. Currently digital I&C systems are mostly analyzed simply and conventionally in PRA, based on failure mode and effects analysis and fault tree modelling. More dynamic approaches are still in the trial stage and can be difficult to apply in full scale PRA-models. As basic events CPU failures, application software failures and common cause failures (CCF) between identical components are modelled.The primary goal is to model dependencies. However, it is not clear which failure modes or system parts CCF:s should be postulated for. A clear distinction can be made between the treatment of protection and control systems. There is a general consensus that protection systems shall be included in PRA, while control systems can be treated in a limited manner. OECD/NEA CSNI Working Group on Risk Assessment (WGRisk) has set up a task group, called DIGREL, to develop taxonomy of failure modes of digital components for the purposes of PRA. The taxonomy is aimed to be the basis of future modelling and quantification efforts. It will also help to define a structure for data collection and to review PRA studies.

• Nuclear Engineering and Technology
• /
• 제53권1호
• /
• pp.103-110
• /
• 2021

In the development of a Risk Monitor probabilistic safety assessment (PSA) model from the basic PSA model of a nuclear power plant, the modeling of common-cause failure (CCF) is very important. At present, some approximate modeling methods are widely used, but there lacks criterion of modeling accuracy and error analysis. In this paper, aiming at ensuring the accuracy of risk assessment and minimizing the Risk Monitor PSA models size, we present three basic issues of CCF model resulted from the changes of a nuclear power plant configuration, put forward corresponding modeling methods, and derive accuracy criteria of CCF modeling based on minimum cut sets and risk indicators according to the requirements of risk monitoring. Finally, a nuclear power plant Risk Monitor PSA model is taken as an example to demonstrate the effectiveness of the proposed modeling method and accuracy criteria, and the application scope of the idea of this paper is also discussed.