• 한국안전학회지
• /
• 제14권2호
• /
• pp.178-191
• /
• 1999

More than twenty methods were suggested for Human Reliability Analysis (HRA) in the field of safety analysis for Nuclear Power Plants (NPPs). However, there is still a high uncertainty on the analysis and a difficulty in performing HRA. New methods and approaches are under studying to overcome such limitations of current HRA. This paper presents some results of study to analysis limitations of current HRA in viewpoint of user, i.e., HRA analyst. The limitation analysis was based on 89 human error events modeled in a Probabilistic Safety Assessment (PSA) project for NPPs in Korea. Total 17 specific limitations were identified and categorized into seven groups. Important analysis has also been undertaken to assess the order of priority among those limitations. Finally, seven requirements with priority ranking were generated for an advanced framework and methodology of HRA.

• 한국신뢰성학회:학술대회논문집
• /
• 한국신뢰성학회 2000년도 추계학술대회
• /
• pp.269-277
• /
• 2000

The component reliability database is required in PSA (Probabilistic Safety Analysis) for NPP (Nuclear Power Plant). We have applied a generic database to the PSA for the Korean NPPs, since there is no specific component reliability database. Therefore we are developing the plant-specific component reliability database for domestic NPPs. We also extend the experience and knowledge of PSA and component reliability database for NPP to chemical industry We collect the raw data like component operation history and maintenance history and then input the required data for the component reliability database through failure analysis. With the database, we can not only perform PSA with real data but also perform maintenance optimization.

• 대한안전경영과학회지
• /
• 제24권1호
• /
• pp.61-72
• /
• 2022

This study aims to propose a conceptual design of information displays for supporting responsive actions under severe accidents in Nuclear Power Plants (NPPs). Severe accidents in NPPs can be defined as accident conditions that are more severe than a design basis accident and involving significant core degradation. Since the Fukushima accident in 2011, the management of severe accidents is increasing important in nuclear industry. Dealing with severe accidents involves several cognitively complex activities, such as situation assessment; accordingly, it is significant to provide human operators with appropriate knowledge support in their cognitive activities. Currently, severe accident management guidelines (SAMG) have been developed for this purpose. However, it is also inevitable to develop information displays for supporting the management of severe accidents, with which human operators can monitor, control, and diagnose the states of NPPs under severe accident situations. It has been reported that Ecological Interface Design (EID) framework can be a viable approach for developing information displays used in complex socio-technical systems such as NPPs. Considering the design principles underlying the EID, we can say that EID-based information displays can be useful for dealing with severe accidents effectively. This study developed a conceptual design of information displays to be used in severe accidents, following the stipulated design process and principles of the EID framework. We particularly attempted to develop a conceptual design to make visible the principle knowledge to be used for coping with dynamically changing situations of NPPs under severe accidents.

• 자원ㆍ환경경제연구
• /
• 제27권2호
• /
• pp.261-286
• /
• 2018

본 논문은 제 7차 전력수급기본계획에서 제시한 신규 원자력발전, 석탄 발전, 그리고 LNG 복합 발전의 균등화발전비용과, 고리 1호기(가압형 경수로, PWR) 및 월성 1호기(가압형 중수로, PHWR)의 계속운전 기간별(10년과 20년) 균등화발전비용을 추정하여 비교해서 원전 계속운전의 노형별 및 계속운전 기간별 경제성을 평가하였다. 균등화발전비용을 이용한 원자력 발전의 계속운전 경제성은 노형, 계속운전기간, 할인율, 이용률 등으로부터 영향을 받는다. 분석결과에 따르면 가압형 경수로(고리 1호)는 가압형 중수로(월성 1호)보다 경제성이 높다. 원자력발전의 계속운전과 다른 전원의 경제성 비교 결과를 보면 가압형 경수로(고리 1호)의 경우에 20년 계속운전이 신규 원자력 발전 및 석탄발전보다 경제적이다. 그러나 가압형 중수로(월성 1호)의 경우에 20년 계속운전은 LNG 복합 발전보다 경제적이지만, 신규 원전 및 신규 석탄발전보다 비경제적이다. 원자력발전의 계속운전에서 보면 20년 계속운전이 경제적이며, 특히 가압형 경수로는 다른 전원보다 비용효율적이다. 원자력발전의 계속운전 정책은 모든 원전을 폐로하기 보다는 안전성과 경제성을 동시에 고려하는 선별적 접근 방식이 유효하다.

• 한국안전학회지
• /
• 제35권5호
• /
• pp.121-127
• /
• 2020

Since the Fukushima nuclear accident in 2011, the public were concerned about the safety of Nuclear Power Plants (NPPs) in extreme natural disaster situations, such as earthquakes, flooding, heavy rain and tsunami, have been increasing around the world. Accordingly, the Stress Test was conducted in Europe, Japan, Russia, and other countries by reassessing the safety and response capabilities of NPPs in extreme natural disaster situations that exceed the design basis. The extreme natural disaster can put the NPPs in beyond-design-basis conditions such as the loss of the power system and the ultimate heat sink. The behaviors and capabilities of NPPs with losing their essential safety functions should be measured to find and supplement weak areas in hardware, procedures and coping strategies. The Loss of Ultimate Heat Sink (LUHS) accident assumes impairment of the essential service water system accompanying the failure of the component cooling water system. In such conditions, residual heat removal and cooling of safety-relevant components are not possible for a long period of time. It is therefore very important to establish coping strategies considering all available equipment to mitigate the consequence of the LUHS accident and keep the NPPs safe. In this study, thermal hydraulic behavior of the LUHS event was analyzed using RELAP5/Mod3.3 code. We also performed the sensitivity analysis to identify the effects of the operator recovery actions and operation strategy for charging pumps on the results of the LUHS accident.

• Nuclear Engineering and Technology
• /
• 제52권1호
• /
• pp.87-100
• /
• 2020

Performance shaping factors (PSFs) in a human reliability analysis (HRA) are one that may influence human performance in a task. Most currently applicable HRA methods for nuclear power plants (NPPs) use PSFs to highlight human error contributors and to adjust basic human error probabilities (HEPs) that assume nominal conditions of NPPs. Thus far, the effects of PSFs have been treated independently. However, many studies in the fields of psychology and human factors revealed that there may be relationships between PSFs. Therefore, the inter-relationships between PSFs need to be studied to better reflect their effects on operator errors. This study investigates these inter-relationships using two data sources and also suggests a context-based approach to treat the inter-relationships between PSFs. Correlation and factor analyses are performed to investigate the relationship between PSFs. The data sources are event reports of unexpected reactor trips in Korea and an experiment conducted in a simulator featuring a digital control room. Thereafter, context-based approaches based on the result of factor analysis are suggested and the feasibility of the grouped PSFs being treated as a new factor to estimate HEPs is examined using the experimental data.

• 한국안전학회지
• /
• 제14권1호
• /
• pp.177-184
• /
• 1999

PSA(Probabilistic Safety Assessment) methodology is widely used on assessing the safety of Nuclear Power Plants(NPPs) quantitatively in the domestic nuclear field. Initiating event frequencies are absolutely needed to conduct PSA, and they considerably affect PSA results. There is no domestic database where domestic trip event cases are reflected, so they are used to assess the safety of NPPs that are from the foreign database. In this paper, operating experience data from the Korean NPPs was collected and analyzed for the trip event cases, which are necessary to determine the initiating events and their frequencies. Korean NPPs have experienced five of 16 initiating events, which we LOFW. LOCV, LOCCW, LOOP and GTRN as a result of analyzing the trip event cases. Initiating frequencies based on the domestic trip event cases are analyzed, and they are similar to that from the foreign database.

• 한국압력기기공학회 논문집
• /
• 제7권3호
• /
• pp.23-29
• /
• 2011

The renewed global interest in nuclear power has arisen from the need to reduce greenhouse gas emissions and to provide sufficient electricity for a growing global population before the accident at Fukushima Dai-ichi nuclear power plant in Japan. In spite of the safety issues of nuclear power plants raised by the ongoing Japanese nuclear crisis, many countries with nuclear power plants (NPPs) are still implementing license extensions of 10~20 years, and even consideration is being given to the concept of life-beyond-60, a further period of license extension from 60 to 80 years. To solving the materials aging problem is integral to its success. To evaluate the plant aging phenomena, a lot of background information such as materials and environment of the parts of the reactor and plant systems is needed by the experts. Information on degradation mechanisms is also used. In this paper, a materials degradation evaluation program called OnMDE-SYS (On-line Materials Degradation Evaluation System) is introduced. The developed program provides a variety of information on the materials and stressors as well as operational experience to the experts. It is also anticipated that the experts can perform materials degradation assessment on the web directly by referring to domestic and international information about the degradation of a nuclear power plants through OnMDE-SYS.

• Journal of Welding and Joining
• /
• 제29권2호
• /
• pp.34-39
• /
• 2011

Nickel-based austenitic alloys such as Alloy 82 and 182 had been employed as the weld metals in nuclear power plants (NPPs) due to their high corrosion resistance as well as good mechanical properties. However, since the 2000s, the occurrence of primary water stress corrosion cracking has been reported in conjunction with these alloys in domestic and oversea NPPs. In the present work, we assumed an imaginary crack at the inner surface of a surge nozzle weld that had previously experienced the outside repair welding, and constructed its finite element model. Finite element analysis was performed with respect to the heat transfer, and then to the residual stress for obtaining the total applied stress distributions. These stress distributions were finally converted to the stress intensity factors for estimating crack growth rate. From the comparison of crack growth rate curves for the cases of no repair welding and outside repair welding, it was found that the outside repair welding did not exhibit negative effect on the crack growth for the surge nozzle under consideration in this work; in both cases, the cracks stopped growing before they became the through-wall cracks.

• 대한인간공학회지
• /
• 제32권1호
• /
• pp.27-36
• /
• 2013

Objective: This paper describes an analysis of electroencephalography(EEG) signals to identify human errors during using digital devices in nuclear power plants(NPPs). Background: The application of an advanced main control room(MCR) has accompanied with lots of changes in different forms and features by virtue of new digital technologies. The characteristics of these digital technologies and devices provide several opportunities for the use of interface management. It can integrate into a compact single workstation in an advanced MCR, allowing workers to operate the plant with minimum physical burden under any operating condition. However these devices may introduce new types of human errors, and thus we need a means to assess and prevent such errors especially those related to digital devices. Method/Conclusion: The EEG data are relatively objective, and thus we introduce several measures to EEG analysis for obtaining the feasibility of human error identification. Application: This study may support to ensure the safety when applying digital devices in NPPs.