• 한국안전학회지
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• 제33권6호
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• pp.136-143
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• 2018

In this study, we conducted a pilot study on the multiple spurious operations (MSO) analysis in the fire probabilistic safety assessment (PSA) of domestic nuclear power plant (NPP) to identify the degree of influence of the operator actions used in the MSO mitigation strategies. The MSO scenario of the domestic reference NPP selected for this study is refueling water tank (RWT) drain down event. It could be caused by spurious operations of the containment spray system (CSS) of the reference NPP. The RWT drain down event can be stopped by the main control room (MCR) operator actions for stopping the operation of CSS pump or closing the CSS motor operated valve if the containment spray actuation signal (CSAS) is spuriously actuated. Outside the MCR, it can be stopped by operator actions for closing the CSS manual valves or motor operated valve or stopping the operation of CSS pump. The quantification result of a fire PSA model that takes into account all recovery actions for the RWT drain down event lead to risk reduction by about 95%, compared with quantification result of fire PSA model without considering them. Among the various operator actions, the recovery action for the spurious CSAS operations and the operator action for the manual valve are identified as the most important operator actions. This study quantitatively showed the extent to which the operator actions used as MSO countermeasures have affected the fire PSA quantification results. In addition, we can see the rank of importance among the operator recovery actions in quantitative terms.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 춘계학술논문요약집
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• pp.29-30
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• 2018

• Nuclear Engineering and Technology
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• 제15권4호
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• pp.219-228
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• 1983

릴리프 밸브의 차단까지 TMI-2 사고의 blowdown history를 검토하고 TMI-2 사고와 같은 소규모의 냉각제 상실 사고 동안 노심 파괴를 막기 위해 더 가산해야할 측정 기구에 대하여 논의하였다. 가산된 기구를 이용하여 어떻게 노심의 uncovered level과 operator time margin을 계산하는 가를 검토하였으며, TMI-2 사고에 대해 uncovered level과 operator time margin을 결정하기 위한 샘플 계산을 수행하였다. 이 방법을 이용해서 측정되는 변수들의 함수로써 uncovered level과 operator time margin을 보여주는 도표를 작성하였다.

• Nuclear Engineering and Technology
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• 제50권4호
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• pp.562-569
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• 2018

Recently, human errors have very rarely occurred during power generation at nuclear power plants. For this reason, many countries are conducting research on smart support systems of nuclear power plants. Smart support systems can help with operator decisions in severe accident occurrences. In this study, a smart support system was developed by integrating accident prediction functions from previous research and enhancing their prediction capability. Through this system, operators can predict accident scenarios, accident locations, and accident information in advance. In addition, it is possible to decide on the integrity of instruments and predict the life of instruments. The data were obtained using Modular Accident Analysis Program code to simulate severe accident scenarios for the Optimized Power Reactor 1000. The prediction of the accident scenario, accident location, and accident information was conducted using artificial intelligence methods.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.1261-1262
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• 2005

• Nuclear Engineering and Technology
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• 제39권2호
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• pp.123-128
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• 2007

In order to apply a static fault-tree (FT) method to a system or a plant whose configuration changes dynamically, condition gates and a post processing method are used to effectively accommodate these changes. An operator's performance change, which can be caused by these configuration changes, should also be considered to assess the risk to a plant in a more realistic manner. This study aims to develop an integrated framework to accommodate various configuration changes and their effect on an operator’s performance by using the FT model. We applied a condition-based human reliability assessment (CBHRA) method to consider various conditions endured by an operator. That is, we integrated the CBHRA method with the conventional post processing method for modeling the system configuration changes. The effect of the condition monitoring systems installed in a plant is also considered. In this study, we show an example application of the integrated framework to a probabilistic safety assessment for the shutdown phase of a nuclear power plant.

• 대한인간공학회지
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• 제32권1호
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• pp.1-7
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• 2013

Objective: This study proposes the application of Ecological Interface Design(EID) method that is effective for situation awareness in digitalized environment. Background: While cognitive interface design method such as Information Rich Display(IRD) is simply focused on existing information for user, EID method helps users' resource to be solved to higher ion task such as diagnostic and problem solving. Method: Using EID method based on Work Domain Analysis (WDA), it was analyzed and designed for Steam Generator(SG) Water Level control process in a digitalized Main Control Room of Nuclear Power Plant. Proposed EID example is evaluated through interviews by expert & operator. Results: The result of expert & operator showed that EID display might give an aid for operator's decision. Conclusion: The results can reduce critical accidental damage that occurred due to cognitive load and so critical human error. Application: This study may be impact on situation awareness study for digitalized interface design.

• Nuclear Engineering and Technology
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• 제34권4호
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• pp.358-369
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• 2002

Abundant information about analyses of single steam generator tube rupture (SGTR) events is available because of its importance in terms of safety. However, there are few literatures available on analyses of multiple steam generator tube rupture (MSGTR) events. In addition, knowledge of transients and consequences following a MSGTR event are very limited as there has been no occurrence of MSGTR event in the commercial operation of nuclear reactors. In this study, a postulated MSGTR event in an APR1400 is analyzed using thermal-hydraulic system code MARSI.4. The present study aims to examine the effects of affected steam generator selection. The main steam safety valve (MSSV) lift time for four cases are compared in order to examine how long operator response time is allowed depending on which steam generate. (S/G) is affected. The comparison shows that the cases where two steam generators are simultaneously affected allow longer time for operator action compared with the cases where a single steam generator is affected. Furthermore, the tube ruptures in the steam generator where a pressurizer is connected leads to the shortest operator response time.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.185-190
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• 1995

Conventional alarm system has many difficulties in the operator's identifying the plant status during special situations such as design basis accidents. To solve the shortcomings, an on-line alarm annunciator system, called dynamic alarm console (DAC), was developed. In the DAC, a signal is generated as alarm by the use of an adaptive setpoint check strategy based on operating mode, and time delay technique is used not to generate nuisance alarms. After alarm generation, if activated alarm is a level precursor alarm or a consequencial alarm, it would be suppressed, and the residual alarms go through dynamic prioritization which provide the alarms with pertinent priorities to the current operating mode. Dynamic prioritization is achieved by going through the system- and mode-oriented prioritization. The DAC has the alarm hierarchical structure based on the physical and functional importance of alarms. Therefore the operator can perceive alarm impacts on the safety or performance of the plant with the alarm propagation from equipment level to plant functional level. In order to provide the operator with the most possible cause of the event and quick cognition of the plant status even without recognizing the individual alarms, reactor trip status tree (RTST) was developed. The DAC and the RTST have been simulated with on-line data obtained from the full-scope simulator for several abnormal cases. The results indicated that the system can provide the operator with useful and compact information fur the earlier termination and mitigation of an abnormal state.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1554-1560
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• 2019

Development of digital main control rooms (MCRs) has greatly changed operating environments by altering operator tasks, and thus the unique characteristics of digital MCRs should be considered in terms of human reliability analysis. Digital MCR tasks can be divided into primary tasks that directly supply control input to the plant equipment, and secondary tasks that include interface management conducted via soft controls (SCs). Operator performance regarding these secondary tasks must be evaluated since such tasks did not exist in previous analog systems. In this paper, we analyzed SC-related tasks based on simulation data, and classified the error modes of the SCs following analysis of all operational tasks. Then, we defined the factors to be considered in human reliability analysis methods regarding the SCs; such factors are mainly related to interface management and computerized operator support systems. As these support systems function to reduce the number of secondary tasks required for SC, we conducted an assessment to evaluate the efficiency of one such support system. The results of this study may facilitate the development of training programs as well as help to optimize interface design to better reflect the interface management task characteristics of digitalized MCRs.