• Nuclear Engineering and Technology
• /
• v.52 no.9
• /
• pp.1983-1989
• /
• 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.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.5 no.2
• /
• pp.40-44
• /
• 2009

In this study, Event Corrective Action Supporting System (ECAS) is developed for the accident evaluation in nuclear power plant. The ECAS system can be used in supporting regulator and/or operator under event situation in nuclear power plants. The ECAS system consists of 5 modules including failure location module, failure analysis module, failure integrity evaluation module, system vulnerability evaluation module, and reporting and operating experience feedback module. The ECAS system will be used as sub module of Knowledge-Based Event Evaluation Network (K-EvENT) which is developing for the against the accident in nuclear power plants.

• Nuclear Engineering and Technology
• /
• v.55 no.10
• /
• pp.3599-3616
• /
• 2023

After a nuclear power plant (NPP) accident, it would be helpful to predict the movement of the radioactive plume emitted from the NPP as accurately as possible to protect the nearby population. Radioactive plumes are mainly affected by wind direction and speed. Since it is difficult to identify the wind direction and speed immediately after the accident, a good understanding of the historical wind data could save many lives and ensure smoother evacuation procedures. In this study, wind data for the past 10 years are analyzed for the five NPPs in the Republic of Korea (ROK). The analyzed data include wind direction and wind speed from 2012 to 2021. In particular, the characteristics of the wind field blowing from the NPPs to the nearest densely populated regions are examined. Finally, suggestions to improve evacuation plans are made.

• Proceedings of the Korean Society of Precision Engineering Conference
• /
• 2012.10a
• /
• pp.849-850
• /
• 2012

• Korean System Dynamics Review
• /
• v.15 no.1
• /
• pp.51-74
• /
• 2014

Despite the high efficiency of nuclear power plant, people in Korea do not give approvals and supports the facilities because the risk of the accidents and incidents. In particular, the low level of safety culture is a crucial mechanism that damages the robustness of the NPP. By considering the various definitions of safety culture and analyzing the major reasons of incidents, the conceptual safety culture model is made by using Causal Loop Diagramming. For sustaining development of nuclear power, social supports, incentives and organizational learning are needed. It also requires the coordination of work schedules and the expansion of human resource for protecting the rules and procedures in NPP. Decommissioning aging nuclear power plants will prevent a serious accident. In order to promote the safety culture, Korea Hydro & Nuclear Power Corporation should disclose more information to the public and promote the internal and external communications.

• Nuclear Engineering and Technology
• /
• v.47 no.1
• /
• pp.11-25
• /
• 2015

The accident at Japan's Fukushima Daiichi nuclear power plant in March 2011, caused by an earthquake and a subsequent tsunami, resulted in a failure of the power systems that are needed to cool the reactors at the plant. The accident progression in the absence of heat removal systems caused Units 1-3 to undergo fuel melting. Containment pressurization and hydrogen explosions ultimately resulted in the escape of radioactivity from reactor containments into the atmosphere and ocean. Problems in containment venting operation, leakage from primary containment boundary to the reactor building, improper functioning of standby gas treatment system (SGTS), unmitigated hydrogen accumulation in the reactor building were identified as some of the reasons those added-up in the severity of the accident. The Fukushima accident not only initiated worldwide demand for installation of adequate control and mitigation measures to minimize the potential source term to the environment but also advocated assessment of the existing mitigation systems performance behavior under a wide range of postulated accident scenarios. The uncertainty in estimating the released fraction of the radionuclides due to the Fukushima accident also underlined the need for comprehensive understanding of fission product behavior as a function of the thermal hydraulic conditions and the type of gaseous, aqueous, and solid materials available for interaction, e.g., gas components, decontamination paint, aerosols, and water pools. In the light of the Fukushima accident, additional experimental needs identified for hydrogen and fission product issues need to be investigated in an integrated and optimized way. Additionally, as more and more passive safety systems, such as passive autocatalytic recombiners and filtered containment venting systems are being retrofitted in current reactors and also planned for future reactors, identified hydrogen and fission product issues will need to be coupled with the operation of passive safety systems in phenomena oriented and coupled effects experiments. In the present paper, potential hydrogen and fission product issues raised by the Fukushima accident are discussed. The discussion focuses on hydrogen and fission product behavior inside nuclear power plant containments under severe accident conditions. The relevant experimental investigations conducted in the technical scale containment THAI (thermal hydraulics, hydrogen, aerosols, and iodine) test facility (9.2 m high, 3.2 m in diameter, and $60m^3$ volume) are discussed in the light of the Fukushima accident.

• Asian Journal for Public Opinion Research
• /
• v.5 no.1
• /
• pp.1-14
• /
• 2017

The Fukushima nuclear accident caused by an earthquake and a subsequent tsunami on March 11, 2011 has seriously impacted the environment surrounding the Fukushima Daiichi nuclear power plant. While all the residents near the plant were evacuated from the area deemed uninhabitable after the accident, residents of the neighboring area outside of the evacuation zone still seem to live in fear of invisible radiation. To understand Fukushima residents' thinking about the environmental risks that accompany a nuclear disaster, we utilize a poll of the residents of Fukushima conducted in 2013. Based on the survey data, we reveal factors that seem to strongly affect their knowledge and concerns about nuclear power plants. The results of the multivariate analysis show the importance of the following two factors: (1) confidence in mass media, and (2) trust in institutions in charge of administering the accident, especially the central government, the Nuclear and Industrial Safety Agency, and Tokyo Electric Power Company. We conclude that the more people trust mass media and particular institutions, the more likely it is that they are have an elevated sense of anxiety and fear of the presence of nuclear plants.

• Nuclear Engineering and Technology
• /
• v.53 no.2
• /
• pp.532-548
• /
• 2021

Nuclear power plants contain several monitoring systems that can identify the in-vessel phenomena of a severe accident (SA). Though a lot of analysis and research is carried out on SA, right from the development of the nuclear industry, not all the possible circumstances are taken into consideration. Therefore, to improve the efficacy of the safety of nuclear power plants, additional analytical studies are needed that can directly monitor severe accident phenomena. This paper presents an interacting multiple model (IMM) based fault detection and diagnosis (FDD) approach for the identification of in-vessel phenomena to provide the accident propagation information using reactor vessel (RV) out-wall temperature distribution during severe accidents in a nuclear power plant. The estimation of wall temperature is treated as a state estimation problem where the time-varying wall temperature is estimated using IMM employing three multiple models for temperature evolution. From the estimated RV out-wall temperature and rate of temperature, the in-vessel phenomena are identified such as core meltdown, corium relocation, reactor vessel damage, reflooding, etc. We tested the proposed method with five different types of SA scenarios and the results show that the proposed method has estimated the outer wall temperature with good accuracy.

• Journal of the Korean Society of Safety
• /
• v.16 no.1
• /
• pp.84-87
• /
• 2001

Severe accident management can be defined as the use of existing and alternative resources, systems, and actions to prevent or mitigate a core-melt accident in nuclear power plants. TRAIN (Training pRogram for AMP In NPP), developed for training control room staff and the technical group, is introduced in this paper. The TRAIN composes of phenomenological knowledge base (KB), accident sequence KB and accident management procedures with AM strategy control diagrams and information needs. This TRAIN might contribute to training them by obtaining phenomenological knowledge of severe accidents, understanding plant vulnerabilities, and solving problems under high stress.

• Nuclear Engineering and Technology
• /
• v.28 no.5
• /
• pp.467-481
• /
• 1996

An integrated framework of modeling the human operator cognitive behavior during nuclear power plant accident scenarios is presented. It incorporates both plant and operator models. The basic structure of the operator model is similar to that of existing cognitive models, however, this model differs from those existing ones largely in too aspects. First, using frame and membership function, the pattern matching behavior, which is identified as the dominant cognitive process of operators responding to an accident sequence, is explicitly implemented in this model. Second, the non-task-related human cognitive activities like effect of stress and cognitive biases such as confirmation bias and availability bias, are also considered. A computer code, OPEC is assembled to simulate this framework and is actually applied to an SGTR sequence, and the resultant simulated behaviors of operator are obtained.