• Nuclear Engineering and Technology
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• v.44 no.3
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• pp.225-236
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• 2012

The Fukushima-Daiichi accident shook the world, as a well-known plant design, the General Electric BWR Mark I, was heavily damaged in the tsunami, which followed the Great Japanese Earthquake of 11 March 2011. Plant safety functions were lost and, as both AC and DC failed, manoeuvrability of the plants at the site virtually came to a full stop. The traditional system of Emergency Operating Procedures (EOPs) and Severe Accident Management Guidelines (SAMG) failed to protect core and containment, and severe core damage resulted, followed by devastating hydrogen explosions and, finally, considerable radioactive releases. The root cause may not only have been that the design against tsunamis was incorrect, but that the defence against accidents in most power plants is based on traditional assumptions, such as Large Break LOCA as the limiting event, whereas there is no engineered design against severe accidents in most plants. Accidents beyond the licensed design basis have hardly been considered in the various designs, and if they were included, they often were not classified for their safety role, as most system safety classifications considered only design basis accidents. It is, hence, time to again consider the Design Basis Accident, and ask ourselves whether the time has not come to consider engineered safety functions to mitigate core damage accidents. Associated is a proper classification of those systems that do the job. Also associated are safety criteria, which so far are only related to 'public health and safety'; in reality, nuclear accidents cause few casualties, but create immense economical and societal effects-for which there are no criteria to be met. Severe accidents create an environment far surpassing the imagination of those who developed EOPs and SAMG, most of which was developed after Three Mile Island - an accident where all was still in place, except the insight in the event was lost. It requires fundamental changes in our present safety approach and safety thinking and, hence, also in our EOPs and SAMG, in order to prevent future 'Fukushimas'.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.2
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• pp.66-70
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• 2016

Following the 2011 Fukushima Nuclear Power Plant accident, the IAEA has issued a revised version of the Nuclear Safety Standard for beyond design basis earthquake to consider the core meltdown accident. In Korea, relevant laws and regulations were also revised to consider beyond design basis earthquake to nuclear components. In this paper, CAV, an seismic damage factor that determines the restart of nuclear power plant after operating breakdown earthquake, is proposed for extension to the beyond design basis earthquake. For pipings not satisfying the beyond design basis earthquake condition, several evaluation methods are suggested, such as strain-based evaluation methods, simple nonlinear analysis method and cumulative damage evaluation method.

• Proceedings of the KSME Conference
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• 2002.08a
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• pp.51-54
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• 2002

When a cold HPSI (High Pressure Safety Injection) fluid associated with an design basis accident, such as LOCA (Loss of Coolant Accident), enters the cold legs of a stagnated primary coolant loop, thermal stratification phenomena will arise due to incomplete mixing. If the stratified flow enters a reactor pressure vessel downcomer, severe thermal stresses are created in a radiation embrittled vessel wall by local overcooling. Previous thermal-mixing analyses have assumed that the thermal stratification phenomena generated in stagnated loop of a partially stagnated coolant loop are neutralized in the vessel downcomer by strong flow from unstagnated loop. On the basis of these reasons, this paper presents the thermal-mixing analysis results in order to identify the fact that the cold plume generated in the vessel downcomer due to the thermal stratification phenomena of the stagnated loop is affected by the strong flow of the unstagnated loop.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.56-61
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• 2016

After Japanese Fukushima nuclear power plant accident caused by the beyond design basis earthquake and tsunami, it has turned to be a major challenge for nuclear safety. IAEA, US NRC and EU have provided new safety design standards for beyond design basis event, Domestic regulatory bodies have also enacted guidances for licensees and applicants on additional methods related to beyond design basis events. This paper describes several evaluation methods for applying to nuclear power plants piping for beyond design basis earthquake. As a results, energy method based on the absorbed energy on nuclear power plant, deterministic method following design code and theory, experience method considering past earthquake data and information and probabilistic methods similar to probabilistic risk assessment were reviewed.

• Nuclear Engineering and Technology
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• v.37 no.4
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• pp.327-362
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• 2005

The design, construction, and operation of a light water reactor (LWR) are subject to compliance with safety criteria specified for accident situations, such as loss-of-coolant accident (LOCA) and reactivity-initiated accident (RIA). Because reactor fuel is the primary source of radioactivity and heat generation, such a criterion is established on the basis of the characteristics and performance of fuel under the specific accident condition. As such, fuel behavior under accident situations impact many aspects of fuel design and power generation, and in an indirect manner, even spent fuel storage and management. This paper provides a comprehensive review of: the history of the current LOCA criteria, results of LOCA-related investigations on conventional and new classes of fuel, and status of on-going studies on high-burnup fuel under LOCA situations. The objective of the paper is to provide a better understanding of important issues and an insight helpful to establish new LOCA criteria for modem LWR fuels.

• Nuclear Engineering and Technology
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• v.17 no.1
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• pp.53-67
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• 1985

The severe core meltdown accident, which is not included as a design basis accident, has high consequence and low probability of occurrence and turns out to be a major risk factor in the overall risk assessment. The physical mechanisms of containment failure in core meltdown accidents are identified as steam explosion, debris bed coolability, hydrogen burning, steam spike and concrete interaction. The state of technology review is made for each subtopic about the previous and current researches for better understanding of the phenomenon.

• Journal of ILASS-Korea
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• v.8 no.1
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• pp.23-28
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• 2003

When a cold HPSI (High Pressure Safety Injection) fluid associated with a design basis accident, such as LOCA (Loss of Coolant Accident), enters the cold legs of a stagnated primary coolant loop, thermal stratification phenomena may arise due to incomplete mixing. If the stratified flow enters a reactor pressure vessel downcomer, severe thermal stresses are created in a radiation embrittled vessel wall by local overcooling. Previous thermal-mixing analyses have assumed that the thermal stratification phenomena generated in stagnated loop of a partially stagnated collant loop are neutralized in the vessel downcomer by strong flow from unstagnated loop. On the basis of these reasons, this paper presents the thermal-mixing analysis results in order to identify the fact that the cold plume generated in the vessel downcomer due to the thermal stratification phenomena of the stagnated loop is affected by the strong flow of the unstagnated loop.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.535-536
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• 2004

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.165-166
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• 2018

• Journal of the Korea Safety Management & Science
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• v.24 no.1
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• pp.61-72
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• 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.