• Journal of the Korean Society of Safety
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• v.24 no.3
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• pp.54-58
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• 2009

A substitute energy development has been required due to the exhaust of the fossil fuel and an environmental problem. Consequently, possible technologies producing hydrogen from water that does not release carbon is a very promising technology. Also, Iodine-Sulfur(IS) thermochemical water decomposition is one of the promising processes that are used to produce hydrogen efficiently using the high temperature gas-cooled reactor(HTGR) as an energy source that is possible to supply heat over 900$^{\circ}C$. In this study, to make a initiating events identification for the IS process, Master Logic Diagram(MLD) is used and 9 initiating events that cause a leakage of the chemical material are identified. Also, 6 events are identified among 9 initiating events above and are quantified using event tree.

• Nuclear Engineering and Technology
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• v.31 no.6
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• pp.51-65
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• 1999

The present review report introduces the existing analysis codes and physical modeling of two-phase flow associated with initiating event of HCDA in Liquid Metal Reactors for the effective study in the future, because the related research has not been systematically carried out in Korea compared with other areas. The description in this report is specifically addressed to the results yielded from careful review of the technical concepts on the two-phase flow modeling in the SAS2A code which was developed in ANL. The report is prepared in 2 parts based on the definite physical phenomena. The liquid slug and gas behavior models are main representations in the part (I) and (II), respectively. In this regard, it is expected that this report provide a fundamental knowledge on the two-phase flow model in LMR and, thus, contribute to establishment of the necessary HCDA analysis technology concerned with the LMR development in Korea.

• Journal of Korean Society for Quality Management
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• v.19 no.2
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• pp.96-107
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• 1991

At the design stage of a plant, the plausible causes and pathways of release of hazardous materials are not clearly known. Thus there exist large amount of uncertainties on the consequences resulting from the operation of a fusion plant. In order to better handle such uncertain circumstances, we utilize the Probabilistic Risk Assessment(PRA) for the safety analyses on fusion power plant. In this paper, we concentrate on the tritium release accident. We develop a simple model that describes the process and flow of tritium, by which we figure out the locations of tritium inventory and their vulnerability. We construct event tree models that lead to various levels of tritium release from abnormal initiating events. Branch parameters on the event tree are assessed from the fault tree analysis. Based on the event tree models we construct influence diagram models which are more useful for the parameter updating and analysis. We briefly discuss the parameter updating scheme, and finally develop the methodology to obtain the predictive distribution of consequences resulting from the operating a fusion power plant. We also discuss the way to utilize the results of testing on sub-systems to reduce the uncertain ties on over all system.

• Journal of the Korean Society of Safety
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• v.31 no.3
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• pp.34-41
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• 2016

The purpose of this paper is to conduct ETA on six items of ESS: the whole system, battery, BMS, PCS, ESS and cable. To achieve that, ESS work flow and its components are categorized. Based on performance, human, environmental, management, and safety, this paper drew initiation events (IE) and end states (ES). ETA is applied to the main functions of each item, and the end states that may occur in one initiation event are suggested. In addition, detailed classification was performed to induce various end states on the basis of the suggested initiation events ; loss of grid electricity of ESS, loss of battery electricity(DC) of battery, impairment of electric function of BMS, loss of grid electricity(AC) of PCS, loss of data of EMS, Mechanical damage of cable, event sequence analysis conducted on the basis of event trees. If the suggested IEs and ESs are applied on the basis of ESS event cases, it is expected to prevent the same kinds of accident and operate ESS safely.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.35-39
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• 2013

The Operating Experience Report(OER) has written about the event and accident happened at a Nuclear Power Plant(NPP). The purpose of publishing the OER is to prevent the similar event or accident repeatedly by spreading the experience of a single plant to other plants personnel. Before initiating the analysis mentioned in this paper, 2,298 review reports for the same number of OER published from 2007 to June 2012 have been written to achieve the correct and objective statistics. The analysis introduced in this paper is performed with the various factors such as year, plant type, equipment, type of work, root-cause. The root-cause analysis is showed that the equipment problem is the major factor in domestic NPPs, but on the other hand human-error is the main part of the foreign NPPs. Moreover, while the number of the man-made event is decreasing, the equipment-made event is rapidly increasing in domestic NPPs.

• Journal of the Society of Naval Architects of Korea
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• v.59 no.3
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• pp.164-172
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• 2022

A HAZID is a brainstorming workshop to identify hazards in an early phase of a project. It should be flexible to capture all probable accidents allowing experienced participants to exploit their expertise and experiences. A bowtie analysis is a graphical representation of major accident hazards elaborating safety measures i.e. barriers. The result of these workshops should be documented in an organized manner to share as good as possible details of the discussion through the lifetime of the project. Currently results are documented using a three-step representation of an accident; causes, top event and consequences, which cannot capture correctly sequence of events leading to various accidents and roles of barrier between two events. Another problem is that barriers would be shown repeatedly leading to a misunderstanding that there are an enough number of safety measures. A new bowtie analysis method is proposed to describe an accident in multiple steps showing relations among causes or consequences. With causes and consequences shown in a format of a tree, the frequencies of having the top event (Fault tree analysis) and various consequences (Event tree analysis) are evaluated automatically based on the frequency of initiating causes and the probabilities of failure of barriers. It will provide a good description of the accident scenario and help the risk to be assessed transparently.

• Journal of the Korean Society of Safety
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• v.18 no.3
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• pp.75-80
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• 2003

This paper presents a set of typical pseudo-accident scenarios related to major equipments in petrochemical plants, which would be useful for performing such quantitative risk analysis techniques as fault tree analysis, event tree analysis, etc. These typical scenarios address what the main hazard of each equipment might be and how the accident might develop from an "initiating event". The proposed set of accident scenarios consists of total thirteen (13) scenarios specific for five (5) major equipments like reactor, distillation column, etc., and has been determined and screened out of one hundred and twenty-five (125) potential accident scenarios that were generated by performing semi-quantitative risk analysis practically for twenty-five (25) petrochemical processes, considering advices from the operation experts. It is assumed that with simple consideration or incorporation of plant-specific conditions only, the proposed accident scenarios could be easily reorganized or adapted for the relevant process with less time and labor by the safety engineers concerned in the petrochemical industries.ndustries.

• International journal of steel structures
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• v.18 no.4
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• pp.1470-1481
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• 2018

Building-level response to post-earthquake fire hazards in steel buildings has been assessed using primarily two-dimensional analyses of the lateral force resisting system. This approach may not adequately consider potential vulnerabilities in the gravity framing system. For this reason, three-dimensional (3D) finite element models of a 10-story case study building with perimeter moment resisting frames were developed to analyze post-earthquake fire events and better understand building response. Earthquakes are simulated using ground motion time histories, while Eurocode parametric time-temperature curves are used to represent compartment fires. Incremental dynamic analysis and incremental fire analysis procedures capture a range of hazard intensities. Findings show that the structural response due to earthquake and fire hazards are somewhat decoupled from one another. Regardless of the level of plastic hinging present in the moment framing system due to a seismic event, gravity column failure is the initiating failure mode in a fire event.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.spc
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• pp.63-74
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• 2020

The operational safety assessment is an important part of a safety case for the deep geological repository of spent fuels. It consists of different stages such as the identification of initiating events, event tree analysis, fault tree analysis, and evaluation of exposure doses to the public and radiation workers. This study develops a probabilistic safety assessment method for the operational safety assessment and establishes an assessment framework. For the event and fault tree analyses, we propose the advanced information management system for probabilistic safety assessment (AIMS-PSA Manager). In addition, we propose the Radiological Safety Analysis Computer (RSAC) program to evaluate exposure doses to the public and radiation workers. Furthermore, we check the applicability of the assessment framework with respect to drop accidents of a spent fuel assembly arising out of crane failure, at the surface facility of the KRS⁺ (KAERI Reference disposal System for SNFs). The methods and tools established through this study can be used for the development of a safety case for the KRS⁺ system as well as for the design modification and the operational safety assessment of the KRS⁺ system.

• Nuclear Engineering and Technology
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• v.44 no.3
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• pp.311-322
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• 2012

Since the crisis at the Fukushima plants, severe accident progression during a station blackout accident in nuclear power plants is recognized as a very important area for accident management and emergency planning. The purpose of this study is to investigate the comparative characteristics of anticipated severe accident progression among the three typical types of nuclear reactors. A station blackout scenario, where all off-site power is lost and the diesel generators fail, is simulated as an initiating event of a severe accident sequence. In this study a comparative analysis was performed for typical pressurized water reactor (PWR), boiling water reactor (BWR), and pressurized heavy water reactor (PHWR). The study includes the summarization of design differences that would impact severe accident progressions, thermal hydraulic/severe accident phenomenological analysis during a station blackout initiated-severe accident; and an investigation of the core damage process, both within the reactor vessel before it fails and in the containment afterwards, and the resultant impact on the containment.