• Journal of the Korean Society of Safety
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• v.27 no.6
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• pp.192-204
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• 2012

For a severe accident of nuclear power plant, an approach to estimation of the radiological source term using a severe accident code(MELCOR) has been proposed. Although the MELCOR code has a capability to estimate the radiological source term, it has been hardly utilized for the radiological consequence analysis mainly due to a lack of understanding on the relevant function employed in MELCOR and severe accident phenomena. In order to estimate the severe accident source term to be linked with the radiological consequence analysis, this study proposes 4-step procedure: (1) selection of plant condition leading to a severe accident(i.e., accident sequence), (2) analysis of the relevant severe accident code, (3) investigation of the code analysis results and post-processing, and (4) generation of radiological source term information for the consequence analysis. The feasibility study of the present approach to an early containment failure sequence caused by a fast station blackout(SBO) of a reference plant (OPR-1000), showed that while the MELCOR code has an integrated capability for severe accident and source term analysis, it has a large degree of uncertainty in quantifying the radiological source term. Key insights obtained from the present study were: (1) key parameters employed in a typical code for the consequence analysis(i.e., MACCS) could be generated by MELCOR code; (2) the MELOCR code simulation for an assessment of the selected accident sequence has a large degree of uncertainty in determining the accident scenario and severe accident phenomena; and (3) the generation of source term information for the consequence analysis relies on an expert opinion in both areas of severe accident analysis and consequence analysis. Nevertheless, the MELCOR code had a great advantage in estimating the radiological source term such as reflection of the current state of art in the area of severe accident and radiological source term.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.19 no.4
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• pp.370-377
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• 2009

A simple preliminary analysis is often useful to check a validity of design alternatives before the detailed analysis phase in the viewpoint of efficiency. This paper describes a preliminary analysis procedure for the selection among basket design candidates for the spent fuel shipping cask of Korean standard nuclear power plant. As the cask should maintain the structural integrity in hypothetical accident condition, the case of 9 m drop is significantly considered as the worst scenario among the accident conditions in structural design viewpoint in this paper. As basket design options, totally four different types are considered and analyzed in the point of structural integrity at drop impact and weldability for fabrication. As a result, an insertion round plate type with densely spaced supports turns out to be the best in both of the viewpoints, though the weld plate type shows a bit more design margin.

• Journal of Hydrogen and New Energy
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• v.31 no.6
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• pp.605-613
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• 2020

In July and October of this year, the government announced the 'Green new deal plan within the Korean new deal policy' and 'Strategies for proliferation of future vehicles and market preoccupation'. And, in response to changes in the global climate agreement, it has decided to expand green mobility such as electric vehicles and hydrogen electric vehicles with the aim of a "net-zero" society. Accordingly, the goal is to build 310 hydrogen refueling stations along with the supply of 60,000 hydrogen vehicles in 2022, and the hydrogen infrastructure is being expanded. however, it is difficult to secure hydrogen infrastructure due to expensive construction costs and difficulty the selection of a site. In Korea, it is possible to build a mobile hydrogen station according to the safety standards covering special case of the Ministry of Industry. Since the mobile hydrogen station can be charged while moving between authorized place, it has the advantage of being able to meet a large number of demands with only one hydrogen refueling station, so it is proposed as a model suitable for the early market of hydrogen infrastructure. This study demonstrates the establishment of a hydrogen refueling station by deriving a virtual accident scenario for leakage and catastrupture for each facility for the risk factors in a mobile hydrogen station, and performing a quantitative risk assessment through the derived scenario. Through the virtual accident scenario, direction of demonstration and implications for the construction of a mobile hydrogen refueling station were derived.

• Fire Science and Engineering
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• v.15 no.4
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• pp.24-33
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• 2001

This article proposes a strategy for producing accident scenarios in quantitative risk, which is peformed in process design or operation steps. Present worldwide chemical processes need off-site risk assessment as well as on-site one. Most governments in the world require industrial companies to submit the proper emergency plans through off-site risk assessment. Korea is also preparing for executing Integrated Risk Management System along with PSM and SMS. However.

• Journal of the Korean Institute of Gas
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• v.17 no.4
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• pp.26-32
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• 2013

The best way to prevent major occupational accidents is prohibiting use of hazardous substances such as flammable gas, toxic gas whereas using alternative substances that ensured safety. but if there are no economic efficiency and substituting technologies of alternative substances, the best way is preparing to prevent accidents thoroughly. Therefore, this study has developed and selected release scenarios to use and apply for consequence analysis and emergency action plan for HF charging process of chemical plants that have HF release accidents and high probability of release accidents.

• Nuclear Engineering and Technology
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• v.56 no.8
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• pp.2893-2905
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• 2024

Passive safety systems are integrated into the latest generation of Light Water Reactors (LWRs), including small modular reactors. This paper employs the US-NRC TRACE thermal hydraulic code to examine the performance of a passive safety condenser known as SACO, designed to serve as the ultimate heat sink for dissipating decay heat during accident scenarios. The TRACE model is constructed with reference to the PKL/SACO test facility. The safety condenser (SACO) is interconnected with the PKL facility via the secondary side of steam generator 1, effectively serving as a third natural circulation cooling loop during accident scenarios. In the present research, the thermal-hydraulic behavior of the PKL facility is investigated in the presence of the SACO passive safety system during an extended SBO with Loss of AC Power accident scenario. This SBO can be categorized into three distinct phases depending on the activation of the SACO system and the refilling process of the SACO pool. The first phase is depressurizing using primary and secondary relief valves, the second phase is cooling down using SACO system, and the third phase is the refilling of SACO pool. The findings indicate that the SACO system effectively manages to dissipate all decay heat, even though there is temporary evaporation of the SACO water pool. Furthermore, this study provides sensitivity analysis for the assessments of system codes on the selection of maximum time step.

• Journal of the Korean Institute of Gas
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• v.21 no.1
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• pp.43-51
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• 2017

Recently, ammonia leak occurred frequently in the domestic refrigeration manufacturing facilities. Ammonia caused great damage to the environment and human health in the event of an accident as combustible gases and toxic gases. After considering the types of ammonia accidents of domestic refrigeration manufacturing facilities and selected accident scenarios and to analyze the risk analysis through Impact range estimates and frequency analysis and there was a need to establish measures to minimize accident damage. In this study, depending on the method of analysis quantitative risk assessment we analyzed the risk of the receiver tank of ammonia system. Scenario analysis conditions were set according to the 'Technical guidelines for the selection of accident scenario' under the chemicals control act and 'Guidelines for chemical process quantitative risk analysis' of center for chemical process safety. The risk estimates were utilized for consequence analysis and frequency analysis by SAFETI program of DNV, event tree analysis methodology and part count methodology. The individual risk of ammonia system was derived as 7.71E-04 / yr, social risk were derived as 1.17E-03 / yr. The derived risk was confirmed to apply as low as reasonably practicable of the national fire protection association and through risk calculation, it can be used as a way to minimize accidents ammonia leakage accident damage.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.18 no.11
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• pp.16-22
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• 2017

This study examined the status of resident evacuation sites notified to nearby communities, centered on business sites subject to the risk management plan of Incheon, Siheung, and Ansan. Through an analysis of the meteorological data, the direction of improvement of the site selection process for the safe evacuation of chemical accidents was studied. Among a total of 111 evacuation sites, 30 schools were selected the most, and 2-3 sites were usually selected for evacuation. As a result of an analysis of the Incheon meteorological data of 2016, the frequency of occurrences was 18.8525% in the NNE wind direction, 18.0328% in the NNW wind direction, 12.2951% in the WSW wind direction, 9.0164% in the SSE direction, 8.4700% in the SW direction, 6.5574% in the W direction, and 5.7376% in the S direction. The NNE wind direction showed the highest frequency, but the other winds showed a relatively high frequency, indicating that the annual wind direction was not biased toward one side.

• Journal of the Korean Institute of Gas
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• v.22 no.2
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• pp.90-96
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• 2018

Design techniques for minimizing the damage caused by leakage of $H_2S$ gas, contained in natural gas and petroleum, have been widely studied abroad in chemical plants that purify and process natural gas and petroleum. However, there is no domestic engineering practice and regulation of $H_2S$. In accordance with the circumstances, this study proposes the quantitative criteria of process equipment to install $H_2S$ toxic gas detector as 500 ppm and explains the valid basis. The $H_2S$ gas dispersion radius up to IDLH 100 ppm is calculated by ALOHA under previous $H_2S$ gas leak accident scenario. The weather conditions of modeling include the conditions of Ulsan, Yeosu and Daesan, the three major petrochemical complexes in Korea. The long radius up to 100 ppm was derived in order of Ulsan, Daesan, Yeosu. For emergency safety the dispersion radius up to 100 ppm of the $H_2S$ gas obtained in this study should be extended to apply the additional $H_2S$ toxic gas detector, and local climate conditions should be considered.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.4
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• pp.237-245
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• 2011

As necessity of safety re-evaluation for spent fuel storage facility has emphasized after the Fukushima accident, accuracy improvement of heat load evaluation has become more important to acquire reliable thermal-hydraulic evaluation results. As groundwork, parametric and sensitivity analyses of various storage conditions for Kori Unit 4 spent fuel storage pool and spent fuel depletion parameters such as axial burnup effect, operation history, and specific heat are conducted using ORIGEN2 code. According to heat load evaluation and parametric sensitivity analyses, decay heat of last discharged fuel comprises maximum 80.42% of total heat load of storage facility and there is a negative correlation between effect of depletion parameters and cooling period. It is determined that specific heat is most influential parameter and operation history is secondly influential parameter. And decay heat of just discharged fuel is varied from 0.34 to 1.66 times of average value and decay heat of 1 year cooled fuel is varied from 0.55 to 1.37 times of average value in accordance with change of specific power. Namely depletion parameters can cause large variation in decay heat calculation of short-term cooled fuel. Therefore application of real operation data instead of user selection value is needed to improve evaluation accuracy. It is expected that these results could be used to improve accuracy of heat load assessment and evaluate uncertainty of calculated heat load.