• Nuclear Engineering and Technology
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• 제50권5호
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• pp.627-638
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• 2018

This article describes an integrated Level 1-Level 2 probabilistic risk assessment (PRA) methodology to evaluate the radiological risk during postulated accident scenarios initiated during the decommissioning phase of a typical Mark I containment boiling water reactor. The fuel damage scenarios include those initiated while the reactor is permanently shut down, defueled, and the spent fuel is located into the spent fuel storage pool. This article focuses on the integrated Level 1-Level 2 PRA aspects of the analysis, from the beginning of the accident to the radiological release into the environment. The integrated Level 1-Level 2 decommissioning PRA uses event trees and fault trees that assess the accident progression until and after fuel damage. Detailed deterministic severe accident analyses are performed to support the fault tree/event tree development and to provide source term information for the various pieces of the Level 1-Level 2 model. Source terms information is collected from accidents occurring in both the reactor pressure vessel and the spent fuel pool, including simultaneous accidents. The Level 1-Level 2 PRA model evaluates the temporal and physical changes in plant conditions including consideration of major uncertainties. The goal of this article is to provide a methodology framework to perform a decommissioning Probabilistic Risk Assessment (PRA), and an application to a real case study is provided to show the use of the methodology. Results will be derived from the integrated Level 1-Level 2 decommissioning PSA event tree in terms of fuel damage frequency, large release frequency, and large early release frequency, including uncertainties.

• 대한안전경영과학회지
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• 제2권4호
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• pp.33-43
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• 2000

The consequence analysis for the unconfined vapor cloud explosion(UVCE) accident by the continuous release of butane vapor was performed and effects of process parameters on consequences were analyzed in standard conditions. For the case of continuous release(87.8 kg/s) of butane vapor at 8 m elevated height in the debutanizing process of tile naphtha cracking plant operating at 877 kPa ＆ 346.75 K, we found that combustion ranges of dispersed vapor estimated by HMP model were 11.2~120.2 m and overpressures estimated by TNT equivalency model at 200 m were about 37.35~55.1 kPa. Also, overpressures estimated by Model UVCE I based on advective travel time to $X_{LFL}$ were smaller than those estimated by Model UVCE IIbased on real travel time between $X_{UFL}$ and $X_{LFL}$. At the same time, damage intensities at 200 m and effect ranges by overpressure could be predicted. Furthermore, simulation results showed that effects of operating pressures on consequences were larger than those of operating temperatures and results of accidents were increased with increasing operating pressures. At this time, sensitivities of overpressures for UVCE accident by the continuous release were about 5 kPa/atm.

• 한국가스학회지
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• 제17권4호
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• pp.26-32
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• 2013

산업현장에서 중대산업사고를 예방하기 위해서는 원천적으로 위험물질의 사용을 금지하고 안전이 확보된 대체물질을 사용하는 것이 최상의 안전을 확보하는 방법이다. 그러나 대체물질의 비효율적인 경제성과 생산기술의 부재로 인해 위험물질을 취급할 수밖에 없는 상황이라면 사고가 발생하지 않도록 예방을 철저하게 하는 것이 차선의 안전대책이라 하겠다. 이에 본 연구는 최근 연속적인 누출사고로 인해 위험성이 대두 되었음에도 산업현장에서 사용 및 취급될 수밖에 없는 HF에 대해 누출사고가 발생함과 동시에 향후에도 누출사고 가능성이 높은 HF 충진공정의 위험성 평가시 사고결과 영향분석과 비상조치계획 수립에 효율적으로 활용 할 수 있는 사고 시나리오를 발굴 및 선정하였다.

• 한국안전학회지
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• 제38권3호
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• pp.43-50
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• 2023

Silicone tetrachloride (SiCl₄) leak accidents cause enormous human and environmental damage because it is highly toxic. Some handling facilities use water curtains to reduce the impact range of SiCl₄. Although the water curtain is known as one of the most efficient technologies for post-release mitigation, its effect on reducing SiCl₄ concentration needs to be investigated scientifically and quantitatively. In this study, three-dimensional computational fluid dynamics (CFD) was used to investigate the physical and chemical effects of water curtains as a release-mitigation system for SiCl₄. SiCl₄ is released and dispersed five seconds prior to the operation of the water curtain. Once the water curtain works, the SiCl₄ reacts chemically with the water and its concentration decreases rapidly; it reaches an emergency response planning guidelines level 2 (ERPG-2) of 5 parts per million (ppm) at about 570 m. We observed, however, that the physical effect of water curtains on reducing SiCl₄ concentration is insignificant when the chemical effect is eliminated. These results are crucial since they can be a scientific and quantitative basis for the 'technical guidelines for estimating the accident affected range'. In order to protect the public from chemical accidents, more toxic gas mitigation technologies need to be developed.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.125-137
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• 2019

The purpose of this paper is to study the source term behavior after severe accidents by using a semi-kinetic model for simulation and calculation of in-containment activity. The reactor containment specification and the safety features of the containment under different accident conditions play a great role in evaluating the in-containment activity. Assuming in-vessel and instantaneous release of radioactivity into the containment, the behavior of in-containment isotopic activity is studied for noble gasses (Kr and Xe) and the more volatile elements of iodine, cesium, and aerosols such as Te, Rb and Sr as illustrative examples of source term release under LOCA conditions. The results of the activity removal mechanisms indicates that the impact of volumetric leakage rate for noble gasses is important during the accident, while the influence of deposition on the containment surfaces for cesium, mainly iodine isotopes and aerosol has the largest contribution in removal of activity during evolution of the accident.

• 대한안전경영과학회지
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• 제9권3호
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• pp.21-27
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• 2007

This paper analyses the effect of parameters on the consequences of the unconfined vapor cloud explosion accident (UVCE) by the release of heavy gas (xylene vapor). Simulation results showed that the overpresure was increased with the increase of the release hole diameter and with the decrease of the interested distance and the wind speed. While, the overpresure was not nearly affected by the release height, weather and environmental conditions. From the results of the consequence analysis and analysis of affecting the consequences of UVCE, the emergency plan should be established taking into account these parameters.

• 방사성폐기물학회지
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• 제20권3호
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• pp.321-338
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• 2022

The transport of radionuclides at oceanic scales can be assessed using a Lagrangian model. In this review an application of such a model to the Atlantic, Indian and Pacific oceans is described. The transport model, which is fed with water currents provided by global ocean circulation models, includes advection by three-dimensional currents, turbulent mixing, radioactive decay and adsorption/release of radionuclides between water and bed sediments. Adsorption/release processes are described by means of a dynamic model based upon kinetic transfer coefficients. A stochastic method is used to solve turbulent mixing, decay and water/sediment interactions. The main results of these oceanic radionuclide transport studies are summarized in this paper. Particularly, the potential leakage of ¹³⁷Cs from dumped nuclear wastes in the north Atlantic region was studied. Furthermore, hypothetical accidents, similar in magnitude to the Fukushima accident, were simulated for nuclear power plants located around the Indian Ocean coastlines. Finally, the transport of radionuclides resulting from the release of stored water, which was used to cool reactors after the Fukushima accident, was analyzed in the Pacific Ocean.

• Nuclear Engineering and Technology
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• 제47권1호
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• pp.1-10
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• 2015

After the Fukushima Daiichi nuclear power plant (NPP) accident, new regulatory requirements were enforced in July 2013 and a backfit was required for all existing nuclear power plants. It is required to take measures to prevent severe accidents and mitigate their radiological consequences. The Regulatory Standard and Research Department, Secretariat of Nuclear Regulation Authority (S/NRA/R) has been conducting numerical studies and experimental studies on relevant severe accident phenomena and countermeasures. This article highlights fission product (FP) release and hydrogen risk as two major areas. Relevant activities in the S/NRA/R are briefly introduced, as follows: 1. For FP release: Identifying the source terms and leak mechanisms is a key issue from the viewpoint of understanding the progression of accident phenomena and planning effective countermeasures that take into account vulnerabilities of containment under severe accident conditions. To resolve these issues, the activities focus on wet well venting, pool scrubbing, iodine chemistry (in-vessel and ex-vessel), containment failure mode, and treatment of radioactive liquid effluent. 2. For hydrogen risk: because of three incidents of hydrogen explosion in reactor buildings, a comprehensive reinforcement of the hydrogen risk management has been a high priority topic. Therefore, the activities in evaluation methods focus on hydrogen generation, hydrogen distribution, and hydrogen combustion.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2878-2887
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• 2021

As the amount of fission product released from ISLOCA was overestimated because of conservative assumptions in the past, several studies have been recently conducted to evaluate the actual release amount. Among several pathways for the ISLOCA, most studies were focused on the pathway with the highest possibility. However, different ISLOCA pathways may have different fission product release characteristics. In this study, fission product behavior was analyzed for various pathways at the Westinghouse two-loop plant using MELCOR. Four pathways are considered: the pipes from a cold leg, from a downcomer, from a hot leg to the outlet of RHR heat exchanger, and the pipe from the hot leg to the inlet of RHR pump (Pathway 1-4). According to the analysis results, cladding fails at around 2.5 h in Pathways 1 and 2, and on the other hand, about 3.3 h in Pathways 3 and 4 because the ISLOCA pathways affect the safety injection flow path. While the release amount of cesium and iodine ranges between 20 and 26% in Pathways 1 to 3, Pathway 4 allows only 5% to the environment because the break location is submerged. Also, as more than 90% of cesium released to the environment passes through the personnel door, reinforcing the pressure capacity of the doors would be a significant factor in the accident management of the ISLOCA.

• 한국수소및신에너지학회논문집
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• 제25권6호
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• pp.609-619
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• 2014

In the present study, the frequency of the undesired accident was estimated for a quantitative risk assessment of a large-scale hydrogen liquefaction plant. As a representative example, the hydrogen liquefaction plant located in Ingolstadt, Germany was chosen. From the analysis of the liquefaction process and operating conditions, it was found that a $LH_2$ storage tank was one of the most dangerous facilities. Based on the accident scenarios, frequencies of possible accidents were quantitatively evaluated by using both fault tree analysis and event tree analysis. The overall expected frequency of the loss containment of hydrogen from the $LH_2$ storage tank was $6.83{\times}10^{-1}$times/yr (once per 1.5 years). It showed that only 0.1% of the hydrogen release from the $LH_2$ storage tank occurred instantaneously. Also, the incident outcome frequencies were calculated by multiplying the expected frequencies with the conditional probabilities resulting from the event tree diagram for hydrogen release. The results showed that most of the incident outcomes were dominated by fire, which was 71.8% of the entire accident outcome. The rest of the accident (about 27.7%) might have no effect to the population.