• Journal of Hydrogen and New Energy
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• v.25 no.6
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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.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.474-483
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• 2021

To establish the exclusive role of hydrogen on burst behaviour of Zircaloy-4 during loss-of-coolant accident transients, an extensive single-rod burst tests were conducted on both unirradiated as-received and hydrogenated Zircaloy-4 cladding tubes at different heating rates and internal overpressures. The visual observations of cladding tubes during bursting as well as post-burst are presented in detail to understand the effect of hydrogen concentration, heating rate, and internal pressure. Impact of hydrogen on burst parameters-burst stress, burst strain, burst temperature-during loss-of-coolant accident transients are compared and discussed. Rupture at multiple locations for hydrogenated cladding at lower internal pressure and higher heating rate is reported for the very first time. A novel burst criterion accounting hydrogen concentration in nuclear fuel cladding is proposed.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.617-648
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• 2009

Under the government supported long-term nuclear R&D program, the severe accident research program at KAERI is directed to investigate unresolved severe accident issues such as core debris coolability, steam explosions, and hydrogen combustion both experimentally and numerically. Extensive studies have been performed to evaluate the in-vessel retention of core debris through external reactor vessel cooling concept for APR1400 as a severe accident management strategy. Additionally, an improvement of the insulator design outside the vessel was investigated. To address steam explosions, a series of experiments using a prototypic material was performed in the TROI facility. Major parameters such as material composition and void fraction as well as the relevant physics affecting the energetics of steam explosions were investigated. For hydrogen control in Korean nuclear power plants, evaluation of the hydrogen concentration and the possibility of deflagration-to-detonation transition occurrence in the containment using three-dimensional analysis code, GASFLOW, were performed. Finally, the integrated severe accident analysis code, MIDAS, has been developed for domestication based on MELCOR. The data transfer scheme using pointers was restructured with the modules and the derived-type direct variables using FORTRAN90. New models were implemented to extend the capability of MIDAS.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2960-2973
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• 2022

Since Fukushima nuclear power plant (NPP) accident in 2011, the importance of research on various severe accident phenomena has been emphasized. Particularly, detailed analysis of combustion risk is necessary following the containment damage caused by combustion in the Fukushima accident. Many studies have been conducted to evaluate the risk of local hydrogen concentration increases and flame propagation using computational code. In particular, the potential for combustion by local hydrogen concentration in specific areas within the containment has been emphasized. In this study, the process of flame propagation generated inside a reactor core to containment during a loss of coolant accident (LOCA) was analyzed using MELCOR 2.1 code. Later in the LOCA scenario, it was expected that hydrogen combustion occurred inside the reactor core owing to oxygen inflow through the cold leg break area. The main driving force of the oxygen intrusion is the elevated containment pressure due to the molten corium-concrete interaction. The thermal and mechanical loads caused by the flame threaten the integrity of the containment. Additionally, the containment spray system effectiveness in this situation was evaluated because changes in pressure gradient and concentrations of flammable gases greatly affect the overall behavior of ignition and subsequent containment integrity.

• Nuclear Engineering and Technology
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• v.37 no.6
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• pp.537-556
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• 2005

A primary-pipe rupture accident is one of the design-basis accidents of a High-Temperature Gas-cooled Reactor (HTGR). When the primary-pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. This paper describes an experiment and analysis of the air ingress phenomena and the method fur the prevention of air ingress into the reactor during the primary-pipe rupture accident. The numerical results are in good agreement with the experimental ones regarding the density of the gas mixture, the concentration of each gas species produced by the graphite oxidation reaction and the onset time of the natural circulation of air. A hydrogen production system connected to the High-Temperature Engineering Test Reactor (HTTR) Is being designed to be able to produce hydrogen by themo-chemical iodine-Sulfur process, using a nuclear heat of 10 MW supplied by the HTTR. The HTTR hydrogen production system is first connected to a nuclear reactor in the world; hence a permeation test of hydrogen isotopes through heat exchanger is carried out to obtain detailed data for safety review and development of analytical codes. This paper also describes an overview of the hydrogen permeation test and permeability of hydrogen and deuterium of Hastelloy XR.

• Nuclear Engineering and Technology
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• v.38 no.2
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• pp.195-200
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• 2006

In order to confirm the integrity of IRWST following a severe accident, the hydrogen behavior inside and around the IRWST has been investigated for an SBO accident. A detailed containment model, including 18 control volumes for IRWST, has been developed. Analysis results show that the peak hydrogen concentration is about 57% during the core melting period. The combustion regime shows that flame acceleration and DDT are possible in the IRWST. The flame acceleration criterion is met when the peak hydrogen concentration occurs; the 7 -DDT criterion is also met during some periods. These results show certain measures may be required to assure IRWST integrity against an SBO accident.

• Nuclear Engineering and Technology
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• v.51 no.8
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• pp.1939-1950
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• 2019

The MELCOR code useful for a plant-specific hydrogen risk analysis has inevitable limitations in prediction of a turbulent flow of a hydrogen mixture. To investigate the accuracy of the hydrogen risk analysis by the MELCOR code, results for the turbulent gas behavior at pipe rupture accident were compared with CFX results which were verified by the American National Standard Institute (ANSI) model. The postulated accident scenario was selected to be surge line failure induced by station blackout of an Optimized Power Reactor 1000 MWe (OPR1000). When the surge line failure occurred, the flow out of the surgeline was strongly turbulent, from which the MELCOR code predicted that a substantial amount of hydrogen could be released. Nevertheless, the results indicated nonflammable mixtures owing to the high steam concentration released before the failure. On the other hand, the CFX code solving the three-dimensional fluid dynamics by incorporating the turbulence closure model predicted that the flammable area continuously existed at the jet interface even in the rising hydrogen mixtures. In conclusion, this study confirmed that the MELCOR code, which has limitations in turbulence analysis, could underestimate the existence of local combustible gas at pipe rupture accident. This clear comparison between two codes can contribute to establishing a guideline for computational hydrogen risk analysis.

• Journal of computational fluids engineering
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• v.10 no.3 s.30
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• pp.9-18
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• 2005

During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by the active reaction of fuel-cladding and steam in the reactor pressure vessel and released with steam into the containment. In order to mitigate hydrogen hazards possibly occurred in the NPP containment, hydrogen mitigation system (HMS) is usually adopted. The design of the next generation NPP (APR1400) designed in Korea specifies 26 passive autocatalytic recombiners and 10 igniters installed in the containment for the hydrogen mitigation. in this study, the analysis of the hydrogen and steam behavior during a total lose of feed water (TLOFW) accident in the APR1400 containment has been conducted by using the CFD code GASFLOW. During the accident, a huge amount of hot water, steam, and hydrogen is released in the in-containment refueling water storage tank (IRWST). The current design of the APR1400 includes flap-type dampers at the IRWST vents which are operated depending on the pressure difference between inside and outside of the IRWST. it was found that the flaps strongly affects the flow structure of the steam and hydrogen in the containment. The possibilities of a flame acceleration and transition from deflagration to detonation (DDT) were evaluated by using Sigma-Lambda criteria. Numerical results indicate the DDT possibility could be heavily reduced in the IRWST compartment when the flaps are installed.

• Journal of Environmental Health Sciences
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• v.39 no.4
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• pp.346-353
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• 2013

Objectives: The goal of this study is to identify the distribution of the foliar fluorine content of vegetation surrounding the area where hydrofluoric acid was accidently released in Gumi, Gyeongsangbuk-do on September 27, 2012. In addition, it also aims to estimate the concentration of hydrogen fluoride in the air on the day of the accident. Methods: Samples of plant leaves were collected on October 7, 2012 within 1 km from the site where the accident occurred. These samples were analyzed for soluble fluorine ion with an ion selective electrode. The ambient concentration of hydrogen fluoride was calculated using the fluoride content in the plant via the dose-rate equation (${\Delta}F$=KCT). Results: The arithmetic and geometric means of the concentrations were 2158.2 and 1183.7mg F $kg^{-1}$ for leaves and, 2.4 and 1.1 ppm HF for the air, respectively. The highest concentration of hydrogen fluoride in the air was 14.7 ppm, which is higher than the maximum concentration reported by the government (1 ppm) and the exposure limit (ceiling, 3 ppm). The concentrations of both fluorine and hydrogen fluoride decreased with increasing distance from the accident site and showed a significant decrease outside of a 500m radius from the site (p <0.05). Conclusions: The area around the accident site was highly polluted with hydrogen fluoride according to the results of this study. Considering the persistency of hydrogen fluoride in the environment, long-term monitoring and environmental impact assessment should be pursued.

• Nuclear Engineering and Technology
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• v.46 no.2
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• pp.207-216
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• 2014

This paper revisits the Fukushima accident to draw lessons in the aspect of nuclear safety considering the fact that the Fukushima accident resulted in core damage for three nuclear power plants simultaneously and that there is a high possibility of a failure of the integrity of reactor vessel and primary containment vessel. A brief review on the accident progression at Fukushima nuclear power plants is discussed to highlight the nature and characteristic of the event. As the severe accident management measures at the Fukushima Daiich nuclear power plants seem to be not fully effective, limitations of current severe accident management strategy are discussed to identify the areas for the potential improvements including core cooling strategy, containment venting, hydrogen control, depressurization of primary system, and proper indication of event progression. The gap between the Fukushima accident event progression and current understanding of severe accident phenomenology including the core damage, reactor vessel failure, containment failure, and hydrogen explosion are discussed. Adequacy of current safety goals are also discussed in view of the socio-economic impact of the Fukushima accident. As a conclusion, it is suggested that an investigation on a coherent integrated safety principle for the severe accident and development of innovative mitigation features is necessary for robust and resilient nuclear power system.