• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.2
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• pp.183-191
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• 2023

The potential impact of hypothetical accidents that occur during the immediate and deferred dismantling of the Kori Unit 1 steam generator has been comprehensively evaluated. The evaluation includes determining the inventory of radionuclides in the Steam Generator based on surface contamination measurements, assuming a rate of release for each accident scenario, and applying external and internal exposure dose coefficients to assess the effects of radionuclides on human health. The evaluation also includes calculating the atmospheric dispersion factor using the PAVAN code and analyzing three years of meteorological data from Kori NPP to determine the degree of diffusion of radionuclides in the atmosphere. Overall, the effective dose for residents living in the Exclusion Area Boundary (EAB) of Kori NPP is predicted, an it is found that the maximum level of the dose is 0.034% compared to the annual dose limit of 1 mSv for the general public. This implies that the potential impact of hypothetical accidents on human health discussed above is within acceptable limits.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2120-2134
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• 2022

A hypothetical core destructive accident (HCDA) has received widespread attention as one of the most serious accidents in sodium-cooled fast reactors. This study combined recent advantages in numerical methods to realize realistic modeling of the complex fluid-structure interactions during HCDAs in a full-scale sodium-cooled fast reactor. The multi-material arbitrary Lagrangian-Eulerian method is used to describe the fluid-structure interactions inside the container. Both the structural deformations and plug rises occurring during HCDAs are evaluated. Two levels of expansion energy are considered with two different reactor models. The simulation results show that the container remains intact during an accident with small deformations. The plug on the top of the container rises to an acceptable level after the sealing between the it and its support is destroyed. The methodology established in this study provides a reliable approach for evaluating the safety feature of a container design.

• Journal of the Korea Safety Management & Science
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• v.16 no.1
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• pp.101-108
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• 2014

This study highlights the theme of human error of train drivers, conducting empirical analysis on the relationship between emergency stress, psychological fatigue, safety behavior, and accident. The hypothetical test results based on questionnaires received from 223 train drivers working at A subway firm indicate that emergency stress shows a significant positive effect on psychological fatigue, which in turn shows a significant negative influence on safety behavior. And safety behavior is shown having a significant negative relationship with accident. These results suggest the necessity of corporate-level approaches to depict the drastic causes of drivers' emergency stress, and to effectively manage this stress, as well as the necessity of making effort to enhance safety behavior, and to prevent or reduce accidents.

• 한국전산유체공학회:학술대회논문집
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• 2009.11a
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• pp.121-128
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• 2009

During a hypothetical high-pressure accident in a nuclear power plant (NPP), molten corium can be ejected through a breach of a reactor pressure vessel (RPV) and dispersed by a following jet of a high-pressure steam in the RPV. The dispersed corium is fragmented into smaller droplets in a reactor cavity of the NPP by the steam jet and released into other compartments of the NPP by a overpressure in the cavity. The fragments of the corium transfer thermal energy to the ambient air in the containment or interact chemically with steam and generate hydrogen which may be burnt in the containment. The thermal loads from the ejected molten corium on the containment which is called direct containment heating (DCH) can threaten the integrity of the containment. DCH in a NPP containment is related to many physical phenomena such as multi-phase hydrodynamics, thermodynamics and chemical process. In the evaluation of the DCH load, the melt dispersion rates depending on the RPV pressure are the most important parameter. Mostly, DCH was evaluated by using lumped-analysis codes with some correlations obtained from experiments for the dispersion rates. In this study, MC3D code was used to evaluate the dispersion rates in the APR1400 NPP during the high-pressure accidents. MC3D is a two-phase analysis code based on Eulerian four-fields for melt jet, melt droplets, gas and water. The dispersion rates of the corium melt depending on the RPV pressure were obtained from the MC3D analyses and the values specific to the APR1400 cavity geometry were compared to a currently available correlation.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.467-472
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• 1996

This study discusses special aspects of severe accident related recriticality modeling and analysis in the Advanced Neutron Source (ANS) reactor.$^{1, 2)}$ The analytical comparison of three elements core to former two elements case is conducted including evaluation of suitable nuclear cross-section sets to account for the effects of system configulation, fuel and moderator mixture temperature, material dispersion and the other thermal-hydraulics. Three elements core ANS reactor is the alternative core design which was proposed as a modified core design, with three fuel elements instead of two, that would allow operation with only 50% enriched uranium (former uranium fuel is the baseline design value of 93%) A comprehensive test matrix of calculations to evaluate the threat of a criticality event in the ANS is described. Strong dependencies still on geometry, material constituents, and thermal-hydraulic conditions are verified. Therefore, the concepts of mitigative design features are qualified.d.

• International Journal of Computer Science & Network Security
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• v.22 no.3
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• pp.107-112
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• 2022

The automotive industry is motivated to provide more and more amenities to its customers. The industry is taking advantage of artificial intelligence by increasing different sensors and gadgets in vehicles machoism is forward collision warning, at the same time road accidents are also increasing which is another concern to address. So there is an urgent need to provide an A.I based system to avoid such incidents which can be address by using artificial intelligence and global positioning system. Automotive/smart vehicles protection has become a major study of research for customers, government and also automotive industry engineers In this study a two layered novel hypothetical approach is proposed which include in-time vehicle/obstacle detection with auto warning mechanism for collision detection & avoidance and later in a case of an accident manifestation GPS & video camera based alerts system and interrupt generation to nearby ambulance or rescue-services units for in-time driver rescue.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.3
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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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• v.45 no.6
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• pp.721-730
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• 2013

Within the framework of the ASTRID project, core design studies are being conducted by the CEA with support from AREVA and EDF. The pre-conceptual design studies are being conducted in accordance with the GEN IV reactor objectives, particularly in terms of improving safety. This involves limiting the consequences of 1) a hypothetical control rod withdrawal accident (by minimizing the core reactivity loss during the irradiation cycle), and 2) an hypothetical loss-of-flow accident (by reducing the sodium void worth). Two types of cores are being studied for the ASTRID project. The first is based on a 'large pin/small spacing wire' concept derived from the SFR V2b, while the other is based on an innovative CFV design. A distinctive feature of the CFV core is its negative sodium void worth. In 2011, the evaluation of a preliminary version (v1) of this CFV core for ASTRID underlined its potential capacity to improve the prevention of severe accidents. An improved version of the ASTRID CFV core (v2) was proposed in 2012 to comply with all the control rod withdrawal criteria, while increasing safety margins for all unprotected-loss-of-flow (ULOF) transients and improving the general design. This paper describes the CFV v2 design options and reports on the progress of the studies at the end of pre-conceptual design phase 1 concerning: - Core performance, - Intrinsic behavior during unprotected transients, - Simulation of severe accident scenarios, - Qualification requirements. The paper also specifies the open options for the materials, sub-assemblies, absorbers, and core monitoring that will continue to be studied during the conceptual design phase.

• Journal of Radiation Protection and Research
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• v.9 no.1
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• pp.26-32
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• 1984

The exposure doses by the radioactive gaseous effluents from nuclear power plants are investigated in the two cases of normal operation and hypothetical accident. Gaussian equation is adapted in the normal operation as the diffusion model of effluents for long period, which uses annual average meteorological data. But the real time models have been used in the case of accidents which analyze the changes of wind direction and speed. In this study the annual exposure doses by the normal operation of Kori unit 1 during $1977{\sim}1982$ were calculated on the basis of the atmospheric diffusion factor by the Gaussian straight line model. And the image processing technique was suggested as the effective method through the wind tunnel experiments to get the characteristic value of atmospheric diffusion coefficient required especially in the accidents of nuclear power plants.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.937-946
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• 2020

The aim in the present work is to simulate accident scenarios of AP1000 during the small-break loss-of-coolant accident (SBLOCA) and investigate the performance and behavior of automatic depressurization system (ADS) during accidents by using MIDAC (The Module In-vessel Degradation severe accident Analysis Code). Four types of accidents with different hypothetical conditions were analyzed in this study. The impact on the thermal-hydraulic of the reactor coolant system (RCS), the passive core cooling system and core degradation was researched by comparing these types. The results show that the RCS depressurization becomes faster, the core makeup tanks (CMT) and accumulators (ACC) are activated earlier and the effect of gravity water injection is more obvious along with more ADS valves open. The open of the only ADS1-3 can't stop the core degradation on the basis of the first type of the accident. The open of ADS1-3 has a great impact on the injection time of ACC and CMT. The core can remain intact for a long time and the core degradation can be prevent by the open of ADS-4. The all results are significant and meaningful to understand the performance and behavior of the ADS during the typical SBLOCA.