• 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.

• 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.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.475-479
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• 1995

During the hypothetical loss of coolant accident in the nuclear power plant the emergency core cooling water could not penetrate to the reactor core when the steam flow rate from the reactor core exceeds CCFL (Countercurrent flow limitation). The CCFL generated by earlier investigators are developed under the steady gas flow. However the flow instability in the reactor loop could generate oscillatory steam flow, hence their applicability under oscillating flow should be investigated. In this work, an experimental investigation of countercurrent flow in the vertical flow channel has been conducted under oscillatory gas flow. Pulsation of gas under oscillatory flow disturbs the flow pattern significantly and prevents flooding (CCFL) when its minimum value is less than the threshold gas flow rate value.

• Proceedings of the Korean Society for Technology of Plasticity Conference
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• 2002.05a
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• pp.98-102
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• 2002

As the nuclear power plant has been operated continuously and increased gradually, transportation and storage of spent fuel are seriously considered nowadays. The transportation cask which contains radioactive material needs to be inspected about structural safety. About safety verification, description of IAEA Safety Standards states that cask must withstand hypothetical accident conditions. In this paper, 9m free drop impact analysis was performed for transportation cask and impact limiter by using the finite element methods. Furthermore, we obtained the dynamic behavior of wood to as compared with safety test results, and verified the safety of transportation cask.

• Nuclear Engineering and Technology
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• v.30 no.5
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• pp.462-469
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• 1998

A package to transport high-level radioactive materials is required to withstand normal transport and hypothetical accident conditions pursuant to the IAEA and domestic regulations. The package should maintain the structural safety not to release radioactive material in any condition. The structural safety of the package has been evaluated by tests using proto-type or scaled-down models, however, the method by analysis is gradually utilized due to recent advancement of computers and computer codes. In this paper, to evaluate the structural safety of a radioisotope package of the KAERI, the three dimensional impact analyses under 9m free drop and 1m puncture were performed with an explicit finite-element code, the LS-DYNA3D code. The maximum stress intensity on each part was calculated and the structural safety of the package was evaluated in accordance with the regulations.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1999.11a
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• pp.149-154
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• 1999

This paper describes a passive safety injection system with free piston Stirling pump working withabundant decay heat in the nuclear reactor during the hypothetical accident. The water column in the tube assembly connected from the hot chamber to the cold chamber in the pump oscillates periodically due to thermal volume changes of non-condensable gas in each chamber. The oscillating pressure in the water column is converted into the pumping power with a suction-and-bleed type valve assembly. In this paper a dynamic model describing the frequency of oscillation and pumping pressure is developed. It was found that the pumping pressure is a function of the temperature difference between the chambers. Also, the frequency oscillation depends on the length of the tube with water column.

• 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.2
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• pp.133-149
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• 2022

The design of a wooden impact limiter equipped to a transportation cask for radioactive materials was optimized. According to International Atomic Energy Agency Safety Standards, 9 m drop tests should be performed on the transportation cask to evaluate its structural integrity in a hypothetical accident condition. For impact resistance, the size of the impact limiter should be properly determined for the impact limiter to absorb the impact energy and reduce the impact force. Therefore, the design parameters of the impact limiter were optimized to obtain a feasible optimal design. The design feasibility criteria were investigated, and several objectives were defined to obtain various design solutions. Furthermore, a probabilistic approach was introduced considering the uncertainties included in an engineering system. The uncertainty of material properties was assumed to be a random variable, and the probabilistic feasibility, based on the stochastic approach, was evaluated using reliability. Monte Carlo simulation was used to calculate the reliability to ensure a proper safety margin under the influence of uncertainties. The proposed methodology can provide a useful approach for the preliminary design of the impact limiter prior to the detailed design stage.

• 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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.17 no.1
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• pp.18-27
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• 2021

This study aims to develop an improved evaluation technology for assessing CANDU-6 safety. For this purpose, the multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout (SBO) in a CANDU-6 plant was selected as a hypothetical event scenario and the analysis model to evaluate the plant responses was envisioned into the MARS-KS input model. The model includes logic models for controlling the pressure and inventory of the primary heat transport system (PHTS) decreasing due to the u-tubes' rupture, as well as the main features of PHTS with a simplified model for the horizontal fuel channels, the secondary heat transport system including the shell side of steam generators, feedwater and main steam line, and moderator system. A steady state condition was successfully achieved to confirm the stable convergence of the key parameters. Until the turbine trip, the fuel channels were adequately cooled by forced circulation of coolant and supply of main feedwater. However, due to the continuous reduction of PHTS pressure and inventory, the reactor and turbine were shut down and the thermal-hydraulic behaviors between intact and broken loops got asymmetric. Furthermore, as the conditions of low-flow coolant and high void fraction in the broken loop persisted, leading to degradation of decay heat removal, it was evaluated that the peak cladding temperature (PCT) exceeded the limit criteria for ensuring nuclear fuel integrity. This study is expected to provide the technical bases to the accident management strategy for transient conditions with multiple events.