• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.4
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• pp.355-361
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• 2014

Steam explosions can be caused by fuel-coolant interactions resulting from failure of the external vessel cooling system in a new nuclear power plant. This can threaten the integrity of structures, including the nuclear reactor and the containment building. In the present study, an improved technique for analyzing the steam explosion phenomenon was proposed on the basis of previous research and was verified by simulations involving alumina experiments. Also, the improved analysis technique was applied to determine the pressure history of the reactor cavity in accordance with postulated failure locations. The results of the analysis revealed that the effects of vessel side failure are more serious than those of vessel bottom failure, with approximately 70% higher maximum pressure.

• Fire Science and Engineering
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• v.19 no.3 s.59
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• pp.20-27
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• 2005

This paper evaluates the fire safety level of eight pump rooms in the nuclear power plant using a fire model, CFAST We estimate the Conditional Core Damage Probability (CCDP) of each room based on the analyzed results of CFAST Eight rooms located on the primary auxiliary building of the nuclear power plant are high pressure safety injection pump room A/B, low pressure safety injection pump room Am. containment sprdy pump room A/B, and motor-driven auxiliary feed water pump room A/B. The upper layer gas temperature of each room is estimated and the integrity of cable is reviewed. Based on the results, the integrity of the cable located at the upper part of compartment is maintained without thermal damage. The Conditional Core Damage Probability Is reduced to half of the old values. Accordingly, the fire safety assessment for eight pump rooms using the fire model will be capable of reducing the uncertainty and to develop a more realistic model.

• Journal of the Earthquake Engineering Society of Korea
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• v.9 no.6 s.46
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• pp.13-19
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• 2005

Floor response spectra for dynamic response of subsystem such as equipment, or piping in nuclear power plants are usually generated without considering dynamic interaction between main structure and subsystem. This study describes the analytic method in which equipment response spectra can be obtained through dynamic analysis considering equipment-structure Interaction(ESI). In this method, dynamic response of the equipment by this method is based on a dynamic substructure method in which the equipment-structure system is partitioned into the single-degree-ol-freedom system(SDOF) representing the equipment and the equipment support impedance representing the dynamic charactenstics of the structure ai the equipment support. A family of equipment response spectra is developed by applying this method to calculate the maximum responses of a family of SDOF equipment systems with wide banded equipment frequency, damping ratio, and mass. The method is validated by comparing the floor response spectrum from this method with the floor response spectrum generated from the rigorous analysis including equipments on the containment building of a prototypical nuclear power plant. in order to Investigate ESI effect in the response of equipment, response values from the method and the conventional approach without considering ESI are compared for the equipment having the mass less than 1% of the total structural mass. Response spectra from the method showed lower spectral amplitudes than those of the conventional floor response spectra around controlling frequencies.

• Journal of Advanced Marine Engineering and Technology
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• v.41 no.4
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• pp.302-309
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• 2017

If a hydrogen explosion occurs in a containment building, its multiplex defense wall may be destroyed and a large amount of radioactive material may be released. The hydrogen occurred interacting with melting fuel rods must be effectively controlled and removed. however, the countermeasures for reducing explosion risk are difficult to carry out, due owing to the various variety of accident scenarios causes and the irregularity of hydrogen distribution and behavior. In this study, We examine the guide structures while considering the ambient flows, in order to improve the efficiency of PAR the widely used Passive Autocatalytic Recombiner(PAR). We simulate the fluid behavior and the hydrogen reduction rate were simulated when a guide is attached to the two-step catalyst PAR. For an upward flow, the consisting of a height of 150mm, a gap of 0mm, and a performs $60^{\circ}$ showed the best. In contrast, for a sideways flow, a consisting of the height of 150mm, a gap of 100mm, and a performs $60^{\circ}$ showed the best in the case of side ward flow. for a downward flow, a consisting of the height of 50mm and a directly attached guide produce the best in the case of down ward flow results.

• Korean Journal of Optics and Photonics
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• v.28 no.4
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• pp.166-171
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• 2017

Hydrogen gas is a green energy sources because it features no emission of pollutants during combustion. But hydrogen gas is very dangerous, being flammable and very explosive. Hydrogen gas detection is very important for the safety of a nuclear power plant. Hydrogen gas is generated by oxidation of nuclear fuel cladding during a critical accident, and leads to serious secondary damage in the containment building. This paper discusses the development of a Raman lidar system for remote detection and measurement of hydrogen gas. A small, portable Raman lidar system was designed, and a measurement algorithm was developed to quantitatively measure hydrogen gas concentration. To verify the capability of measuring hydrogen gas with the developed Raman lidar system, experiments were carried out under daytime outdoor conditions by using a gas chamber that can adjust the hydrogen gas density. As results, our Raman lidar system is able to measure a minimum density of 0.67 vol. % hydrogen gas at a distance of 20 m.

• Korean Journal of Legislative Studies
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• v.26 no.3
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• pp.125-159
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• 2020

Amid global pandemic of covid-19, Korean government's response has drawn wide attention among social scientists as well as medical studies. The role of Korean state and civil society has attracted particular attention among others. Yet, this paper criticizes extant studies on Korean case which focus on the extensive intervention of the strong state and subjective attitude of Korean citizens in coping with covid-19. The concept of the strong state lacks social scientific specification and subjective citizens do not match with Korean realities. This article argues that Korean state's capacity in collecting and mobilizing digital data may offer better understanding for the successful responses to the pandemic. First, Korean state is the ultimate coordinator in collecting, analyzing and applying big data about the expansion of covid-19 with its huge network of dataveillance. Also, such role has been largely based upon relevant legal framework and well prepared manuals and cooperation with civic actors and companies. In other words, Korean digital dataveillance had demonstrated its transparency and cooperative governance. Second, such dataveillance capacity has deep roots in the long-term development of Korean state's big data management. Korean state has evolved about thirty years while enhancing digital data network within governments, companies and private sectors. Third, the relationship between Korean state's dataveillance and civil society can be characterized as a state centered push model. This model demonstrates highly effective governmental responses to covid-19 crisis but fall short of building social consensus in balancing individual freedom, human rights and effective containment policies. It means communitarian solidarity among citizens has not been a major factor in Korea's successful response yet.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.33 no.1
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• pp.63-72
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• 2020

The purpose of this study is to investigate the effects of the application of various numerical models and frequency contents of earthquakes on the performances of the reactor containment building (RCB) in a nuclear power plant (NPP) equipped with an advanced power reactor 1400. Two kinds of numerical models are developed to perform time-history analyses: a lumped-mass stick model (LMSM) and a full three-dimensional finite element model (3D FEM). The LMSM is constructed in SAP2000 using conventional beam elements with concentrated masses, whereas the 3D FEM is built in ANSYS using solid elements. Two groups of ground motions considering low- and high-frequency contents are applied in time-history analyses. The low-frequency motions are created by matching their response spectra with the Nuclear Regulatory Commission 1.60 design spectrum, whereas the high-frequency motions are artificially generated with a high-frequency range from 10Hz to 100Hz. Seismic responses are measured in terms of floor response spectra (FRS) at the various elevations of the RCB. The numerical results show that the FRS of the structure under low-frequency motions for two numerical models are highly matched. However, under high-frequency motions, the FRS obtained by the LMSM at a high natural frequency range are significantly different from those of the 3D FEM, and the largest difference is found at the lower elevation of the RCB. By assuming that the 3D FEM approximates responses of the structure accurately, it can be concluded that the LMSM produces a moderate discrepancy at the high-frequency range of the FRS of the RCB.