• Journal of Advanced Marine Engineering and Technology
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• v.40 no.3
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• pp.165-173
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• 2016

On March 11, 2011, a massive earthquake measuring 9.0 on the Richter scale and subsequent 10-.14 m waves struck the Fukushima Daiichi (FD) Nuclear Power Plant. The main and backup electric power was damaged preventing the cooling system from functioning. Fuel rods overheated and led to hydrogen explosions. If heat in the fuel rods is not dissipated, the nuclear fuel coating material (e.g., Zircaloy) reacts with water vapor to generate hydrogen at high temperatures. This hydrogen is released into the containment area. If the released hydrogen burns, the stability of the containment area is significantly impacted. In this study, researchers performed an explosion analysis in a high-risk explosion area, analyzing the hydrogen distribution in a containment building [1] and the effects of a hydrogen explosion on containment safety. Results indicated that a hydrogen explosion was possible throughout the containment building except the middle area. If an explosion occurs at the top of the containment building with more than 40% of the hydrogen collected or in the bottom right or left side of the of containment building, safety of the containment building could be threatened.

• Proceedings of the Korea Concrete Institute Conference
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• 2000.04a
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• pp.805-810
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• 2000

The purpose of this study is to predict long-term structural safety on the Yonggwang Unit 3 prestressed concrete containment building. The aging-related degradations of its main structural materials are investigated and the effects of the property variation of time-dependent materials on the structural behavior of containment building are also assessed through the analysis on the ultimate pressure capacity. The nonlinear finite element analyses for both the design criteria condition a the present aging condition are conducted to assess the present structural capacity of the containment building As a result, it is verified that the structural capacity of the Yonggwang Unit 3 containment building under the present aging condition is judged to be still rugged. n addition, the sensitivity of the ultimate pressrue capacity of containment building according to th degradation levels of the structural materials are assessed. Finally, it is showed that the sensitivity levels are in the order of the tendon, rebar and concrete in case of individual material degradations, and the tendon-rebar, tendon-concrete and rebar-concrete in case of coupled material degradations.

• Journal of ILASS-Korea
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• v.4 no.3
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• pp.42-52
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• 1999

An effectiveness of thermal utilization of a dousing system in the 600 MW PHWR Nuclear Power Plant has been evaluated. The behavior and conditions of water droplet sprayed in a postulated accident conditions in containment configuration has been calculated. In this calculation, two pressure conditions with the consideration of obstruction area and containment wall effect has been established : one being the minimum containment pressure of 7 kPa(g) encountered for dousing shut off and the other being the containment design pressure 124 kPa(g). The results revealed that the effectiveness of the thermal utilization ranges from 93% to 97%. In the analysis on two cases without/with side wall effect in the containment building, the thermal utilization decreases with obstruction area from 89% to 85%, which satisfies the design criteria set for the containment pressure against the accident condition.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.597-604
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• 2013

Containment Filtered Vent Systems (CFVSs) have been mainly equipped in nuclear power plants in Europe and Canada for the controlled depressurization of the containment atmosphere under severe accident conditions. This is to keep the containment integrity against overpressure during the course of a severe accident, in which the radioactive gas-steam mixture from the containment is discharged into a system designed to remove the radionuclides. In Korea, a CFVS was first introduced in the Wolsong unit-1 nuclear power plant as a mitigation measure to deal with the threat of over pressurization, following post-Fukushima action items. In this paper, the overall features of a CFVS installation such as risk assessments, an evaluation of the performance requirements, and a determination of the optimal operating strategies are analyzed for the Wolsong unit 1 nuclear power plant using a severe accident analysis computer code, ISAAC.

• Proceedings of the Korea Concrete Institute Conference
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• 2004.11a
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• pp.643-646
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• 2004

The posttensioning system of nuclear containment have to be verified its structural integrity by the periodic inspection because the structural behavior of the containment is changed by the variation of the physical property of concrete and tendon as time passes. In this study a program 'SAPONC-CANDU' which is able to monitor and analysis the micro structural behavior of the domestic CANDU type containment at all times was developed. The readings of vibrating-wire strain gauges embedded into the concrete of containment were used as input data for operating the program. This program provides the long-term prediction values and bands of the concrete strain due to the time dependent factors of the concrete and tendon of the domestic CANDU type containment.

• Nuclear Engineering and Technology
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• v.15 no.2
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• pp.123-134
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• 1983

Estimates of the rate of heat removal from the containment atmosphere following a loss of colant accident (LOCA) are important to the prediction of containment peak pressure and temperature which are essential parameters in designing the containment building. An overall survey and discussion of mechanisms, experimental results, empirical correlations and analytical models that are relevant to the heat transfer inside the containment have been made. As a result of this review, the current state of the knowledge about tile containment heat transfer can be understood and it is known that more investigations are needed to avoid the misuse of various correlations.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1781-1788
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• 2022

The containment building Kori Unit 1 may require sequential steps for full decommissioning. This study assumes that the containment building is to be used as an auxiliary building that handles nuclear power systems and materials during decommissioning before conversion into a greenfield. Through the derivation of guidelines and dose evaluation, it was confirmed whether the radiation workers were satisfied with the ALARA decision. The specific modeling of the external radiation exposure was performed based on the facility investigation procedures. The external radiation specific derived concentration guideline levels (DCGLs) for radiation workers in containment building were obtained using the RESRAD-BUILD code and were applied to the VISIPLAN 3D ALARA Planning Tool code to calculate the working dose and check worker safety. The derivation of site-specific and realistic DCGLs and dose evaluation via 3D modeling can contribute to the scenario development for the decommission and remediation of containment building.

• Journal of the Korean GEO-environmental Society
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• v.11 no.8
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• pp.57-63
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• 2010

For optimum scale design of containment area, a series of laboratory tests using column were performed in this study as followings; sedimentation test and self-weight consolidation test for dredged soil, and suspended solid concentration test for supernatant. Containment area has been designed and evaluated, based on field condition and concentration of suspended solid of effluent water. In addition, the relation between width of containment area and target concentration of suspended solid was analyzed. The results show that concentration of suspended solid decreases as the width of containment area decreases and the length of containment area increases. It was also observed that influence of change in ponding depth should be considered to predict the change in suspended solid concentration in supernatant discharged as disposal is conducted; the lower target suspended solid concentration of effluent water, the more important.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.25 no.2
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• pp.79-84
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• 2013

In this study, a cold aisle containment system was proposed among various strategies to reduce the energy waste by recirculation air from the hot aisle. To verify the effectiveness of the cold aisle containment system, a test bed which is similar to an actually existing server room was set up in the Internet Data Center(IDC) building. Comparative experiments, conventional open type cooling system and cold aisle containment system were carried out under actual conditions. The result revealed that the range of inlet temperature of the server system was $20{\sim}25^{\circ}C$ in an existing cooling system and the range of inlet temperature dropped below $20^{\circ}C$ by the cold aisle containment system. After all, cold aisle containment system was proved to be the solution for energy saving cooling system.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.673-682
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• 2018

One of the most widely used methods to estimate core damage during a nuclear power plant accident is containment radiation measurement. The evolution of severe accidents is extremely complex, leading to uncertainty in the containment dose rate (CDR). Therefore, it is difficult to accurately determine core damage. This study proposes to conduct uncertainty analysis of CDR for core damage assessment. First, based on source term estimation, the Monte Carlo (MC) and point-kernel integration methods were used to estimate the probability density function of the CDR under different extents of core damage in accident scenarios with late containment failure. Second, the results were verified by comparing the results of both methods. The point-kernel integration method results were more dispersed than the MC results, and the MC method was used for both quantitative and qualitative analyses. Quantitative analysis indicated a linear relationship, rather than the expected proportional relationship, between the CDR and core damage fraction. The CDR distribution obeyed a logarithmic normal distribution in accidents with a small break in containment, but not in accidents with a large break in containment. A possible application of our analysis is a real-time core damage estimation program based on the CDR.