• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.05a
• /
• pp.581-582
• /
• 2005

• Proceedings of the Korean Nuclear Society Conference
• /
• 2004.10a
• /
• pp.1327-1328
• /
• 2004

• Proceedings of the Korean Nuclear Society Conference
• /
• 2005.05a
• /
• pp.727-728
• /
• 2005

• Journal of the Korea Concrete Institute
• /
• v.22 no.6
• /
• pp.817-824
• /
• 2010

The concrete containment structures have been widely used in nuclear power plants, LNG storage tanks, etc., due to their high safety and economic efficiency. The containment structure consists of a bottom slab, wall, ring beam and dome. The shape of the roof dome has a very significant effect on structural safety, the quantity of materials, and constructability; the thickness and curvature of the dome should therefore be determined to give the optimum design. The ring beam plays the role as supports for the dome, resulting in a minimized deformation of the wall. The main issues in designing the ring beam are the correct dimensions of the section and the prestress level. In this study, an efficient design procedure is proposed that can be used to determine an optimal shape and prestress level of the dome and ring beam. In the preliminary design stage of the procedure, the membrane theory of shells of revolution is adopted to determine several plausible alternatives which can be obtained even by hand calculation. Based on the proposed procedures, domes and ring beams of the existing domestic containment structures are analyzed and some improvements are discussed.

• Nuclear Engineering and Technology
• /
• v.47 no.1
• /
• pp.11-25
• /
• 2015

The accident at Japan's Fukushima Daiichi nuclear power plant in March 2011, caused by an earthquake and a subsequent tsunami, resulted in a failure of the power systems that are needed to cool the reactors at the plant. The accident progression in the absence of heat removal systems caused Units 1-3 to undergo fuel melting. Containment pressurization and hydrogen explosions ultimately resulted in the escape of radioactivity from reactor containments into the atmosphere and ocean. Problems in containment venting operation, leakage from primary containment boundary to the reactor building, improper functioning of standby gas treatment system (SGTS), unmitigated hydrogen accumulation in the reactor building were identified as some of the reasons those added-up in the severity of the accident. The Fukushima accident not only initiated worldwide demand for installation of adequate control and mitigation measures to minimize the potential source term to the environment but also advocated assessment of the existing mitigation systems performance behavior under a wide range of postulated accident scenarios. The uncertainty in estimating the released fraction of the radionuclides due to the Fukushima accident also underlined the need for comprehensive understanding of fission product behavior as a function of the thermal hydraulic conditions and the type of gaseous, aqueous, and solid materials available for interaction, e.g., gas components, decontamination paint, aerosols, and water pools. In the light of the Fukushima accident, additional experimental needs identified for hydrogen and fission product issues need to be investigated in an integrated and optimized way. Additionally, as more and more passive safety systems, such as passive autocatalytic recombiners and filtered containment venting systems are being retrofitted in current reactors and also planned for future reactors, identified hydrogen and fission product issues will need to be coupled with the operation of passive safety systems in phenomena oriented and coupled effects experiments. In the present paper, potential hydrogen and fission product issues raised by the Fukushima accident are discussed. The discussion focuses on hydrogen and fission product behavior inside nuclear power plant containments under severe accident conditions. The relevant experimental investigations conducted in the technical scale containment THAI (thermal hydraulics, hydrogen, aerosols, and iodine) test facility (9.2 m high, 3.2 m in diameter, and $60m^3$ volume) are discussed in the light of the Fukushima accident.

• Journal of Aerospace System Engineering
• /
• v.14 no.3
• /
• pp.60-68
• /
• 2020

If a broken blade in the aircraft engine penetrates the casing and ejects outside the aircraft, it will impact the fuselage, threatening the safety of the passengers. Thus, the development of a engine case should be certified for stability evaluation by the Aviation Administration. In this paper, we investigated the requirements and development technology for the containment certification of the engine casing necessary for the independent engine development in the country. An experimental/analytical method has been identified to summarize the contact safety requirements presented by the U.S. and European aviation agencies to verify the containment of debris in the casing corresponding to this certification. Also, we analyzed recent research on the containment casing and verification methods in casing development.

• Journal of the Korean Institute of Gas
• /
• v.12 no.1
• /
• pp.36-41
• /
• 2008

Using the finite element analysis, this paper presents the strength safety of a side wall of an outer tank and a roof structures in a full containment LNG storage tank system. The outer tank structure in which is constructed with a prestressed concrete is forced by internal hydrostatic and hydrodynamic pressures of a leaked LNG and an external wind pressure including a typhoon one. The FEM computed results show that the ring beam between a side wall of an outer tank and a roof structure supports most of the internal and the external loads. This means that the design point of the outer tank system is a ring beam structure and the other one is a center part of the roof structure. In this FE analysis model of a full containment LNG tank system, the outer tank and the roof structures are safe for the given combined loads such as an internal leaked LNG pressure and an external typhoon pressure.

• Nuclear Engineering and Technology
• /
• v.33 no.2
• /
• pp.241-253
• /
• 2001

Under severe accidents, the pressure and temperature response has an important role for the integrity of a nuclear power plant containment. The history of the pressure and temperature is characterized by the amount and state of steam/air mixture in a containment. Recently, the heat transfer rate to the structure surface is supposed to be increased by the wavy interface formed on condensate film. However, in the calculation by using CONTAIN code, the condensation heat transfer on a containment wall is calculated by assuming the smooth interface and has a tendency to be underestimated for safety. In order to obtain the best- estimate heat transfer calculation, we investigated the condensation heat transfer model in CONTAIN 1.2 code and adopted the new forced convection correlation which is considering wavy interface. By using the film tracking model in CONTAIN 1.2 code, the condensate film is treated to consider the effect of wavy interface. And also, it was carried out to investigate the effect of the different cell modelings - 5-cell and 10-cell modeling - for KNGR(Korean Next Generation Reactor) containment phenomena during a severe accident. The effect of wavy interface on condensate film appears to cause the decrease of peak temperature and pressure response . In order to obtain more adequate results, the proper cell modeling was required to consider the proper flow of steam/air mixture.

• Nuclear Engineering and Technology
• /
• v.29 no.3
• /
• pp.229-239
• /
• 1997

To investigate the realistic behavior of mass and energy release and resultant containment response during large break Loss of Coolant accident (LOCA), analyses are performed for Yonggwang (YGN) 3&4 nuclear power plants by using a merged version of RELAP5/CONTEMPT4 computer code. Comparative analyses by using conservative design computer codes are also peformed. The break types analyzed are the double-ended guillotine breaks at the cold leg and hot leg. The design analysis resulted in containment peak pressure during post-blowdown phase for the cold leg break. However, the RELAP5/CONTEMPT4 analyses show that the containment pressure has a peak during blowdown phase, thereafter it decreases monotonously without the second port-blowdown peak. For the hot leg break, revised design analysis shows much lower pressure than that reported in YGN 3&4 final safety analysis report. The RELAP5/CONTEMPT4 analysis shoos similar trend and confirmed that the bypass flow through the broken loop steam generator during post-blowdown is negligibly small compared to that of cold leg break. The low pressure and temperature predicted tv realistic analysis presented in this paper suggest that the design analysis methodology contains substantial margin and it can be improved to provide benefit in investment protection, such as, relaxing plant technical specifications and reducing containment design pressure.

• Nuclear Engineering and Technology
• /
• v.46 no.6
• /
• pp.797-802
• /
• 2014

This study investigated the thermal hydraulic issues in the Containment Filtered Venting System (CFVS) for a long operating time using the MELCOR computer code. The modeling of the CFVS, including the models for pool scrubbing and the filter, was added to the input file for the OPR-1000, and a Station Blackout (SBO) was chosen as an accident scenario. Although depressurization in the containment building as a primary objective of the CFVS was successful, the decontamination feature by scrubbing and filtering in the CFVS for a long operating time could fail by the continuous evaporation of the scrubbing solution. After the operation of the CFVS, the atmosphere temperature in the CFVS became slightly above the water saturation temperature owing to the release of an amount of steam with high temperature from the containment building to the scrubbing solution. Reduced pipe diameters at the inlet and outlet of the CFVS vessel mitigated the evaporation of scrubbing water by controlling the amount of high-temperature steam and the water saturation temperature.