• International Journal of Air-Conditioning and Refrigeration
• /
• v.11 no.2
• /
• pp.51-60
• /
• 2003

In the nuclear power plant, emergency core coolant system (ECCS) is furnished at reactor coolant system (RCS) in order to cool down high temperature water in case of emergency. However, in this coolant system, thermal stratification phenomenon can occur due to coolant leaking in the check valve. The thermal stratification produces excessive thermal stresses at the pipe wall so as to yield thermal fatigue crack (TFC) accident. In the present study, effects of turbulence penetration on the thermal stratification into T-branches with square cross-section in the modeled ECCS are analysed numerically. Standard k-$\varepsilon$ model is employed to calculate the Reynolds stresses in momentum equations. Results show that the length and strength of thermal stratification are primarily affected by the leak flow rate of coolant and the Reynolds number of duct. Turbulence penetration into the T-branch of ECCS shows two counteracting effects on the thermal stratification. Heat transport by turbulence penetration from main duct to leaking flow region may enhance thermal stratification while the turbulent diffusion may weaken it.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.6 no.1
• /
• pp.43-49
• /
• 2010

The piping failure probability of the reactor coolant system in Kori unit 1 was evaluated considering stress corrosion cracking. The P-PIE program (Probabilistic Piping Integrity Evaluation Program) developed in this study was used in the analysis. The effect of some variables such as oxygen concentration during start up and steady state operation, and operating temperature, which are related with stress corrosion cracking, on the piping failure probabilities was investigated. The effects of leak detection capability, the size of big leak, piping loops, and reactor types on the piping failure probability were also investigated. The results show that (1) LOCA (loss of coolant accident) probability of Kori unit 1 is extremely low, (2) leak probability is sensitive to oxygen concentration during steady state operation and operating temperature, while not sensitive to the oxygen concentration during start up, and (3) the piping thickness and operating temperature play important roles in the leak probabilities of the cold leg in 4 reactor types having same inner diameter.

• Journal of Electrical Engineering and Technology
• /
• v.11 no.2
• /
• pp.285-299
• /
• 2016

Although South Korea experienced a rolling blackout in 2011, the possibility of a blackout in South Korea continues to increase due to rapid electrification. This study examines the problems of energy taxation and price distortions as possible reasons for the rapid electrification in South Korea, which is occurring at a faster rate than in Japan, Europe, and other developed countries. Further, we suggest new energy taxation and price systems designed to normalize electricity prices. In order to do so, we consider two possible scenarios: the first imposes a tax on bituminous coal for electricity generation and the second levies a tax to provide compensation for the potential damages from a nuclear accident. Based on these scenarios, we analyze the effects of a new energy system on electricity price and demand. The results show that a new energy system could guarantee the power generation costs and balance the relative prices between energy sources, and could also help prevent rapid electrification. Therefore, the suggested new energy system is expected to be utilized as a basis for energy policy to decrease the speed of electrification, thus preventing a blackout, and to induce the rational consumption of energy in South Korea.

• Journal of Energy Engineering
• /
• v.22 no.1
• /
• pp.8-16
• /
• 2013

A core catcher is considered as a promising engineered system to stabilize the molten corium in the containment during a postulated severe accident in a nuclear power plant. Conceptually, the core catcher consists of a carbon steel body, sacrificial material, protection material, and engineered cooling channel. The cooling capacity of the engineered cooling channel should be guaranteed to remove the decay heat of the molten corium. The flow in ex-vessel core catcher is a combined problem of a two-phase flow in the engineered cooling channel and a single-phase natural circulation in the whole core catcher system. In this study, the analysis of the test facility for the core catcher using the CUPID code, which is a three-dimensional thermal-hydraulic code for the simulation of two-phase flows, was carried out to evaluate its cooling capacity.

• Korean Journal of Plant Resources
• /
• v.24 no.3
• /
• pp.330-336
• /
• 2011

The recent disaster of earthquake and tsunami on March 11, 2011 severely attacked East Japanese cities and people, and in addition, the accident of nuclear power station will inevitably damage the agricultural activities there. The economical depression influences on the horticulture and floriculture industry as well, through the reluctant consumption all over the country. Such a situation reflects a conventional perception that the garden plants or ornamental plants have been regarded as a symbol of capitalism, representing the success, luxury, beauty or other metaphors indicating the winners of business war. But as the word "culture" means "cultivation" originally, horti "culture" or flori "culture" should have played some roles in cultivating lands as well as cultivation of human minds, leading to develop a modern "culture" which may lay emphases on personality, originality, partnership, cooperation, diversity and so forth. In this article, a brief history of garden plants in Japan, as well as some current movements in Japanese horticulture and floriculture, is reviewed with some commodities which possess messages on creating a new humane culture.

• 한국전산유체공학회:학술대회논문집
• /
• 2008.03b
• /
• pp.308-311
• /
• 2008

As a severe accident mitigation strategy in a nuclear power plant, ERVC(External Reactor Vessel Cooling) has been proposed. Under ERVC conditions, where a molten corium is relocated in a reactor vessel lower head, a natural circulation two-phase flow is driven in the annular gap between the reactor vessel wall and its insulation. This flow should be sufficient to remove the decay heat of the molten corium and maintain the integrity of the reactor vessel. Preliminary experimental study was performed to estimate the natural circulation two-phase flow. The experimental facility which is one dimensional, the half height, and the 1/238 channel area of APR1400, was prepared and the experiments were carried out to estimate the natural circulation two-phase flow with varying the parameters of the coolant inlet area, the heat rate, and the coolant inlet subcooling. In results, the periodic circulation flow was observed and the characteristics were varied from the experimental parameters. The frequency of the natural circulation flow rate increased as the wall heat flux increased.

• Proceedings of the KSME Conference
• /
• 2003.11a
• /
• pp.568-573
• /
• 2003

When a safety valve equipped in a nuclear power plant opens in an instant by an accident, a moving shock wave propagates downstream the valve, inducing a complicated unsteady flow field. The moving shock wave may exert severe load to the structure. So, to reduce the load acting on the wall of POSRV, a gradual opening of POSRV is adopted in general. In theses connections, a numerical work is performed to investigate the effect of valve opening time on the unsteady flow fields downstream of the valve. Compressible, two-dimensional Navier-Stokes equations are used with the finite volume method. The obtained results show that sharp pressure rise through moving shock tor the case of instant opening is attenuated by employing the gradual opening of valve. It is turned that the flows for the two cases of gradual valve opening time show the similar to that of highly under-expanded one in jet structure having expansion and compression waves and Mach stem. Also, comparing with the results for the two cases of opening time, the shorter the valve opening is, the pressure gradient at the downstream of the valve becomes softly.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
• /
• v.15 no.3
• /
• pp.239-245
• /
• 2003

In the nuclear power plant, emergency core coolant system (ECCS) is furnished at reactor coolant system (RCS) in order to cool down high temperature water in case of emergency. However, in this coolant system, thermal stratification phenomenon can occur due to coolant leaking in the check valve. The thermal stratification produces excessive thermal stresses at the pipe wall so as to yield thermal fatigue crack (TFC) accident. In the present study, effects of turbulence penetration on the thermal stratification into T-branches with square cross-section in the modeled ECCS are analysed numerically. $textsc{k}$-$\varepsilon$ model is employed to calculate the Reynolds stresses in momentum equations. Results show that the length and strength of thermal stratification are primarily affected by the leak flow rate of coolant and the Reynolds number of the main flow in the duct. Turbulence penetration into the T-branch of ECCS shows two counteracting effects on the thermal stratification. Heat transport by turbulence penetration from the main duct to leaking flow region may enhance thermal stratification while the turbulent diffusion may weaken it.

• Journal of the Korea institute for structural maintenance and inspection
• /
• v.5 no.2
• /
• pp.147-154
• /
• 2001

The prestressed concrete containment is one of the most important structures in nuclear power plants, which is required to prevent release of radioactive or hazardous effluents to the environment even in the case of a severe accident. Numerical analyses are carried out by using the ABAQUS finite element program to assess the ultimate pressure capacity of the Y prestressed concrete containment with light water reactor at design criteria condition and aging condition considering varied properties of time-dependant materials respectively. From the results, it is verified that the structural capacity of the Y prestressed concrete containment building under the present, aging condition is still robust. In addition, the parameter studies for the reduction of the ultimate pressure capacity of containment building according to the degradation levels of the main structural materials are carried out. The results show that when the degradations of each materials are considered as individual and combined forms, the influence is large in the order of tendon, rebar and concrete degradation, and tendon-rebar, tendon-concrete and rebar-concrete degradation respectively.

• Journal of Institute of Control, Robotics and Systems
• /
• v.20 no.10
• /
• pp.1067-1072
• /
• 2014

The ability to overcome obstacles is necessary for field robots for various applications including the ability to climb stairs. While much research has been performed focusing on overcoming obstacles, the resulting robots do not have sufficient ability to overcome obstacles such as stairs. In this research, the purpose is to overcome relatively large obstacles by flipping locomotion through the modification of the stair climbing robotic platform of the previous research. We propose two scenarios to overcome large obstacles: a rear wheel driving system and an elevation system using a ball screw. The research is performed based on static analyses on obstacle-climbing. As the simulation results indicate, we determined the optimal posture of the robot for climbing obstacles for rear wheel driving. Also, an elevation system is analyzed for obstacle climbing. Between the two scenarios an elevation system is determined to reduce the operating torque of the actuator, and the prototype was recently assembled. The climbing ability of the robotic platform is verified. We expect the application area for this robotic platform will be in accident areas of nuclear power plants.