• Journal of Energy Engineering
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• 제22권1호
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• pp.8-16
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• 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.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2008년도 춘계학술대회논문집
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• pp.308-311
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• 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
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• 대한기계학회 2003년도 추계학술대회
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• pp.568-573
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• 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
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• 제15권3호
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• pp.239-245
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• 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.

• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
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• 한국정보통신학회 2017년도 추계학술대회
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• pp.714-715
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• 2017

Much of the interest in ocean radiation detection has been heightened as a lot of radioactivity has leaked to the ocean due to the accident at the Fukushima nuclear power plant in Japan. In the study, MCNP simulation for radiation detection in the ocean was performed. Unlike in the air, the marine environment must ensure the stability of the sensor from water depth, temperature, pressure, and salinity. In the marine environment, too much radiation is shielded. Therefore, it is an object to select a housing with a low radiation shielding ratio.

• Structural Engineering and Mechanics
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• 제50권5호
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• pp.617-633
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• 2014

Because the elevated temperature degrades the mechanical properties of materials used in containments, the global behavior of containments subjected to the internal pressure under high temperature is remarkably different from that subjected to the internal pressure only. This paper concentrates on the nonlinear finite element analyses of the nuclear power plant containment structures, and the importance for the consideration of the elevated temperature effect has been emphasized because severe accident usually accompanies internal high pressure together with a high temperature increase. In addition to the consideration of nonlinear effects in the containment structure such as the tension stiffening and bond-slip effects, the change in material properties under elevated temperature is also taken into account. This paper, accordingly, focuses on the three-dimensional nonlinear analyses with thermal effects. Upon the comparison of experiment data with numerical results for the SNL 1/4 PCCV tested by internal pressure only, three-dimensional analyses for the same structure have been performed by considering internal pressure and temperature loadings designed for two kinds of severe accidents of Saturated Station Condition (SSC) and Station Black-out Scenario (SBO). Through the difference in the structural behavior of containment structures according to the addition of temperature loading, the importance of elevated temperature effect on the ultimate resisting capacity of PCCV has been emphasized.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• 제15권4호
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• pp.391-402
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• 2017

The Korea Radioactive Waste Agency (KORAD) has developed a concrete cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. During long-term storage of spent nuclear fuel in concrete casks kept in dry conditions, the integrity of the concrete cask and spent nuclear fuel must be maintained. In addition, the radiation dose rate must not exceed the storage facility's design standards. A suitable shielding design for radiation protection must be in place for the dry storage facilities of spent nuclear fuel under normal and accident conditions. Evaluation results show that the appropriate distance to the annual dose rate of 0.25 mSv for ordinary citizens is approximately 230 m. For a $2{\times}10$ arrangement within storage facilities, rollover accidents are assumed to have occurred while transferring one additional storage cask, with the bottom of the cask facing the controlled area boundary. The dose rates of 12.81 and 1.28 mSv were calculated at 100 m and 230 m from the outermost cask in the $2{\times}10$ arrangement. Therefore, a spent nuclear fuel concrete cask and storage facilities maintain radiological safety if the distance to the appropriately assessed controlled area boundary is ensured. In the future, the results of this study will be useful for the design and operation of nuclear power plant on-site storage or intermediate storage facilities based on the spent fuel management strategy.

• Journal of Korea Society of Industrial Information Systems
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• 제25권5호
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• pp.31-39
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• 2020

According to lessons learned from an accident of Fukushima Daiichi nuclear power plant, it is advisable to make a comprehensive radiation survey by the accident phase for efficient response and risk management using diverse survey platforms. This study focuses on the technical status of environmental radiation monitoring using a UAV (Unmanned aerial vehicle) and the performance test of developed aerial survey system based on two detectors with an high energy resolution through the field application to contaminated areas. Finally, the performance of aerial survey at diverse flight heights was successfully achieved by introducing the correction factor to represent the results into ambient dose rate at 1m above the ground.

• Korean Security Journal
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• 제41호
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• pp.37-66
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• 2014

In recent years, with the development of computers and electronics, electronics and communication technology in a growing and each part is dependent on the cross-referencing makes all electronic equipment is obsolete due to direct or indirect damage EMP. Korea and the impending standoff North Korea has a considerable level of technologies related to the EMP, EMP weapons you already have or in a few years, the development of EMP weapons will complete. North Korea launched a long-range missile and conducted a nuclear test on several occasions immediately after, when I saw the high-altitude nuclear blackmail has been strengthening the outright offensive nuclear EMP attacks at any time and practical significance for the EMP will need offensive skills would improve. At this point you can predict the damage situation of Korea's security reality that satisfy the need, more than anything else to build a protective system of the EMP. The scale of the damage that unforeseen but significant military damage and socio-economic damage and fatalities when I looked into the situation which started out as a satellite communications systems and equipment to attack military and security systems and transportation, finance, national emergency system, such as the damage elsewhere. In General, there is no direct casualties reported, but EMP medical devices that rely on lethal damage to people who can show up. In addition, the State power system failure due to a power supply interruption would not have thought the damage would bring State highly dependent on domestic power generation of nuclear plants is a serious nuclear power plant accident in the event of a blackout phenomenon can lead to the plant's internal problems should see a forecast. First of all, a special expert Committee of the EMP, the demand for protective facilities and equipment and conduct an investigation, he takes fits into your budget is under strict criteria by configuring the contractors should be sifting through. He then created the Agency for verification of performance EMP protection after you have verified the performance of maintenance, maintenance, safety and security management, design and construction company organized and systematic process Guard facilities or secret communications equipment and perfect for the EMP, such as protective equipment maneuver system should take.

• Journal of the Korean Society for Nondestructive Testing
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• 제28권2호
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• pp.191-198
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• 2008

Flow accelerated corrosion (FAC) phenomenon of low alloy carbon steels in nuclear power plant has been known as one of major degradation mechanisms. It has a potential to cause nuclear pipe rupture accident which may directly impact on the plant reliability and safety. Recently, the equipotential switching direct current potential drop (ES-DCPD) method has been developed, by the present authors, as a method to monitor wall loss in a piping. This method can rapidly monitor the thinning of piping, utilizing either the wide range monitoring (WiRM) or the narrow range monitoring (NaRM) technique. WiRM is a method to monitor wide range of straight piping, whereas NaRM focuses significantly on a narrow range such as an elbow. WiRM and NaRM can improve the reliability of the current FAC screening method that is based on computer modeling on fluid flow conditions. In this paper, the measurements by ES-DCPD are performed with signal sensitivity analyses in the laboratory environment for extended period and showed the viability of ES-DCPD for real plant applications.