• Nuclear Engineering and Technology
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• v.43 no.4
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• pp.375-382
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• 2011

The prospects for growth of the nuclear power industry in Korea have improved remarkably as the demand for energy increases in stride with economic development. Meanwhile, as nuclear energy development is enhanced, nuclear technology has also improved evolutionarily and innovatively in the areas of reactor design and safety measures. As nuclear technology development in Korea advances, more human resources are required. Accordingly, the need for a well-managed program of human resource development (HRD) aimed at assuring needed capacities, skills, and knowledge and maintaining valuable human resources through education and training in various nuclear-related fields has been recognized. A well-defined and object-oriented human resource development and management (HRD&M) is to be developed in order to balance between the dynamics of supply and demand of the workforce in the nuclear industry. The HRD&M schemes include a broad base of disciplines, education, sciences, and technologies within a framework of national sustainable development goals, which are generally considered to include economics, environment, and social concerns. In this study, the projection methodology considering a variety of economic, social, and environmental factors was developed. Using the developed methodology, medium- and long-term nuclear human resources projections up to 2030 were conducted in compliance with the national nuclear technology development programmes and plans.

• Nuclear Engineering and Technology
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• v.44 no.6
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• pp.697-708
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• 2012

The majority of instrumentation and control (I&C) systems in today's nuclear power plants (NPPs) are based on analog technology. Thus, most existing I&C systems now face obsolescence problems. Existing NPPs have difficulty in repairing and replacing devices and boards during maintenance because manufacturers no longer produce the analog devices and boards used in the implemented I&C systems. Therefore, existing NPPs are replacing the obsolete analog I&C systems with advanced digital systems. New NPPs are also adopting digital I&C systems because the economic efficiencies and usability of the systems are higher than the analog I&C systems. Digital I&C systems are based on two technologies: a microprocessor based system in which software programs manage the required functions and a programmable logic device (PLD) based system in which programmable logic devices, such as field programmable gate arrays, manage the required functions. PLD based systems provide higher levels of performance compared with microprocessor based systems because PLD systems can process the data in parallel while microprocessor based systems process the data sequentially. In this research, a bistable trip logic in a reactor protection system (RPS) was developed using very high speed integrated circuits hardware description language (VHDL), which is a hardware description language used in electronic design to describe the behavior of the digital system. Functional verifications were also performed in order to verify that the bistable trip logic was designed correctly and satisfied the required specifications. For the functional verification, a random testing technique was adopted to generate test inputs for the bistable trip logic.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.534-539
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• 1997

A CHF table method has been assessed in this study for rod bundle CHF predictions. At the conceptual design stage for a new reactor, a general critical heat flux (CHF) prediction method with a wide applicable range and reasonable accuracy is essential to the thermal-hydraulic design and safety analysis. In many aspects, a CHF table method (i.e., the use of a round tube CHF table with appropriate bundle correction factors) can be a promising way to fulfill this need. So the assessment of the CHF table method has been performed with the bundle CHF data relevant to pressurized water reactors (PWRs). For comparison purposes, W-3R and EPRI-1 were also applied to the same data base. Data analysis has been conducted with the subchannel code COBRA-IV-I. The CHF table method shows the best predictions based on the direct substitution method. Improvements of the bundle correction factors, especially for the spacer grid and cold wall effects, are desirable for better predictions. Though the present assessment is somewhat limited in both fuel geometries and operating conditions, the CHF table method clearly shows potential to be a general CHF predictor.

• Nuclear Engineering and Technology
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• v.30 no.4
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• pp.342-352
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• 1998

A nuclear design feasibility of soluble boron free(SBF core for the medium-sized(600MWe) PWR was investigated. The result conformed that soluble boron free operation could be performed by using current PWR proven technologies. Westinghouse advanced reactor, AP-600 was chosen as a design prototype. Design modification was applied for the assembly design with burnable poison and control rod absorber material. In order to control excess reactivity, large amount of gadolinia integral burnable poison rods were used and B4C was used as a control rod absorber material. For control of bottom shift axial power shape due to high temperature feedback in SBF core, axial zoning of burnable poison was applied to the fuel assemblies design. The combination of enrichment and rod number zoning for burnable poison could make an excess reactivity swing flat within around 1% and these also led effective control on axial power offset and peak pin power, The safety assessment of the designed core was peformed by the calculation of MTC, FTC and shutdown margin. MTC in designed SBF core was greater around 6 times than one of Ulchin unit 3&4. Utilization of enriched BIO(up to 50w1o) in B4C shutdown control rods provided enough shutdown margin as well as subcriticality at cold refueling condition.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.590-595
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• 1998

차세대 원자로(KNGR : Korea Next Generation Reactor)에는 새로운 안전개념으로서 피동형 안전주입탱크(Safety Injection Tank. SIT)의 도입을 고려하고 있는데, 이러한 피동형 유량조절기능은 안전주입탱크내의 유체기구(Fluidic device)인 볼텍스챔버(vortex chamber)에 의해 이루어진다. 볼텍스챔버는 내부에서 발생되는 와류강도에 따라 유동저항의 강도가 달라짐을 이용하여 유량을 피동적으로 조절할 수 있는 유체기구이다. 본 연구에서는 볼텍스챔버의 유동특성을 관찰하기 위하여 소규모 실험장치를 구축하고, 이를 이용하여 실험을 수행하였다. 본 연구는 두 단계로 수행되었다. 제1단계 실험에서는 볼텍스챔버의 기하학적 특성이 안전주입탱크의 안전주입수 방출특성에 미치는 영향에 대한 거시적 관점에서의 연구로서. 볼텍스챔버의 기하학적 변수(유입구 및 방출구의 직경)가 안전주입수의 방출과정에서 발생되는 SIT 내의수위 거동, 안전주입수의 방출유량 특성등에 미치는 영향에 대해 중점적으로 고찰하였다 제2단계 실험에서는 1단계 실험에서 관찰된 안전주입탱크의 여러 가지 방출특성과 볼텍스챔버 내부 유동장의 유동특성과의 관련성을 규명하기 위해 PIV (Particle Image Velocimetry)를 이용하여 볼텍스챔버의 기하학적 변수에 따른 유동장 내부의 국소 유속분포를 측정하였다.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.599-604
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• 1997

One of important inconsistencies in the six-equation model predictions has been found to be the force experienced by a single bubble placed in a convergent stream of liquid. Various sets of governing equations yield different amount of forces to hold the bubble stationary in a convergent nozzle. By using the first order potential flow theory, it is found that the six-equation model can not be used to estimate the force experienced by a deformed bubble. The theoretical value of the particle stress of a bubble in a convergent nozzle flow has been found to be a function of the Weber number when bubble distortion is allowed. This force has been calculated by using different sets of governing equations and compared with the theoretical value. It is suggested in this study that the bubble size distribution function can be used to remove the presented inconsistency by relating the interfacial variables with different moments of the bubble size distribution function. This study also shows that the inconsistencies in the thermal-hydraulic governing equation can be removed by mechanistic modeling of the phasic interface.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.3
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• pp.23-29
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• 2011

The renewed global interest in nuclear power has arisen from the need to reduce greenhouse gas emissions and to provide sufficient electricity for a growing global population before the accident at Fukushima Dai-ichi nuclear power plant in Japan. In spite of the safety issues of nuclear power plants raised by the ongoing Japanese nuclear crisis, many countries with nuclear power plants (NPPs) are still implementing license extensions of 10~20 years, and even consideration is being given to the concept of life-beyond-60, a further period of license extension from 60 to 80 years. To solving the materials aging problem is integral to its success. To evaluate the plant aging phenomena, a lot of background information such as materials and environment of the parts of the reactor and plant systems is needed by the experts. Information on degradation mechanisms is also used. In this paper, a materials degradation evaluation program called OnMDE-SYS (On-line Materials Degradation Evaluation System) is introduced. The developed program provides a variety of information on the materials and stressors as well as operational experience to the experts. It is also anticipated that the experts can perform materials degradation assessment on the web directly by referring to domestic and international information about the degradation of a nuclear power plants through OnMDE-SYS.

• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.311-318
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• 1992

During mid-loop operation of Nuclear Power Plant, to prevent the Decay Heat Removal System (DHRS) from failure due to air entrainment of free surface vortex in the piping system, a set of simulating experiments was performed. Through these experiments, a relation between the non-dimensionalized numbers, such as H/d, Froude number, Reynolds number, was found. It was also found that the perturbation of the system by the disturbance such as pump start, valve operation, etc., has a strong effect on the free surface vortex. Furthermore, from viewpoint of reactor safety, a modified inlet device which is reducer type is strongly recommended for the prevention of air entrainment into DHRS.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.2053-2058
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• 2004

The pressurized light water cooled, medium power (330 MWt) SMART (System-integrated Modular Advanced ReacTor) has been under development at KAERI for a dual purpose : seawater desalination and electricity generation. The SMART design verification phase was followed to conduct various separate effects tests and comprehensive integral effect tests. The high temperature / high pressure thermal-hydraulic test facility, VISTA(Experimental Verification by Integral Simulation of Transient and Accidents) has been constructed to simulate the SMART-P (the one fifth scaled pilot plant) by KAERI. Experimental tests have been performed to investigate the thermal-hydraulic dynamic characteristics of the primary and the secondary systems. Heat transfer characteristics and natural circulation performance of the PRHRS (Passive Residual Heat Removal System) of SMART-P were also investigated using the VISTA facility. The coolant flows steadily in the natural circulation loop which is composed of the steam generator (SG) primary side, the secondary system, and the PRHRS. The heat transfers through the PRHRS heat exchanger and ECT are sufficient enough to enable the natural circulation of the coolant.

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.2236-2241
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• 2008

The surface heat flux of nuclear fuel rod is the most important factor which can affect safety of reactor and fuel. If fuel rod surface heat flux exceeds the CHF(${\underline{C}}ritical$ ${\underline{H}}eat$ ${\underline{F}}lux$), fuel can be damaged. In case of double cooled annular fuel, which is under developing, contains two coolant channels. Therefore, a generated heat in the fuel pellet can move to inner or outer channel and heat flow direction is decided by both sides heat resistance which varied by dimension and material property change which caused by temperature and irradiation. The new program(called DUO) was developed. For the calculation of surface heat flux, a both sides convection by inner/outer coolant, s gap temperature jump and conduction in the fuel are modeled. Especially, temperature and time dependent fuel dimension and material property change are considered during the iteration. A sample calculation result shows that the DUO program has sufficient performance for annular fuel thermal hydraulics design.