• Journal of Advanced Marine Engineering and Technology
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• v.25 no.6
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• pp.1324-1329
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• 2001

The depression of the external situation like the departure of WTO system and the plan of EEZ proclaim is forcing fishery into improving their fishing condition. By this international and domestic circumstance, development of the sea water cooling apparatus for fish hold storage is demanded sincerely. This study represents the thermal characteristics of the fish hold storage during transportation. The numerical analysis in this study is the finite volume method with the SIMPLE computational algorithm to study the seawater flow behavior in the fish hold storage. The computation were carried out with the variations of the circulating flow velocity and depth of fish hold storage. As the result of the three dimensional simulations, the mean temperature doesn't almost change by the circulating flow rate. find the mean temperature is suddenly changed by the ratio of depth of fish hold storage.

• Nuclear Engineering and Technology
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• v.42 no.3
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• pp.279-296
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• 2010

For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal hydraulics code, named CUPID, has been developed. The CUPID code adopts a two-fluid, three-field model for two-phase flows, and the governing equations were solved over unstructured grids, which are very useful for the analysis of flows in complicated geometries. To obtain numerical solutions, the semi-implicit numerical method for the REALP5 code was modified for an application to unstructured grids, and it has been further improved for enhanced accuracy and fast running. For the verification of the CUPID code, a set of conceptual problems and experiments were simulated. This paper presents the flow model, the numerical solution method, and the results of the preliminary assessment.

• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.54-67
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• 2018

There have been recent efforts to establish methods for high-fidelity and multi-physics simulation with coupled thermal-hydraulic (T/H) and neutronics codes for the entire core of a light water reactor under accident conditions. Considering the computing power necessary for a pin-by-pin analysis of the entire core, subchannel-scale T/H analysis is considered appropriate to achieve acceptable accuracy in an optimal computational time. In the present study, the applicability of in-house code CUPID of the Korea Atomic Energy Research Institute was extended to the subchannel-scale T/H analysis. CUPID is a component-scale T/H analysis code, which uses three-dimensional two-fluid models with various closure models and incorporates a highly parallelized numerical solver. In this study, key models required for a subchannel-scale T/H analysis were implemented in CUPID. Afterward, the code was validated against four subchannel experiments under unheated and heated single-phase incompressible flow conditions. Thereafter, a subchannel-scale T/H analysis of the entire core for an Advanced Power Reactor 1400 reactor core was carried out. For the high-fidelity simulation, detailed geometrical features and individual rod power distributions were considered in this demonstration. In this study, CUPID shows its capability of reproducing key phenomena in a subchannel and dealing with the subchannel-scale whole core T/H analysis.

• Proceedings of the KIEE Conference
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• 2008.07a
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• pp.223-224
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• 2008

In this paper, a comparison analysis is executed about costs and merits & demerits of medium and large air-conditioning systems which are gas absorbtion chillers and thermal energy storages. These results will be applied to B/C analysis to propose advanced load management subsidy policies in the electrical and gas field.

• Proceedings of the Korean Society for Agricultural Machinery Conference
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• 1993.10a
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• pp.1222-1229
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• 1993

An advanced design environment , which is based on adaptive and knowledge -based finite elements (INTELMESH), has been developed. Unlike other approaches, INTEMMESH incorporates the information about the object geometry as well as the boundary and loading conditions to generate an ${\alpha}$-priori finite element mesh which is more refined around the critical regions of the problem domain. INTEMMESH is designed for planar domains and axisymmetric 3-D structures of elasticity and heat transfer subjected to mechanical and thermal loading . It intelligently identifies the critical regions/points in the problem domain and utilize the new concepts of substructuring and wave propagation to choose the proper mesh size for them. INTEMMESH generates well-shaped triangular elements by applying trangulartion and Laplacian smoothing procedures. The adaptive analysis involves the intial finite elements analyze and an efficient ${\alpha}$-posteriori error analysis involves the initial finite element anal sis and an efficient ${\alpha}$-posteriori error analysis and estimation . Once a problem is defined , the system automatically builds a finite element model and analyzes the problem though automatic iterative process until the error reaches a desired level. It has been shown that the proposed approach which initiates the process with an ${\alpha}$-priori, and near optimum mesh of the object , converges to the desired accuracy in less time and at less cost. Such an advanced design/analysis environment will provide the capability for rapid product development and reducing the design cycle time and cost.

• Nuclear Engineering and Technology
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• v.37 no.1
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• pp.25-78
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• 2005

As a key and core knowledge for the design of various types of nuclear reactors, the discipline of reactor physics has been advanced continually in the past six decades and has led to a very sophisticated fabric of analysis methods and computer codes in use today. Notwithstanding, the discipline faces interesting challenges from next-generation nuclear reactors and innovative new fuel designs in the coming. After presenting a brief overview of important tasks and steps involved in the nuclear design and analysis of a reactor, this article focuses on the currently-used design and analysis methods, issues and limitations, and current activities to resolve them as follows: (1) Derivation of the multi group transport equations and the multi group diffusion equations, with representative solution methods thereof. (2) Elements of modem (now almost three decades old) diffusion nodal methods. (3) Limitations of nodal methods such as transverse integration, flux reconstruction, and analysis of UO2-MOX mixed cores. Homogenization and related issues. (4) Description of the analytic function expansion nodal (AFEN) method. (5) Ongoing efforts for three-dimensional whole-core heterogeneous transport calculations and acceleration methods. (6) Elements of spatial kinetics calculation methods and coupled neutronics and thermal-hydraulics transient analysis. (7) Identification of future research and development areas in advanced reactors and Generation-IV reactors, in particular, in very high temperature gas reactor (VHTR) cores.

• Journal of the Korean Crystal Growth and Crystal Technology
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• v.23 no.5
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• pp.235-240
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• 2013

Pure-carbon materials such as graphite, graphene, carbon nanotubes, and diamond have very high thermal conductivities. The reported thermal conductivity of graphene is in the range 3000~5000W/m-K at room temperature. Here, we developed graphene/cu foam hybrid type heat spreader to obtain higher thermal conductivity than Cu foam. Hybrid materials were characterized using optical microscopy (OM), scanning electron microscopy (SEM) and thermal conductivity measurement system; LFA (Laser Flash Analysis @ LFA 447, NETZSCH). We suggest that excellent thermal properties of graphene/cu foam hybrid structures are beneficial for all proposed electrical applications and can lead to a thermal management application.

• Journal of Welding and Joining
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• v.26 no.2
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• pp.43-49
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• 2008

Fatigue cracks due to thermal stratification or corrosion in pipelines of nuclear power plants can cause serious problems on reactor cooling system. Therefore, the development of an integrated technology including fabrication of standard specimens and their practical usage is needed to enhance the reliability of nondestructive testing. The test material was austenitic STS 304, which is used as pipelines in the Reactor Coolant System of a nuclear power plants. The best condition for fabrication of thermal fatigue cracks at the notch plate was selected using the thermal stress analysis of ANSYS. The specimen was installed from the tensile tester and underwent continuos tension loads of 51,000N. Then, after the specimen was heated to $450^{\circ}C$ for 1 minute using HF induction heater, it was cooled to $20^{\circ}C$ in 1 minute using a mixture of dry ice and water. The initial crack was generated at 17,000 cycles, 560 hours later (1cycle/2min.) and the depth of the thermal fatigue crack reached about 40% of the thickness of the specimen at 22,000 cycles. As a results of optical microscope and SEM analysis, it is confirmed that fabricated thermal fatigue cracks have the same characteristics as real fatigue cracks in nuclear power plants. The crack shape and size were identified.

• Bulletin of the Korean Chemical Society
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• v.18 no.1
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• pp.38-43
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• 1997

Poly(ethylene 2, 6-naphthalate-co-tetramethylene 2, 6-naphthalate) (PEN/PBN) copolymers were synthesized and studied by 13C NMR spectroscopy, DSC analysis and X-ray diffraction. A minimum in the melting point vs. composition curve was found at approximately 60 mol% tetramethylene 2, 6-naphthalate. The PEN/PBN copolymers were shown to be statistically random throughout the range of 1, 4-butanediol compositions. The melting point depression behavior of annealed PEN/PBN copolymers depended upon the sequence propagation probability, PS, which is suggested by indivisual crystal formation of two pure comonomers; that is, ethylene-naphthalate-ethylene, EE, and tetramethylene-naphthalate-tetramethylene, BB. However, it can be seen from the X-ray curve that the peaks of PEN/PBN copolymers appear from a crystal lattice which is governed only by the rich component between two different aliphatic units in the copolymer composition.

• Nuclear Engineering and Technology
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• v.28 no.1
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• pp.36-43
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• 1996

In this work, a PC-based thermal performance monitoring system is developed for the nuclear power plants. The system performs real-time thermal performance monitoring and diagnosis during plant operation. Specifically, a prototype for the KORI-2 nuclear power unit is developed and examined in this work. The analysis and the fault identification of the thermal cycle of a nuclear power plant is very difficult because the system structure is highly complex and the components are very much inter-related. In this study, some major diagnostic performance parameters are selected in order to represent the thermal cycle effectively and to reduce the computing time. The Fuzzy ARTMAP, a self-organizing neural network, is used to recognize the characteristic pattern change of the performance parameters in abnormal situation. By examination, this algorithm is shown to be able to detect abnormality and to identify the fault component or the change of system operation condition successfully. For the convenience of operators, a graphical user interface is also constructed in this work.