• Proceedings of the KSME Conference
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• 2000.11a
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• pp.837-842
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• 2000

In the coiling process of helical steam generator tubes of integral reactor SMART, a considerable amount of spring back, which induces dimensional inaccuracy and difficulty in fabrication, has been arised. In this research, an analytical model was derived to evaluate the amount of the spring back for steam generator tubes. The model was developed on the basis of beam theory and elastic-perfectly plastic material property. This model was extended to consider the effect of plastic hardening and the effect of the tensile force on the spring back phenomena. Parametric studies were performed for various design variables of steam generator tubes in order to minimize the spring back in the design stage. A sensitivity analysis has shown that the low yield strength, the high elastic modulus, the small helix diameter, and the large tube diameter result in a small amount of the spring back. The amount of the spring back can be controlled by the selection of adequate design values in the basic design stage and reduced to an allowable limit by the application of the tensile force to the tube during the coiling process.

• Proceedings of the Korean Operations and Management Science Society Conference
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• 2004.05a
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• pp.5-8
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• 2004

컴퓨터기반 운전지원시스템은 운전원의 발전소 상태감시와 진단기능을 수행을 지원하여 운전원의 안전성을 향상시키는 시스템으로 80년대 중반부터 개발되어왔고, 또한 최근의 컴퓨터 기술과 도구의 빠른 발전으로 컴퓨터기반 운전지원 시스템의 개발 및 원전 적용에 대한 연구가 이루어지고 있다. 본 논문에서는 컴퓨터기반 운전원 조언 시스템의 기술 현황을 분석하고, 현재 설계를 진행중인 일체형원자로(SMART, System-integrated Modular Advanced ReacTor)에 컴퓨터기반 운전지원 시스템 적용을 위한 개발방안을 제안한다.

• 한국전산유체공학회:학술대회논문집
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• 2001.10a
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• pp.118-125
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• 2001

The performance prediction of SMART MCP was performed using a computational fluid dynamics code. General capacity-head performance curve of MCP, which is provided to other design branches as design input, was obtained and it showed the typical type of axial pump performance curve. When four MCPs operate in parallel and one of them stops while the others continue to operate, SMART requires reduced power operation. A procedure for predicting the performance of SMART MCP for that case was developed and verified with available experimental data. An analysis based on the developed procedure was performed for two cases; the impeller of sloped MCP is fixed or free to rotate in reverse direction. According to the results, $73\%$ flow rate of normal operation enters the reactor core in the case of the locked impeller. In case of the impeller free rotation, the flow rate entering the reactor core is $62.8\%$.

• The KSFM Journal of Fluid Machinery
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• v.14 no.2
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• pp.25-28
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• 2011

The control rod assemblies do freely fall into the reactor core by the gravity from the control rod drive mechanism. In order to achieve a rapid shutdown and control the reactor power, it is required to insert control rod assemblies as soon as possible. In this paper, we evaluated the drop time and flow characteristics caused around guide tube for SMART(System-integrated modular advanced reactor) control rod assembly. Numerical analyses are carried out with FLUENT program of computational fluid dynamics. This study results show that the drop time of the control rod assembly in the operating condition of SMART is more 20 percent rapidly than the drop time of the room temperature and ambient atmosphere condition.

• 한국전산유체공학회:학술대회논문집
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• 1998.05a
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• pp.75-81
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• 1998

A new advanced integral reactor of 330 MWt capacity named SMART(System-integrated Modular Advanced ReacTor) is currently under development at KAERI(Korea Atomic Energy Research Institute). One of the major design features of the integral reactor is locating the steam generators(SG) inside reactor vessel and eliminating the possibility of LB LOCA(large Break Loss of Coolant Accident). Orifices are fitted at the low part of steam generator cassette to stabilize and balance coolant flow distribution in the MCP (Main Circulation Pump) pressure header. A sensitivity analysis is conducted to determine the optimum orifice size using computer code 'CFX'.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.441-446
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• 2001

In the integral type reactor, SMART, all the major components such as steam generators, pressurizer and pumps are located inside the single reactor pressure vessel. The objective of this study is to evaluate the structural integrity for RPV of SMART under the postulated pressurized thermal shock by applying the finite element analysis. Input data for the finite element analysis were generated using the commercial code I-DEAS, and the fracture mechanics analysis was performed using the ABAQUS. The crack configurations, the crack aspect ratio and the clad thickness were considered in the parametric study. The effects of these parameters on the reference nil-ductility transition temperature were also investigated.

• The Transactions of the Korean Institute of Electrical Engineers A
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• v.52 no.4
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• pp.147-147
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• 2003

This paper describes the finite element analysis(FEA) for the design of electromagnet for Control Element Drive Mechanism(CEDM) in System-integrated Modular Advanced Reactor(SMART) and compared with the lifting power characteristics of prototype electromagnet. A thermal analysis was performed for the electromagnet. A model for the thermal analysis of the electromagnet was developed and theoretical bases for the model were established. It is important that the temperature of the electromagnet windings be maintained within the allowable limit of the insulation. since the electromagnet of CEDM is always supplied with current during the reactor operation. So the thermal analysis of the winding insulation which is composed of polyimide and air were performed by finite element method. As a result, it is shown that the characteristics of prototype electromagnet have a good agreement with the results of FEA. The thermal properties obtained here will be used as input for the optimization analysis of the electromagnet.

• Proceedings of the Korean Information Science Society Conference
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• 2004.04b
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• pp.394-396
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• 2004

원자력 산업계에서는 원전 MMIS(Man-Machine Interface System)의 디지털 기술 적용을 위해 많은 노력을 기울이고 있고, 특히 원자력 산업의 특수성인 안전성 확보에 필요한 개발기준과 규제방법 정립에 많은 연구가 수행되고 있다. 또한 디지털 MMIS의 핵심기반기술인 고 신뢰도 소프트웨어 개발 방법론이 확립되지 못하여 소프트웨어 공통모드고장 문제, 정량적인 소프트웨어 신뢰도 보장 문제 등이 현안으로 제기되고 있다. 이와 같이 원전 MMIS의 디지털화를 성공하기 위해서는 소프트웨어의 고 신뢰도 확보가 관건이며, 고 신뢰도를 확보하기 위한 소프트웨어 개발 방법론의 정립이 절실히 요구되고 있다. 본 논문에서는 원전 소프트웨어 개발에 적응되는 규제지침을 분석하고, 일체형원자로(SMART, System-integrated Modular Advanced ReacTor) MMIS 소프트웨어 개발에 적응될 소프트웨어 개발 방법론을 제시한다.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2000.11a
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• pp.163-171
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• 2000

AMBIDEXTER 원자력 에너지 시스템은 Th$^{233}$ /U 핵주기를 이용한 용융염 핵연료가 내장형 열교환기를 포함하는 일체형 원자로 시스템을 순환하면서 1차 냉각 계통을 이루고, 독립된 온라인 정화계통에 의해 액상 용융염 핵연료 일부를 연속 추출. 처리, 재주입 함으로 노심의 핵적 자활상태를 유지한다. 이와 같은 시스템 개념은 배관망 파손에 의한 중대사고 방지, 열수송 회로와 방사성 물질 회로의 독립적 구성을 통한 효과적인 원자력 에너지 이용과 고유 안전성을 확보하는 장점을 통해 현안 원자력 문제의 근본적인 해결 방안을 제시하고 있다.(중략)

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1999.05a
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• pp.119-126
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• 1999

농축우라늄 고체핵연료를 사용하는 기존의 발전용 원자로 개념에서는 냉각기능의 상실 또는 반응도 상실사고와 같은 극심한 열적 불균형에 의해 핵연료의 온도가 급격히 증가하고, 결과적으로 핵연료의 파손 및 용융으로 발전할 수 있다. 본 연구는 이러한 기존 발전로의 고유 안정성 문제를 획기적으로 해결할 수 있는 혁신형으로서 Th/$^{233}$ U 용융염핵연료주기를 사용하며 원자로계통 전체를 원자로용기에 내장하는 일체형 원자로개념의 AMBIDEXTER (Advanced Molten-salt Break-even Inherently-safe Dual-mission Experimental and TEst Reactor) 원자력 에너지시스템의 동특성을 해석하기 위해 수행되고 있다.(중략)