• Transactions of the Korean Society of Mechanical Engineers
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• v.13 no.6
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• pp.1193-1202
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• 1989

본 논문에서는 발전소의 소재로 많이 쓰이고 있는 304 스테인리스강(앞으로는 304SS로 표기함)과 316스테인리스강(앞으로는 316SS로 표기함)의 크립 균열 발생 거동 을 각각 600.deg. C와 625.deg. C에서 조사한다. 이 온도는 발전소의 반응기(reactor)에 사용 되는 304SS와 316SS이 받는 온도이다. 즉 304SS와 316SS의 크립 균열 발생을 지배 하는 파괴 매개변수가 무엇인지가 크립 파괴 실험을 통하여 조사된다. 실험 결과는 이미 제안되어 있는 크립 균열 발생 모델에서 예측된 결과와 비교된다. 특히 304SS 와 316SS은 고온에서의 연성도가 변형률 속도에 따라 변하는 것으로 알려져 있다. 본 연구에서는 '변형률 속도에 따른 재료의 연성도의 변화에 근거한 균열 발생 모델' 을 제안하고, 그 모델에서 예측된 크립 발생 거동을 실험 결과와 비교한다.

• Transactions of the Korean Society of Mechanical Engineers
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• v.16 no.11
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• pp.2111-2118
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• 1992

In this study, an analytic model is developed to predict Nusselt numbers for turbulent flow in a rod bundle. Flow channel area is divided into several element channels, and simple algebraic equations of universal velocity and temperature profiles are integrated over each element channel. The integral equations are then added to yield an analytic expression for the nusselt number of a rod bundle. The analytic model reasonably predicts the available heat transfer data.

• Journal of computational fluids engineering
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• v.21 no.3
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• pp.89-97
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• 2016

In this study, direct code coupling, in which two codes share a single flow field, was conducted using 3-dimensional high resolution thermal hydraulics code, CUPID and 1-dimensional system analysis code, MARS. This approach provide the merit to use versatile capability of MARS for nuclear power plants and 3-dimensional T/H analysis capability of CUPID. Numerical Method to directly couple CUPID and MARS was described in this paper. The straight flow and manometer flow oscillation were calculated to verify conservation of coupled CUPID/MARS code in mass, momentum, and energy. This verification calculations indicates that the CUPID/MARS is coupled appropriately in numerical aspect and the coupled code can be applied to nuclear reactor thermal hydraulics after validation against integral transient experiments.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.280-285
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• 2001

As an integral part of the probabilistic fracture mechanics analysis, stress intensity factor calculation scheme for semi-elliptical surface flaws in thin-walled cylinder has been introduced. The approximation solution utilizes the influence coefficients to calculate the stress intensity factor at the crack tip. This method has been compared with other solution methods including 3-D finite element analysis for cooldown boundary condition. The analysis results confirmed that the simplified methods provided sufficiently accurate stress intensity factor values for axial semi-elliptcal flaws on the surface of the reactor pressure vessel.

• Nuclear Engineering and Technology
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• v.32 no.2
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• pp.108-127
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• 2000

Hot leg break LBLOCA(Large Break LOCA) had a potential to be a containment maximum pressure accident in YGN3&4, which was induced from excessive conservatism in the CE analysis methodology of mass and energy release. This study conducted mass and energy release experiment for the hot leg break LBLOCA during post blowdown with an integral test facility, SNUF(Seoul National University Facility). This facility simulated YGN 3&4 with volume ratio of 1/1140 based on Ishii's three level scaling. Experiment showed that SI(Safety Injection) water refilled cold leg first and core later. SI water was vaporized in the core, which resulted in the repressurization of reactor. This increase of pressure drove the water in cold leg to flow up half height of U tubes. However, since the water was drained back soon, the release through the SG side broken section by evaporation was negligibly small. This study also provided experimental assessment of RELAP5 results by KAERI for the release through the SG side broken section.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.639-644
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• 1997

The work reported in this paper identifies the thermal-hydraulic phenomena that are expected to occur during a number of key transients in SMART(System-integrated Modular Advanced ReacTor) which is under development at KAERI. The result of this effort is based on the current design concept of SMART integral reactor. Although the design is still evolving, the preliminary phenomena Identification and Ranking Table(PIRT) has been developed based on the experts' knowledge and experience. The preliminary PIRT has been developed by consensus of KAERI expert panelists and AHP(Analytical Hierarchy Process). Preliminary PIRT developed in this paper is intended to be used to identify and integrate development areas of further experimental tests needed, thermal hydraulic models and correlations and code improvements for the safety analysis of the SMART.

• Proceedings of the KIEE Conference
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• 1999.07a
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• pp.308-311
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• 1999

A squirrel cage canned induction motor for the main coolant pump of the integral reactor, SMART was designed, manufactured and tested. The motor was first designed using the equivalent circuit theory to determine major dimensions and then finalized through finite element analyses for electromagnetic and thermal characteristics. In order to verify the design methodology, a reduced scale canned induction motor was manufactured and tested. The experimental results have shown a good agreement with the analysis results.

• The Transactions of the Korean Institute of Electrical Engineers
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• v.34 no.12
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• pp.491-497
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• 1985

An applicaton of the linear optimal control theory to the problem or load follow control in pressurized water reactors (PWR) is investigated. In order to perform the steady-state and load follow operation in PWR, a nonlinear model for the reactor and steam generator is derived and linearized at 50% rated power. Simulation tests are performed for 10% demanded load. Comparing the dynamic response of the newly developed optimal load follow controller with those of the integral error feedback controller proposed by Yang, the rise time of dynamic response of the former is about 15 seconds faster than those of the latter, thus the results indicate that the fast response of the optimal load follow controller is verified. The results of this work are directly applicable to the design of the load follow control systems for commercially operated PWRs.

• Proceedings of the Korean Information Science Society Conference
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• 2001.04a
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• pp.610-612
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• 2001

디지틀 중심의 원자로 제어시스템 설계에서 소프트웨어 안전성이 중요한 현안으로 부각되고 있다. 컴퓨터기반의 디지틀시스템으로 설계되는 일체형원자로 MMIS에 적용하기 위한 소프트웨어 개발 계획은 이러한 현안을 만족하기 위해 개발할 필요가 있다. 본 논문은 소프트웨어 개발 계획을 소프트웨어 수명주기 설정, 정형화 기법 적용, 위해서도 분석 수행, 소프트웨어 시험 방법을 제시하였다. 본 논문에서 제시된 소프트웨어 개발 계획은 고품질의 소프트웨어 생산을 보장하며, 원자력 규제기관에서 요구하는 소프트웨어 안전성 보장 계획에 대한 규제사항을 만족한다. 본 논문의 소프트웨어 개발 계획을 바탕으로 추후 구체적인 수행방법, 지침, 절아, 문서화 등의 점차적으로 개발되어 일체형원자로 MMIS 소프트웨어 개발시 적용할 예정이다.

• Proceedings of the KIEE Conference
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• 2005.07b
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• pp.948-950
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• 2005

This paper compared the test data with the loss when a conductor is exposed to the magnetic fieldof reactors after generating external magnetic field in specimen by means of an air core reactor model and the calculation of loss came from a tying the combination of FEM and integral method. It was applied to the loss measurement of transformers due to leakage flux.