• 대한전기학회:학술대회논문집
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• 대한전기학회 1995년도 추계학술대회 논문집 학회본부
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• pp.172-174
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• 1995

About a half of Electric power is generated by nuclear power plants in korea. So, the stable operation of nuclear power plant is very important for suppling the essential national electric power. A S/G(Steam Generator) level control is the most difficult system in PWR(Pressurized Water Reactor) nuclear power plant. Because of the non-linear and the non-nominal response of S/G level control, it Is very difficult to control the level by automatic mode or manual mode. The goal of this study is to establish and verify a advanced control algorithm by analyzing, modelling, stability calculation, controller parameter calculation, simulation for S/G level control system.

• 대한전기학회:학술대회논문집
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• 대한전기학회 1999년도 추계학술대회 논문집 학회본부 A
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• pp.229-231
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• 1999

This paper deals with electromagnetic transient modelling of wind power system embedded in distribution networks. Wind power system consists of induction generator link reactor, distribution line, and controlled load unit. The introduction of embedded wind power system presents a new set of conditions to networks both with respect to power quantify needed to be transported and power quality such as sag swell, very short interruption, and flicker. This paper investigates the transient behavior of voltage, frequency, and load flow in wind driven induction generation system embedded in distribution networks.

• Nuclear Engineering and Technology
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• 제31권1호
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• pp.1-8
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• 1999

In this paper, we propose a new type of power control system for the rod drive coil of the CEDM of the PWR NPP in order to supply more reliable DC power The electrical modelling of the controlled rod drive coil was done by referring related documentations. The design of the proposed system is based on this electric81 model satisfying the existing specification. A high power DC-DC converter scheme is adopted utilizing the SMPS technique in the design of the proposed system. In order to show the effectiveness of the proposed system, an experimental system with the capability of 3.2 K Watt was set up for a rod with four cores and some computer simulations and experimentations were carried out. The result shows a very similar tracking performance with that of the existing system to the driving command. As a result of this, the proposed method can be applied to the power control system for the rod drive coil of the CEDM of the PWR NPP.

• 한국막학회:학술대회논문집
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• 한국막학회 1999년도 The 7th Summer Workshop of the Membrane Society of Korea
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• pp.31-34
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• 1999

Membrane technology can be applied in two ways to produce pure enantiomers. In one case, a membrane separation process can be combined with an enantiospecific reaction to obtain so-called 'enantiospecific membrane reactor'. These systems are useful to carry out asymmetric synthesis or kinetic resolution and simultaneously separate the produced enantiomer. As for general membrane reactors, the result is a were compact system with a higher conversion; in fact, removal of a product drives equilibrium-limited reactions towards completion. The other way to apply membrane technology to chiral production is the use of intrinsically enantioselective membranes that are able to distinguish between two isomers favouring preperential transport of only one isomer in absence of reaction. In This paper, the current development of chiral membrane processes will be discussed.

• Nuclear Engineering and Technology
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• 제42권4호
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• pp.365-376
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• 2010

This paper summarizes the results of a Writing Group on the Extension of CFD codes to two-phase flow safety problems, which was created by the Group for Analysis and Management of Accidents of the Nuclear Energy Agency' Committee on the Safety of Nuclear Installations (NEA-CSNI). Two-phase CFD used for safety investigations may predict small scale flow processes, which are not seen by system thermalhydraulic codes. However, the two-phase CFD models are not as mature as those in the single phase CFD and potential users need some guidance for proper application. In this paper, a classification of various modelling approaches is proposed. Then, a general multi-step methodology for using two-phase-CFD is explained, including a preliminary identification of flow processes, a model selection, and a verification and validation process. A list of 26 nuclear reactor safety issues that could benefit from investigations at the CFD scale is identified. Then, a few issues are analyzed in more detail, and a preliminary state-of-the-art is proposed and the remaining gaps in the existing approaches are identified. Finally, guidelines for users are proposed.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(2)
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• pp.1008-1013
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• 1995

The CANFLEX fuel bundle has been developed by KAERI/AECL jointly to facilitate the use of various fuel cycles in CANDU-6 reactor. As one of the design evaluations, the structural analysis of the fuel bundles by hydraulic drag force is performed to evaluate the fuel integrity in the period of the refuelling in CANDU-6. The structural integrity is evaluated by FEM modelling for the complicated bundles configuration in channel. It is noted that the present analysis method is newly developed for the structural integrity evaluation. The analysis results show that the fuel bundle is shown to keep its structural integrity during the refuelling.

• 전산구조공학
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• 제7권3호
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• pp.95-103
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• 1994

원자력발전소 부품중 안전과 관련된 구조물은 지진하중하에서 그 건전성을 유지하도록 설계되어야 한다. 그중 원자로내부구조물부품은 1차 내진분류에 속하는 것으로써 지진하중하에서의 건전성이 발전소 안전과 경제적인 관점에서 매우 중요하다. 지금까지 이러한 원자로내부구조물의 모델링에 대해서는 여러 사람들에 의해 연구되고 발표되었으나, 본 논문에서는 국내 발전소 중에서 Turn-jey base로 건설되어 이미 가동 중에 있는 영광 1&2호기의 원자로내부구조물에 대한 안전정지지진하의 거동을 Global Beam Model이라는 단순화된 모델을 이용하여 분석하였다. 이 모델의 설정을 위해서 주요부품들을 double pendulum의 보요소로 표현하였고, 이들 주요부품들의 특성해석을 범용유한 요소해석 코드인 ANSYS에 의해 구하여 이를 상부 및 하부에서 간격을 갖는 비선형 스프링으로 모델링하였다. 또한 이 비선형 스프링뿐만아니라 원자로용기와 원자로내부구조물부품들 사이의 유체동적현상을 묘사한 유체동력학적 coupling에 의해 pendulum의 보요소를 서로 연결시켜 모델링을 하였다. 가진자료인 안전정지하중은 영광 1&2호기의 원자로용기 지지부에 가해지는 응답스펙트럼을 시간이력함수로 바꾸었으며, 이 모델과 간진 하중을 가지고 비선형해석 code인 KWUSTOSS의 explicit Runge-Kutta-Gills algorithm을 이용하여 적분을 수행하므로써 안전정지지진하의 원자로 내부구조물에 대한 거동을 구하여 이 구조물의 주요부품에 대한 내진검증 및 구조물 내부에 있는 핵연료집합체의 내진 해석을 위한 입력자료를 확보할 수 있었다. 그리고 본 연구에서 사용된 Globa Beam Model의 간편성 및 효율성과 explicit Runge-Kutta-Gills algorithm에 대한 경제성을 확인할 수 있었다.파악되었 다. 그 외에도 '옥외공간이용 편리'(outdoor or recreation convenience)와 ' 이웃만족'(satisfaction with neighbors), 그리고 '주거환경 유형'(building type, building arrangement type)등도 유의한 인과적 관련을 보이므로써, 기존 문헌들이 제시하고 있는 것보다 훨씬 다양한 변수들이 다양한 경로를 통해 거주자 시각만족의 영향인자가 될 수 있는 가능성을 제시하고 있다. 가설 변수의 하나인 '길찾기의 난이 정도'(difficulty of way-finding)와 종 속변수간에 유의한 관련도가 나타나지 않은 이유로 길찾기 변수가 '시각만 족'보다는 거주자의 '안전만족'(safety)과 관련된 변수일 가능성도 아울러 지적되었다. 본 연구의 결과로부터, 주거 계획 및 설계분야 그리고 추후 관 련 연구 분야를 위한 여러 제안들이 제시되었다.에 관한 국가 규격은 국제 규격에서 저술한 바와 같이 특별히 규정된 것이 없고 VDE(Verband Deutscher Elektrotechniker: 서독전기기술 협회)와 SAE(Society of Automotive Engi- neers: 자동차 기술자 협회)에서 비교적 활발하고 Jaso(Japanese Automobile Standards Organization: 일본 자동차 표준협회)에서 많이 진행중에 있다. 본 고에서는 자동차의 전자제어에 따른 잡음 발생 요인과 전자파 간섭 관련 자동차 규격과 시험평가 방법에 대해 간단히 소개 하였다.저하에 저해요인으로서가 아니라, 인위적이던 자연적이던 간에 아들만 두면 단산하는 현행의 출산풍토하에서는 남아선호관이 오히려 출산력저하에 결정적으로 작용하고 있다고 하겠다. 태아의 성 판별을 통한 선택적 인공임신중절의 건수는 1990년 한해에

• Nuclear Engineering and Technology
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• 제27권5호
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• pp.743-752
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• 1995

본 논문은 배관파단에 대한 원자로 내부구조물의 해석시 사용되는 원자로 내부구조물과 노심의 커플(couple)된 모델에서 핵연료집합체의 grouping수에 따른 동적 응답의 영향을 조사한 것이다. 177개의 핵연료집합체를 1, 3, 5 그리고 7개의 그룹으로 나누어 모델링 하였고 그 각각에 대한 응답을 구하였다. 해석결과 원자로 내부구조물과 핵연료집합체의 배관파단에 대한 응답은 핵연료집합체의 grouping수에 거의 영향을 받지 않음을 알 수 있었다. 또한 핵연료집합체의 해석시 사용되는 상세모델에서 2개 이상의 이웃하는 핵 연료다발을 하나의 등가모델로 나타내는 방법을 연구한 결과 집합체의 1차모드 주파수와 일치하는 등가스프링을 사용하고 각 핵연료다발사이의 간격을 그대로 유지했을 때의 모델이 원래의 응답과 가장 잘 일치함을 보였다.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.973-979
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• 2024

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered. Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up. The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3367-3378
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• 2021

The design phase and safety assessment of Generation IV liquid metal-cooled fast reactors calls for the improvement of fuel pin performance codes, in particular the enhancement of their predictive capabilities towards uranium-plutonium mixed oxide fuels and stainless-steel cladding under irradiation in fast reactor environments. To this end, the current capabilities of fuel performance codes must be critically assessed against experimental data from available irradiation experiments. This work is devoted to the assessment of three European fuel performance codes, namely GERMINAL, MACROS and TRANSURANUS, against the irradiation of two fuel pins selected from the SUPERFACT-1 experimental campaign. The pins are characterized by a low enrichment (~ 2 wt.%) of minor actinides (neptunium and americium) in the fuel, and by plutonium content and cladding material in line with design choices envisaged for liquid metal-cooled Generation IV reactor fuels. The predictions of the codes are compared to several experimental measurements, allowing the identification of the current code capabilities in predicting fuel restructuring, cladding deformation, redistribution of actinides and volatile fission products. The integral assessment against experimental data is complemented by a code-to-code benchmark focused on the evolution of quantities of engineering interest over time. The benchmark analysis points out the differences in the code predictions of fuel central temperature, fuel-cladding gap width, cladding outer radius, pin internal pressure and fission gas release and suggests potential modelling development paths towards an improved description of the fuel pin behaviour in fast reactor irradiation conditions.