• 동력기계공학회지
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• 제9권4호
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• pp.77-82
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• 2005

Bellows is a familiar component in piping systems as it provides a relatively simple means of absorbing thermal expansion and providing system flexibility. In routine piping flexibility analysis by finite element methods, bellows is usually considered to be straight pipe runs modified by an appropriate flexibility factor; maximum stresses are evaluated using a corresponding stress concentration factor. In this paper, the dynamic characteristics of bellows were investigated by Finite element methods. Using Anany program, the natural frequencies and evaluation of the reliability of bellows were also investigated.

• 한국압력기기공학회 논문집
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• 제4권2호
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• pp.1-6
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• 2008

Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source for nuclear hydrogen generation. The VHTR can produce hydrogen from heat and water by using a thermo-chemical process or from heat, water, and natural gas by steam reformer technology. A co-axial double-tube primary hot gas duct (HGD) is a key component connecting the reactor pressure vessel and the intermediate heat exchanger (IHX) for the VHTR. In this study, a preliminary design analysis for the primary HGD of the nuclear hydrogen system was carried out. These preliminary design activities include a determination of the size, a strength evaluation and an appropriate material selection. The determination of the size was undertaken based on various engineering concepts, such as a constant flow velocity model, a constant flow rate model, a constant hydraulic head model, and finally a heat balanced model.

• 한국압력기기공학회 논문집
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• 제13권2호
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• pp.31-37
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• 2017

The heat transfer function or thermal performance is the most important function of the steam generator component in nuclear power plants. The declining of thermal performance, fouling does not affect the electric power of the nuclear power plant within a certain fouling level, but it affects the output when goes beyond the governor valve wide open of the turbine. The VWO steam pressure can be predicted through the thermal performance evaluation of steam generators in the nuclear power plant. In consideration of the fouling characteristics of the steam generator, methods of the thermal performance evaluation and fouling cases are reviewed, and also the critical VWO value is estimated through the actual thermal performance evaluation. It is necessary to apply the VWO theory based on the thermal performance of the steam generators.

• 한국압력기기공학회 논문집
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• 제7권3호
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• pp.48-53
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• 2011

A PHE (Process Heat Exchanger) is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR (Very High Temperature Reactor) to the chemical reaction that yields a large quantity of hydrogen. A small-scale PHE prototype made of Hastelloy-X is being tested in a small-scale gas loop at Korea Atomic Energy Research Institute. In this study, as a part of the evaluation on the high-temperature structural integrity of the small-scale PHE prototype, we carried out macroscopic high-temperature structural analysis of the small-scale PHE prototype under the gas loop test conditions considering the pipeline stiffness.

• 한국압력기기공학회 논문집
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• 제13권1호
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• pp.48-53
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• 2017

We diagnosed the defect using the data obtained from the nuclear power plant simulator. In this paper, we diagnosed faults in the nuclear power plant system for discovery instead of the traditional single-component or device unit. We created the six fault scenarios and used a fault simulator to obtain the fault data. It was extracted pattern from acquired failure data. Neural network model was trained and simple pattern matching algorithm was applied. We presented a simulation result and confirmed that the applied algorithm works correctly.

• 비파괴검사학회지
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• 제22권5호
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• pp.500-507
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• 2002

본 연구에서는 감마선과 섬광체 및 광 다이오드로 구성된 64 채널의 선형 디텍터 어레이를 이용하여 보온재로 싸인 배관의 두께를 실시간으로 측정하는 시스템을 개발하였다. 본 측정시스템은 감마선원으로 Ir-192를 사용하였으며, 디텍터는 BGO 섬광체와 광다이오드로 구성하였다. Ir-192 방사선원은 배관의 한쪽 편에, 그리고 디텍터 어레이는 배관을 중심으로 그 반대편에 위치하며 컴퓨터로 제어되는 주행 시스템에 실려 배관을 따라 이송되는 동안 배관과 단열재를 투과한 방사선의 강도는 각 디텍터에서 측정된다. 측정된 디텍터 어레이의 출력은 증폭기에서 증폭된 후 케이블에 의해 컴퓨터로 전송되며 주행시스템이 진행하는 동안 컴퓨터는 수집된 신호를 분석 및 계산하여 실제의 두께를 나타내며 주사간격을 1mm로 할 경우의 최대 측정속도는 분당 120cm이다.

• 대한기계학회논문집B
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• 제35권7호
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• pp.717-722
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• 2011

본 연구의 목적은 국부감육이 일어나고 있는 위치들을 분석하고, 그와 관련된 난류매개변수를 밝혀내는데 있다. 축소 제작된 배관계 티부분에서의 실험과 수치해석이 이루어졌으며, 실제로 배관계 부품내에서의 유동특성을 유추하기 위하여 그 결과들이 비교 검토되었다. 국부감육율과 난류 매개변수간의 상관관계를 결정하기 위하여 급수가열기 주 배관에서의 티 부품에 대한 수치해석이 수행되었고, 실제적인 국부감육 발생 위치를 찾아내기 위해 알칼리 금속염을 사용하여 감육 유로가시화 실험을 수행하였으며, 이를 기초로 한 난류매개 변수와 국부감육의 두께가 비교 분석되었다. 이러한 결과 값 비교를 통하여 얻어낸 바로는 기하학적 형태에 기인하는 배관 벽면에서의 박리로 인한 반경 방향 유속 Vr이 국부 감육 현상과 가장 연관성이 높은 것으로 나타났다.

• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2007년도 제28회 춘계학술대회논문집
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• pp.322-325
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• 2007

본 논문은 터보펌프를 사용하는 액체로켓의 저압 LOX 공급계의 개발 프로세스를 제공한다. 대한항공은 한국항공우주연구원과 협력하여 터보펌프 공급을 위한 LOX 공급계 개발을 수행하였다. LOX 공급계는 극저온의 온도와 무게절감을 위한 얇은 배관두께가 특징이다. 본 프로젝트의 시스템은 주 배관과 LOX 온도 제어를 위한 재순환 배관으로 구성되어 있다. 각 배관시스템은 벨로우즈, 필터, 오리피스, 밸브류, 플랜지와 서포트로 구성되어 있다. 이 논문에서는 시스템 설계 및 제작, 구조 및 열 해석, 단품 시험에 대하여 설명하였다. 최종적으로, 이 시스템은 한국항공우주연구원의 PTF 시험설비에 조립되어 요구 성능을 달성하였다.

• 한국압력기기공학회 논문집
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• 제20권1호
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• pp.1-6
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• 2024

Crud is a corrosion deposit that forms in equipments and piping of nuclear reactor's primary systems. When crud circulates through the reactor's primary system coolant and adheres to the surface of the nuclear fuel cladding tube, it can lead to the Axial Offset Anomaly (AOA) phenomenon. This occurrence is known to potentially reduce the output of a nuclear power plant or to necessitate an early shutdown. Consequently, worldwide nuclear power plants have employed ultrasonic cleaning methods since 2000 to mitigate crud deposition, ensuring stable operation and economic efficiency. This paper details the system configuration of ultrasonic nuclear fuel cleaning equipment, outlining the function of each component. The objective is to contribute to the local domestic production of ultrasonic nuclear fuel cleaning equipment. Additionally, the paper introduces a method for accurately measuring the weight of removed crud, a crucial factor in assessing cleaning effectiveness and providing input data for the BOA code used in core safety evaluations. Accurate measurement of highly radioactive filters containing crud is essential, and weighing them underwater is a common practice. However, the buoyancy effect during underwater weighing may lead to an overestimation of the collected crud's weight. To address this issue, the paper proposes a formula correcting for buoyancy errors, enhancing measurement accuracy. This improved weight measurement method, accounting for buoyancy effects in water, is expected to facilitate the quantitative assessment of filter weights generated during chemical decontamination and system operations in nuclear power plants.

• 한국압력기기공학회 논문집
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• 제7권2호
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• pp.1-6
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• 2011

PHE(Process Heat Exchanger) is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR(Very High Temperature Reactor) to the chemical reaction that yields a large quantity of hydrogen. Korea Atomic Energy Research Institute established a small-scale gas loop for the performance test of components, which are used in the VHTR, and they manufactured a PHE prototype made of Hastelloy-X to be tested in the small-scale gas loop. Results from the elastic structural analysis on the PHE prototype were reported in the previous article. In order to investigate the macroscopic structural characteristics and behavior of the PHE prototype under the test condition of the small-scale gas loop far more in detail, elastic-plastic high-temperature structural-analysis of the PHE prototype was carried out in this study.