• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.2
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• pp.1-6
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• 2012

A PHE (Process Heat Exchanger) in a nuclear hydrogen system is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR (Very High Temperature gas cooled 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. Previous research on the high-temperature structural analysis of the small-scale PHE prototype had been performed only using parent material properties. In this study, high-temperature structural analysis using weld properties in weld zone was performed and the analysis results compared with the previous research.

• Journal of Power System Engineering
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• v.9 no.4
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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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.4 no.2
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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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.2
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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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.3
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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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.1
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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.

• Journal of the Korean Society for Nondestructive Testing
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• v.22 no.5
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• pp.500-507
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• 2002

By this study, on-line real-time radiometric system was developed using a 64 channels linear array of solid state detectors to measure wall thickness of insulated piping system. This system uses an Ir-192 as a gamma ray source and detector is composed of BGO scintillator and photodiode. Ir-192 gamma ray source and linear detector array mounted on a computer controlled robotic crawler. The Ir-192 gamma ray source is located on one side of the piping components and the detector array on the other side. The individual detectors of the detector array measure the intensity of the gamma rays after passing through the walls and the insulation of the piping component under measurement. The output of the detector array is amplified by amplifier and transmitted to the computer through cable. This system collects and analyses the data from the detector array in real-time as the crawler travels over the piping system. The maximum measurable length of pipe is 120cm/min. in the case of 1mm scanning interval.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.35 no.7
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• pp.717-722
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• 2011

The aim of this study is to identify the locations at which local wall thinning occurs and to determine the turbulence coefficients related to local wall thinning. Experiments and numerical analyses of the tee sections of different down-scaled piping components were performed and the results were compared. Numerical analyses of full-scale models of actual plants were performed in order to simulate the flow behaviors inside the piping components. In order to determine the relationship between the turbulence coefficients and the rate of local wall thinning, numerical analyses of the tee components in the main feedwater systems were performed. The turbulence coefficients obtained from the numerical analyses were compared with the local wear rate obtained from the measurement data. From the comparison of the results, the vertical flow velocity component (Vr) flowing to the wall after separating in the wall due to the geometrical configuration and colliding with the wall directly at an angle of some degree was analogous to the configuration of local wall thinning.

• Proceedings of the Korean Society of Propulsion Engineers Conference
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• 2007.04a
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• pp.322-325
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• 2007

This paper shows the development procedure of the low pressure LOX feed system which is used in the liquid rocket with a turbopump. Korean Air has cooperated with KARI in developing the LOX feed system to turbopump. The LOX feed system is characterized with cryogenic temperature and the thin-thickness tube for weight saving. The system in this project is composed with a main feed line and a recirculation line for the LOX temperature conditioning. Each piping system has many components, namely, bellows, filter, orifice, valves, flange and support. In this paper, system design & manufacturing, structural & thermal analyses, and component tests are explained. Finally, the system was assembled to the KARI's PTF test facility and functioned well to meet its required performance.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.20 no.1
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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.