• 한국압력기기공학회 논문집
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• 제14권1호
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• pp.8-14
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• 2018

Printed circuit heat exchangers (PCHEs) are widely used with an increasing demand for industrial applications. PCHEs are capable of operating at high temperatures and pressure. We consider a PCHE as a candidate intermediate heat exchanger type for a high temperature gas-cooled reactor (HTGR). For conventional application using stainless steels, design and manufacturing of PCHEs are well established. For applications to HTGR, knowledge of longitudinal conduction and deformation of channel is required to estimate design margin. This paper analyzes the effects of longitudinal conduction and deformation of channel on thermal performance using a code internally developed for design and analysis of PCHEs. The code has a capability of two dimensional simulations. Longitudinal conduction is estimated using the code. In HTGR operating condition, about ten percent of design margin is required to compensate thermal performance. The cross-sectional images of PCHE channels are obtained using an optical microscope. The images are processed with computer image process technique. We quantify the deformation of channel with dimensional parameters. It is found that the deformation has negative effect on structural integrity. The deformation enhances thermal performance when the shape of channel is straight in laminar flow regime. It reduces thermal performance in cases of a zigzag channel and turbulent flow regime.

• 대한기계학회논문집
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• 제12권5호
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• pp.1003-1015
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• 1988

본 연구에서는 상하운동 또는 좌우흔들림운동하에서의 튜브진동에 대한 실험 을 통하여 튜브지지대 인자의 영향을 고찰하고자 하였다.실험은 양단이 고정된 튜 브의 중앙에 지지대가 있는 두마디 튜브의 실험장치에서 수행되었다. 실험시 고찰된 인자들은 튜브편심율, 튜브지지대 두께, 튜브와 튜브지지대간의 간격, 튜브지지대의 위치, 튜브주파수, 선형도, 그리고 튜브거동 형태(nature of the dynamic interaction )등이다.

• 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.

• 대한기계학회논문집B
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• 제38권11호
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• pp.881-887
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• 2014

본 연구에서는 micro-gap 판형 열교환기 내부의 열유동 특성에 대한 수치해석을 수행하였다. 특히 유량 조건에 따라 열교환기의 주 채널로부터 각 micro-gap 으로의 유동분배에 대한 유동관성의 영향에 대하여 조사하였다. 열교환기 주 채널의 유동을 레이놀즈 수 100 부터 10000 까지 변화시키며 그에 따른 각 micro-gap 으로의 유동분배와 온도분포의 불균일 정도를 평가하였다. 수치해석 결과 유동분배는 유동관성에 의해 크게 영향을 받는 것으로 나타났으며, 관성 효과를 감소시킬 수 있는 헤더 설계를 통해 유동분배 불균일 정도를 줄일 수 있었다. 또한 micro-gap 을 통과한 유체의 온도분포의 불균일 정도는 주유량이 증가함에 따라 증가 후 감소 추세를 나타냈다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2002년도 학술대회지
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• pp.659-662
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• 2002

The present paper presents the experimental results for pressure drop in inclined tube bundles located in a rectangular duct. Measurements are made for pressure drop in triangular and rotated triangular tube arrays having P/d ratio of 1.6 and inclination angles of 30,45,60 and 90 degrees. The Reynolds number based on the free stream velocity and tube diameter ranges from $8{\times}10^2\;to\;6.3{\times}10^{4}$. The experimental results show that the magnitude of dimensionless pressure drop decreases significantly when the inclined angle is less than 45 degree. The measured data are compared with two existing correlations available in the literatures. The ESDU correlation agrees well with the present data far the triangular arrays. But some discrepancies are observed for the rotated triangular arrays when the inclined angles are 30 and 45 degrees. The Idel'chik correlation generally agrees well with the measured data for the rotated triangular arrays except for the inclined angle of 30 degree. The Idel'chik correlation needs modification for the triangular arrays. The modified Idel'chik correlation agrees well with the measured data within $10{\%}$. It is found that the present measured data can be applied to the evaluation and modification of previous correlations.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회A
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• pp.717-724
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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, which can produce hydrogen from water or natural gas. A primary hot gas duct (HGD) as a coaxial double-tube type cross vessel is a key component connecting the reactor pressure vessel and the intermediate heat exchanger for the VHTR. In this study, structural sizing methodology for the primary HGD with a coaxial double-tube of the VHTR that produces heat at temperatures in the order of $950^{\circ}C$ was suggested and a structural pre-sizing of it was carried out as an example.

• 대한기계학회논문집B
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• 제26권3호
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• pp.507-514
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• 2002

A computation routine, capable of performing thermal design analysis of the triple-pressure bottoming system (heat recovery steam generator and steam turbine) of combined cycle power plants, is developed. It is based on thermal analysis of the heat recovery steam generator and estimation of its size and steam turbine power. It can be applied to various parametric analyses including optimized design calculation. This paper presents analysis results for the effects on the design performance of heat exchanger arrangements at intermediate and high temperature parts as well as steam pressures. Also examined is the effect of steam sources for deaeration on design performance.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1083-1095
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• 2016

The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V&V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodiumshowed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

• 대한기계학회논문집A
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• 제33권10호
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• pp.1108-1118
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• 2009

A Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source of the order of $950^{\circ}C$ for nuclear hydrogen generation, which can produce hydrogen from water or natural gas. A primary hot gas duct (HGD) as a coaxial double-tube type cross vessel is a key component connecting a reactor pressure vessel and an intermediate heat exchanger in the VHTR. In this study, a structural sizing methodology for the primary HGD of the VHTR is suggested in order to modulate a flow-induced vibration (FIV). And as an example, a structural sizing of the horizontal HGD with a coaxial double-tube structure was carried out using the suggested method. These activities include a decision of the geometric dimensions, a selection of the material, and an evaluation of the strength of the coaxial double-tube type cross vessel components. Also in order to compare the FIV characteristics of the proposed design cases, a fluid-structure interaction (FSI) analysis was carried out using the ADINA code.

• Nuclear Engineering and Technology
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• 제56권5호
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• pp.1584-1602
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• 2024

It is vital to output the required electrical power following various task requirements when the space reactor power supply is operating in orbit. The dynamic performance of the closed Brayton cycle thermoelectric conversion system is initially studied and analyzed. Based on this, a load tracking power regulation method is developed for the liquid metal cooled space reactor power system, which takes into account the inlet temperature of the lithium on the hot side of the intermediate heat exchanger, the filling quantity of helium and xenon, and the input amount of the heat pipe radiator module. After comparing several methods, a power regulation method with fast response speed and strong system stability is obtained. Under various changes in power output, the dynamic response characteristics of the ultra-small liquid metal lithium-cooled space reactor concept scheme are analyzed. The transient operation process of 70 % load power shows that core power variation is within 30 % and core coolant temperature can operate at the set safety temperature. The second loop's helium-xenon working fluid has a 65K temperature change range and a 25 % filling quantity. The lithium at the radiator loop outlet changes by less than ±7 K, and the system's main key parameters change as expected, indicating safety. The core system uses less power during 30 % load power transient operation. According to the response characteristics of various system parameters, under low power operation conditions, the lithium working fluid temperature of the radiator circuit and the high-temperature heat pipe operation temperature are limiting conditions for low-power operation, and multiple system parameters must be coordinated to ensure that the radiator system does not condense the lithium working fluid and the heat pipe.