• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2001년도 춘계학술대회 논문집
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• pp.811-814
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• 2001

Warpage analysis of bobbin, molded by injection molding process was performed. Concerned with a mold design, cooling system was designed based on Taguchi method, the distance between cavity wall and cooling channel was most influent factor amongst four design variables like an inlet temperature of coolant, a coolant flow rate, a diameter of cooling channel, and the distance between cavity wall and cooling channel. Optimal packing processes to reduce the warpage of molded part was analyzed based on the response surface method by considering holding pressure. Their optimal processing conditions were 9.4 seconds, 5.3 seconds, 15.2 seconds, and 85MPa, respectively.

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2415-2421
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• 2020

The axial heterogeneous void distribution in a fuel channel is a relevant and important issue during nuclear reactor analysis for LWR, especially for boiling water channel-type reactors. Variation of the coolant density in fuel channel has an effect on the neutron spectrum that will in turn have an impact on the values of absolute reactivity, the void reactivity coefficient, and the fuel isotopic compositions during irradiation. This effect is referring to as the history effect in light water reactor calculations. As the void reactivity effect is positive in RBMK type reactors, the underestimation of water density heterogeneity in 3D reactor core numerical calculations could cause an uncertainty during assessment of safe operation of nuclear reactor. Thus, this issue is analysed with different cross-section libraries which were generated with WIMS8 code at different reference water densities. The libraries were applied in single fuel model of the nodal code of QUABOX-CUBBOX/HYCA. The thermohydraulic part of HYCA allowed to simulate axial water distribution along fuel assembly model and to estimate water density history effect for RBMK type fuel.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2011년 춘계학술대회논문집
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• pp.269-274
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• 2011

A CFD analysis was performed to examine the inner channel blockage of dual-cooled fuel which has being developed for the power uprate of a pressurized water reactor (PWR). The dual-cooled fuel consists of an annular fuel pellet($UO_2$) and dual claddings as well as internal and external cooling channels. The dual-cooled annular fuel is different from a conventional solid 려el by employing an internal cooling channel for each fuel pellet as well as an external cooling channel. One of the key issues is the hypothetical event of inner channel blockage because the inner channel is an isolated flow channel without the coolant mixing between the neighboring flow channels. The inner channel blockage could cause the Departure from Nucleate Boiling (DNB) in the inner channel that eventually causes a fuel failure. This paper presents the CFD simulation of the flow through the side holes of the bottom end plug for the complete entrance blockage of the inner channel. Since the amount of coolant supply to the inner channel depends on largely the pressure loss at the side hole, the pressure loss coefficient of the side hole was estimated by the CFD analysis. The CFD prediction of the loss coefficient showed a reasonable agreement with an experimental data for the complete blockage of both the inner channel entrance and the outer channel. The CFD predictions also showed the decrease of the loss coefficient as the outer channel blockage increases.

• 한국추진공학회지
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• 제7권3호
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• pp.8-14
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• 2003

KSR-III 축소형 엔진을 원형으로 하는 8채널형 칼로리미터의 냉각성능해석을 수행하였다. 축대칭 압축성 해석을 통해서 연소실 벽으로의 열유속을 예측하였으며 이를 이용하여 3차원 냉각유로 내부의 열전달 해석을 수행하였다. 연소실 벽으로의 열유속은 문헌에서 제시하는 수준으로 확인되었으며 열전달 해석을 통하여 칼로리미터 개발과 운용에 필요한 냉각수의 압력강하, 온도상승 및 연소실벽의 최고온도를 제시하였다. 연소실 압력증가에 따른 냉각요구량을 결정하였으며 냉각수의 물성변화에 의한 냉각성능 변화를 예측하였다.

• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2017년도 제48회 춘계학술대회논문집
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• pp.643-646
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• 2017

탄화수소계열의 연료를 사용하는 액체로켓엔진에서 높은 열유속으로 인해 연소실 벽면이 구조적으로 손상 및 변형되는 것을 방지하기 위하여 연소실 벽면의 냉각은 필수적이다. 여러 가지 방법 중 연료를 냉각제로 사용 후 연소과정에 투입시키는 재생냉각 방식은 엔진 성능을 높여준다. 본 연구는 구리 단면적, 채널 내 유속, 채널에 가해지는 전류를 변화시켜 냉각제로 사용하는 케로신의 열전달 특성에 대해 알아보았다. 채널의 구리 단면적이 작을수록, 케로신의 유속이 빠를수록 대류 열전달이 빠르게 일어남을 알 수 있었다.

• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3919-3927
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• 2022

This study presents an optimal design of the coolant system of a non-refueling full-life small reactor by analyzing the space-integrated geometrical and electromagnetic variables of an extra vessel electromagnetic pump (EVEMP) for the circulation of a lead-bismuth eutectic (LBE) coolant. The EVEMP is an ideal alternative to the thermal-hydraulic system of non-refueling full-life micro reactors as it possesses no internal structures, such as impellors or sealing structures, for the transportation of LBE. Typically, the LBE passes through the annular flow channel of a reactor, is cooled by the heat exchanger, and then circulates back to the EVEMP flow channel. This thermal-hydraulic flow method is similar to natural circulation, which enhances thermal efficiency, while providing a golden time for cooling cores in the event of an emergency. When the forced circulation technology of the EVEMP was applied, the non-refueling full-life micro reactor achieve an output power of 60 MWt, which is higher than that achievable via the natural circulation method (30 MWt). Accordingly, an optimized EVEMP for Micro URANUS with a flow rate of 4196 kg/s and developed pressure of 73 kPa under a working temperature of 250 ℃ was designed.

• 한국추진공학회지
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• 제6권4호
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• pp.1-6
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• 2002

소형 연소기의 냉각 유로 설계를 위하여 열전달 및 냉각 유로의 구조해석을 수행하였다. 연소기의 고온가스에서의 열전달에 관한 2차원 해석을 수행하여 연소실 벽으로의 열유속을 예측한 다음 이를 3차원 해석을 위한 열경계 조건으로 적용하였다. 위 방법으로 예측한 열 유속은 기존의 경험식과 비교하여 검증하였으며 냉각수의 유량에 둔감한 것으로 판명되어 냉각 조건이 변화하더라도 동일한 열경계 조건을 사용할 수 있었다. 단일 냉각 유로에 대한 3차원 해석을 수행하여 연소실 벽의 최대온도 변화를 예측하였으며 이는 재사용 연소기 개발에 적용될 것이다. 냉각 유로의 정적 구조 해석을 통해 응력 분포와 구조 안전성을 예측하였다.

• Nuclear Engineering and Technology
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• 제34권4호
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• pp.286-300
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• 2002

An algorithm for the sodium boiling model has been developed for calculation of the void reactivity feedback as well as the fuel and cladding temperatures in the KALIMER core after onset of sodium boiling. Modeling of sodium boiling in liquid metal reactors using sodium as a coolant is necessary because of phenomenon difference comparing with that observed generally in light water reactor systems. The applied model to the algorithm is the multiple-bubble slug ejection model. It allows a finite number of bubbles in a channel at any time. Voiding is assumed to result from formation of bubbies that (ill the whole cross section of the coolant channel except for the liquid film left on the cladding surface. The vapor pressure, currently, is assumed to be uniform within a bubble The present study is focused on not only demonstration of the vapor bubble behavior predicted by the developed model, but also confirmation of a qualitative acceptance for the model. As a result, the model can represent important phenomena in the sodium boiling, but it is found that further effort is also needed for its completition.

• International Journal of Fluid Machinery and Systems
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• 제10권2호
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• pp.138-145
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• 2017

The problem of non-uniform inflow exists in many practical engineering applications, such as the elbow suction pipe of waterjet pump and, the channel head of steam generator which is directly connect with reactor coolant pump. Generally, pumps are identical designs and are selected based on performance under uniform inflow with the straight pipe, but actually non-uniform suction flow is induced by upstream equipment. In this paper, CFD approach was employed to analyze unsteady hydrodynamic characteristics of reactor coolant pumps with different inflows. The Reynolds-averaged Naiver-Stokes equations with the $k-{\varepsilon}$ turbulence model were solved by the computational fluid dynamics software CFX to conduct the steady and unsteady numerical simulation. The numerical results of the straight pipe and channel head were validated with experimental data for the heads at different flow coefficients. In the nominal flow rate, the head of the pump with the channel head decreases by 1.19% when compared to the straight pipe. The complicated structure of channel head induces the inlet flow non-uniform. The non-uniformity of the inflow induces the difference of vorticity distribution at the outlet of the pump. The variation law of blade to blade velocity at different flow rate and the difference of blade to blade velocity with different inflow are researched. The effects of non-uniform inflow on radial forces are absolutely different from the uniform inflow. For the radial forces at the frequency $f_R$, the corresponding amplitude of channel head are higher than the straight pipe at $1.0{\Phi}_d$ and $1.2{\Phi}_d$ flow rates, and the corresponding amplitude of channel head are lower than the straight pipe at $0.8{\Phi}_d$ flow rates.

• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1769-1785
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• 2021

The Integral Pressurized Water Reactors (IPWRs) as the innovative advanced and generation-III + reactors are under study and developments in a lot of countries. This paper is aimed at the thermal hydraulic study of the hot and average fuel sub-channel in a Generation III + IPWR by loose external coupling to the neutronic simulation. The power produced in fuel pins is calculated by the neutronic simulation via MCNPX2.6 then fuel and coolant temperature changes along fuel sub-channels evaluated by computational fluid dynamic thermal hydraulic calculation through an iterative coupling. The relative power densities along the fuel pin in hot and average fuel sub-channel are calculated in sixteen equal divisions. The highest centerline temperature of the hottest and the average fuel pin are calculated as 633 K (359.85 ℃) and 596 K (322.85 ℃), respectively. The coolant enters the sub-channel with a temperature of 557.15 K (284 ℃) and leaves the hot sub-channel and the average sub-channel with a temperature of 596 K (322.85 ℃) and 579 K (305.85 ℃), respectively. It is shown that the spacer grids result in the enhancement of turbulence kinetic energy, convection heat transfer coefficient along the fuel sub-channels so that there is an increase in heat transfer coefficient about 40%. The local fuel pin temperature reduction in the place and downstream the space grids due to heat transfer coefficient enhancement is depicted via a graph through six iterations of neutronic and thermal hydraulic coupling calculations. Working in a low fuel temperature and keeping a significant gap below the melting point of fuel, make the IPWR as a safe type of generation -III + nuclear reactor.