• International Journal of Automotive Technology
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• 제7권4호
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• pp.423-439
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• 2006

This paper proposes a hybrid analytic method for the prediction of vibrational and acoustic responses of reverberant system in the medium-to-high frequency ranges by using the PFA(Power Flow Analysis) algorithm and SEA(Statistical Energy Analysis) coupling concepts. The main part of this method is the application of the coupling loss factor(CLF) of SEA to the boundary condition of PFA in reverberant system. The hybrid method developed shows much more promising results than the conventional SEA and equivalent results to the classical PFA for various damping loss factors in a wide range of frequencies. Additionally, this paper presents applied results of hybrid power flow finite element method(hybrid PFFEM) by formulating the new joint element matrix with CLF to analyze the vibrational responses of built-up structures. Finally, the analytic results of coupled plate structures and an automobile-shaped structure using hybrid PFFEM were predicted successively.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.97-109
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• 2022

Multi-dimensional coupling analysis is a research hot spot in nuclear reactor thermal hydraulic study and both the full-scale system transient response and local key three-dimensional thermal hydraulic phenomenon could be obtained simultaneously, which can achieve the balance between efficiency and accuracy in the numerical simulation of nuclear reactor. A one-dimensional to three-dimensional (1D-3D) coupling platform for the nuclear reactor multi-dimensional analysis is developed by XJTU-NuTheL (Nuclear Thermal-hydraulic Laboratory at Xi'an Jiaotong University) based on the CFD code Fluent and system code RELAP5 through the Dynamic Link Library (DLL) technology and Fluent user-defined functions (UDF). In this paper, the International Standard Problem (ISP) No. 43 is selected as the benchmark and the rapid boron dilution transient in the nuclear reactor is studied with the coupling code. The code validation is conducted first and the numerical simulation results show good agreement with the experimental data. The three-dimensional flow and temperature fields in the downcomer are analyzed in detail during the transient scenarios. The strong reverse flow is observed beneath the inlet cold leg, causing the de-borated water slug to mainly diffuse in the circumferential direction. The deviations between the experimental data and the transients predicted by the coupling code are also discussed.

• 대한설비공학회:학술대회논문집
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• 대한설비공학회 2009년도 하계학술발표대회 논문집
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• pp.1031-1036
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• 2009

This study is intended to identify how quick disconnect coupling which connects with refrigerant piping of air-conditioner using R-22 refrigerant has effect on characteristics of flux. in the case where the air-conditioner installs utilizes quick disconnect coupling, COP has an effect on the quantity of cooling load because of changing flow rate and physical properties of refrigerant which flow into an entrance of expansion valve from coupling. Variation of flow rate can be regulated by changing expansion-contraction angle; $\alpha$ of an entrance and an exit of coupling. In this study, quick disconnect coupling is presented flow of coupling by using FLUENT as heat flow program. To have an effect on the expansion entrance valve, and by changing expansion-contraction angle; $\alpha$ of an entrance and an exit

• Nuclear Engineering and Technology
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• 제53권5호
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• pp.1540-1555
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• 2021

The plate-type fuel assembly adopted in nuclear research reactor suffers from complicated effect induced by non-uniform irradiation, which might affect its stress conditions, mechanical behavior and thermal-hydraulic performance. A reliable numerical method is of great importance to reveal the complex evolution of mechanical deformation, flow redistribution and temperature field for the plate-type fuel assembly under non-uniform irradiation. This paper is the first part of a two-part study developing the numerical methodology for the thermal-fluid-structure coupling behaviors of plate-type fuel assembly under irradiation. In this paper, the thermal-fluid-structure coupling methodology has been developed for plate-type fuel assembly under non-uniform irradiation condition by exchanging thermal-hydraulic and mechanical deformation parameters between Finite Element Model (FEM) software and Computational Fluid Dynamic (CFD) software with Mesh-based parallel Code Coupling Interface (MpCCI), which has been validated with experimental results. Based on the established methodology, the effects of non-uniform irradiation and fluid were discussed, which demonstrated that the maximum mechanical deformation with irradiation was dozens of times larger than that without irradiation and the hydraulic load on fuel plates due to differential pressure played a dominant role in the mechanical deformation.

• 한국기계가공학회지
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• 제7권4호
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• pp.76-81
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• 2008

A cryogenic butterfly valve is used to transfer the liquefied natural gas (LNG) which temperature is $-162^{\circ}C$. This valve is core part in the piping system using LNG. This paper performed coupling analysis using FEM to evaluate safety of cryogenic butterfly valve. Flow analysis is calculated numerically the CAE and CFD methods are useful to predict the thermal matter and the inner flow field of the valve. Thermal analysis and structural analysis used ANSYS Workbench.

• 전기학회논문지
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• 제58권7호
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• pp.1329-1335
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• 2009

This paper presents the flow characteristics in the fluid circulation loop using the tubular type linear induction motor(TLIM) electromagnetic pump. A TLIM pump was designed using the equivalent and genetic algorithm for the flow system of 40[1/min]. The flow characteristics are analyzed by coupling the Maxwell equations with the Navier-Stokes equation. The analysis algorithm also takes account of the effects of the thrust. The flow characteristics are analysed with the proposed method and compared with the commercial program and experiment and discussed.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2775-2787
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• 2021

At present, most of the widely used system codes for nuclear safety analysis are one-dimensional, which cannot effectively simulate the flow field of the reactor core or other structures. This is true even for the system codes containing three-dimensional modules with limited three-dimensional simulation function such as RELAP-3D. In contrast, the computational fluid dynamics (CFD) codes excel at providing a detailed three-dimensional flow field of the reactor core or other components; however, the computational domain is relatively small and results in the very high computing resource consuming. Therefore, the development of coupling codes, which can make comprehensive use of the advantages of system and CFD codes, has become a research focus. In this paper, a review focus on the researches of coupled CFD and thermal-hydraulic system codes was carried out, which summarized the method of coupling, the data transfer processing between CFD and system codes, and the verification and validation (V&V) of coupled codes. Furthermore, a series of problems associated with the coupling procedure have been identified, which provide the general direction for the development and V&V efforts of coupled codes.

• 한국전산유체공학회지
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• 제21권3호
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• pp.89-97
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• 2016

In this study, direct code coupling, in which two codes share a single flow field, was conducted using 3-dimensional high resolution thermal hydraulics code, CUPID and 1-dimensional system analysis code, MARS. This approach provide the merit to use versatile capability of MARS for nuclear power plants and 3-dimensional T/H analysis capability of CUPID. Numerical Method to directly couple CUPID and MARS was described in this paper. The straight flow and manometer flow oscillation were calculated to verify conservation of coupled CUPID/MARS code in mass, momentum, and energy. This verification calculations indicates that the CUPID/MARS is coupled appropriately in numerical aspect and the coupled code can be applied to nuclear reactor thermal hydraulics after validation against integral transient experiments.

• 한국해양학회지:바다
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• 제26권2호
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• pp.96-109
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• 2021

담수를 통해 유입되는 육상 기원 물질은 연안 일차 생산을 제어하는 주요 요소이므로 육상-해양 coupling을 파악하는 것은 연안 생태 변화를 이해하는데 중요하다. 본 논문에서는 육상-해양 coupling 양상을 시간에 따라 세 단계(Base flow, Plume event, Residual flow)로 구분하여 개념화하였고, 낙동강 하구에서 각 양상의 출현을 확인하기위해 다양한 플랫폼에서 측정된 자료를 사용하여 분석하였다. 사용된 자료는 원격 탐사 측정 자료(Geostationary Ocean Color Image; GOCI), 현장 실측 자료(Marine Environment Information System; MEIS), 연속 측정 자료(유량 자료, 기상 자료)로 구분될 수 있다. 주성분분석을 통해 MEIS 자료를 육상-해양 coupling의 세 단계로 구분하였고, 이 구분을 2013-2018년 동안의 여타 자료에 적용하여 단기간 육상-해양 coupling 양상을 살펴보았다. 낙동강 하구에서는 예상과는 달리 Plume event때 Chlorophyll-a (Chl-a) 최대값이 나타났다. 이는 담수 증가에도 탁도 증가는 크지 않았고, 플러싱 효과도 약해 식물플랑크톤이 증가 할 수 있는 여건이 조성되었기 때문으로 분석되었다. 육상-해양 coupling을 기반으로 여러 하구들과 비교해보았을 때 육상-해양 coupling은 담수 유입에 영향을 받는 하구에서 흔한 현상이나 하구에서 형성되는 플룸 크기에 따라 육상-해양 coupling이 다르게 나타났다. 낙동강 하구처럼 작은 플룸(~10 km 규모) 이 형성되는 하구에서는 식물플랑크톤의 즉각적인 반응으로 인해 Plume event 단계에 Chl-a 최댓값이 나타나는 반면, ~100 km 이상의 큰 플룸이 형성되는 하구(담수 배출이 크고, 플러싱이 강한 곳)에서는 본 연구에서 개념화한 육상-해양 coupling 양상(Residual flow 때 Chl-a 최대)을 따르는 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.763-775
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• 2021

The thermal hydraulic performances of Steam Generator (SG) under both steady and transient operation conditions are of great importance for the safety and economy in nuclear power plants. In this paper, based on our self-developed SG thermal hydraulic analysis code STAF (Steam-generator Thermalhydraulic Analysis code based on Fluent), an improved new version STAF-CT (fully Coupling and Transient) is developed and introduced. Compared with original STAF, the new version code STAF-CT has two main functional improvements including "Transient" and "Fully Three Dimensional Coupling" features. In STAF-CT, a three dimensional energy transferring module is established which can achieve energy exchange computing function at the corresponding position between two sides of SG. The STAF-CT is validated against the international benchmark experiment data and the results show great agreement. Then the U-shaped SG in AP1000 nuclear power plant is modeled and simulated using STAF-CT. The results show that three dimensional flow fields in the primary side make significant effect on the energy source distribution between two sides. The development of code STAF-CT in this paper can provide an effective method for further SG high fidelity research in the nuclear reactor system.