• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.316-321
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• 1997

The mixing vane effect on the Critical Heat Flux (CHF) is discussed with focus on the vortex now effect. In the subchannel approach, this effect is not quantified by the calculation model, but directly taken into account by the CHF correlation itself through data analysis. The vortex now effect is identified the two Westinghouse correlations, and then the CHF margin issue given rise to by the Vantage-5H design change is evaluated and discussed. It is noted that deficiency about CHF dependency on the vortex flow effect could induce an error in the Departure from Nucleate Boiling Ratio (DNBR) sensitivity Calculation.

• Nuclear Engineering and Technology
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• 제29권5호
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• pp.406-416
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• 1997

A computational method is developed for predicting the steady-state temperature field in an LMR core. Detailed core-wide coolant temperature profiles are efficiently calculated using the simplified energy equation mixing model[1] and the subchannel analysis method. The $\theta$-method is employed for discretizing the energy equations in the axial direction. The interassembly coupling is achieved by interassembly gap flow. Cladding and fuel temperatures are calculated with the one-dimensional conduction model and temperature integrals of conductivities. The accuracy of the method is tested by performing several benchmark calculations for too LMR problems. The results indicate that the accuracy is comparable to the other methods based on ENERGY model. It is also shown that the implicit scheme for the axial discretization is more efficient than the explicit scheme.

• 대한전자공학회:학술대회논문집
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• 대한전자공학회 2003년도 통신소사이어티 추계학술대회논문집
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• pp.3-6
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• 2003

Crosstalk among telephone lines in the same bundles is a major impairment in current VDSL systems. In this paper, we analysis the crosstalk effect of DSL using different frequency band upon ANSI standard VDSL systems. Simulation results show that data rate of ANSI VDSL decrease by crosstalk from another VDSL using different frequency. This study has the potential implication and benefit for ANSI standard DMT VDSL energy distribution in subchannel

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1945-1954
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• 2020

Subchannel code is one of the effective simulation tools for thermal-hydraulic analysis in nuclear reactor core. In order to reduce the computational cost and improve the calculation efficiency, empirical correlation of turbulent mixing coefficient is employed to calculate the lateral mixing velocity between adjacent subchannels. However, correlations utilized currently are often fitted from data achieved in central channel of fuel assembly, which would simply neglect the wall effects. In this paper, the CFD approach based on spectral element method is employed to predict turbulent mixing phenomena through gaps in 3 × 3 bare tight lattice rod bundle and investigate the flow pulsation through gaps in different positions. Re = 5000,10000,20500 and P/D = 1.03 and 1.06 have been covered in the simulation cases. With a well verified mesh, lateral velocities at gap center between corner channel and wall channel (W-Co), wall channel and wall channel (W-W), wall channel and center channel (W-C) as well as center channel and center channel (C-C) are collected and compared with each other. The obvious turbulent mixing distributions are presented in the different channels of rod bundle. The peak frequency values at W-Co channel could have about 40%-50% reduction comparing with the C-C channel value and the turbulent mixing coefficient β could decrease around 25%. corrections for β should be performed in subchannel code at wall channel and corner channel for a reasonable prediction result. A preliminary analysis on fluctuation at channel gap has also performed. Eddy cascade should be considered carefully in detailed analysis for fluctuating in rod bundle.

• 에너지공학
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• 제13권2호
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• pp.101-111
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• 2004

액체금속로 노심 열유체 설계의 기본 목표는 노심을 구성하는 집합체에서 발생하는 열량을 효과적으로 추출하기 위해 각각의 집합체 냉각재 유량을 적절히 분배하고, 이에 따른 온도분포가 적절하게 유지되도록 하는 것이다. 노심 열유체 설계 및 특성 분석은 전체노심에 대한 각 집합체의 유량영역을 구분하고, 집합체별 온도분포를 계산하여, 최종적으로 집합체에 대한 상세 부수로 해석을 하는 과정으로 진행된다. 본 논문에서는 이러한 액체금속로의 노심 열유체 설계 방법론을 기술하고, 이를 바탕으로 KALIMER의 증식특성 노심과 breakeven 노심에 대한 열유체 설계와 특성분석을 수행하였다. KALIMER는 원자력 중장기 과제로 개념설계가 진행 중인 전기출력 150MWe, 열출력 392MWth의 금속핵연료를 사용하는 액체 금속로이다.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3154-3165
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• 2022

During a loss-of-coolant accident (LOCA) in the pressurized water reactor (PWR), there is a possibility that high temperature and internal pressure of the fuel rods lead to ballooning of the cladding, which causes a partial blockage of flow area in a subchannel. Such flow blockage would influence the core coolant flow, thus affecting the core heat transfer during a reflooding phase and subsequent severe accident. However, most of the system analysis codes simulate the accident process based on the assumed channel blockage ratio, resulting in the fact that the simulation results are not consistent with the actual situation. This paper integrates the developed core Fuel Rod Thermal-Mechanical Behavior analysis (FRTMB) module into the self-developed severe accident analysis code ISAA. At the same time, the existing flow blockage model is improved to make it possible to simulate the change of flow distribution due to fuel rod deformation. Finally, the ISAA-FRTMB is used to simulate the QUENCH-LOCA-0 experiment to verify the correctness and effectiveness of the improved flow blockage model, and then the effect of clad ballooning on core heat transfer and subsequent parts of core degradation is analyzed.

• 설비공학논문집
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• 제29권10호
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• pp.538-550
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• 2017

In a PWR (Pressurized Water Reactor), the appropriate heat removal from the surface of fuel rod bundle is important for ensuring thermal margins and safety. Although many CFD (Computational Fluid Dynamics) software have been used to predict complex flows inside fuel assemblies with mixing vanes, there is no domestic regulatory guideline for the comprehensive evaluation of CFD software. Therefore, from the nuclear regulatory perspective, it is necessary to perform the systematic assessment and prepare the domestic regulatory guideline for checking whether valid CFD software is used for nuclear safety problems. In this study, to provide systematic evaluation and guidance on the applicability of CFD software to the domestic nuclear safety area, the results of the sensitivity analysis for the effect of the discretization scheme accuracy for the convection terms and turbulence models, which are main factors that contribute to the uncertainty in the calculation of the nuclear safety problems, on the prediction performance for the turbulent flow distribution inside the fuel assembly with split-type mixing vanes were explained.

• Nuclear Engineering and Technology
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• 제24권2호
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• pp.168-177
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• 1992

이상유동 현상의 해석은 원자력 발전소의 각계통과 가압경수형 원자로의 안전성 분석, 각종 열 수력학적 현상의 해석 그리고 타 산업체의 필요성에 의해 그 연구의 중대성이 커지고 있다. 이러한 이상유동의 현상 해석에 있어서 국부적 영역에서의 기포계수 결정은 매우 중요하다. 본 연구에서는 이러한 이상유동시 국부적 기포계수의 측정을 위하여 원자로내 부수로를 모사한 수직사각 유로를 제작하였다. 또한 국부적 영역에서의 기포계수 측정에 적합한 것으로 알려진 전기탐침 및 그 부가회로를 제작하였으며, 완성된 탐침을 이용하여 실제 비등이 발생하는 실험용 유로내에서 국부적 기포계수의 측정을 시도하였다. 실험 결과 제작된 전기탐침 및 그부가회로의 타당성을 확인 할 수 있었다.

• 설비공학논문집
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• 제28권5호
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• pp.186-194
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• 2016

As a turbulence-enhancing device, a mixing vane installed at a spacer grid of the fuel assembly plays a role in improving the convective heat transfer by generating either swirl flow in the subchannels or cross flow between fuel rod gaps. Therefore, both configuration and arrangement pattern of a mixing vane are important factors that determine the performance of a mixing vane. In this study, in order to examine the flow distribution features inside $5{\times}5$ fuel assembly with swirl-type mixing vanes used in benchmark calculation of OECD/NEA, simulations were conducted with commercial CFD software ANSYS CFX R.14. Predicted results were compared to data measured from MATiS-H (Measurement and Analysis of Turbulent Mixing in Subchannels-Horizontal) test facility. In addition, the effect of swirl-type mixing vanes on flow pattern inside the fuel assembly was described.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.534-539
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• 1997

A CHF table method has been assessed in this study for rod bundle CHF predictions. At the conceptual design stage for a new reactor, a general critical heat flux (CHF) prediction method with a wide applicable range and reasonable accuracy is essential to the thermal-hydraulic design and safety analysis. In many aspects, a CHF table method (i.e., the use of a round tube CHF table with appropriate bundle correction factors) can be a promising way to fulfill this need. So the assessment of the CHF table method has been performed with the bundle CHF data relevant to pressurized water reactors (PWRs). For comparison purposes, W-3R and EPRI-1 were also applied to the same data base. Data analysis has been conducted with the subchannel code COBRA-IV-I. The CHF table method shows the best predictions based on the direct substitution method. Improvements of the bundle correction factors, especially for the spacer grid and cold wall effects, are desirable for better predictions. Though the present assessment is somewhat limited in both fuel geometries and operating conditions, the CHF table method clearly shows potential to be a general CHF predictor.