• Nuclear Engineering and Technology
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• 제50권1호
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• pp.54-67
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• 2018

There have been recent efforts to establish methods for high-fidelity and multi-physics simulation with coupled thermal-hydraulic (T/H) and neutronics codes for the entire core of a light water reactor under accident conditions. Considering the computing power necessary for a pin-by-pin analysis of the entire core, subchannel-scale T/H analysis is considered appropriate to achieve acceptable accuracy in an optimal computational time. In the present study, the applicability of in-house code CUPID of the Korea Atomic Energy Research Institute was extended to the subchannel-scale T/H analysis. CUPID is a component-scale T/H analysis code, which uses three-dimensional two-fluid models with various closure models and incorporates a highly parallelized numerical solver. In this study, key models required for a subchannel-scale T/H analysis were implemented in CUPID. Afterward, the code was validated against four subchannel experiments under unheated and heated single-phase incompressible flow conditions. Thereafter, a subchannel-scale T/H analysis of the entire core for an Advanced Power Reactor 1400 reactor core was carried out. For the high-fidelity simulation, detailed geometrical features and individual rod power distributions were considered in this demonstration. In this study, CUPID shows its capability of reproducing key phenomena in a subchannel and dealing with the subchannel-scale whole core T/H analysis.

• Nuclear Engineering and Technology
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• 제41권3호
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• pp.295-306
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• 2009

The subchannel grade void distributions in the NUPEC (Nuclear Power Engineering Corporation) BFBT (BWR Full-Size Fine-Mesh Bundle Tests) facility were evaluated with the subchannel analysis code MATRA and the system code MARS. Fifteen test series from five different test bundles were selected for an analysis of the steady-state subchannel void distributions. Two transient cases, a turbine trip without a bypass as a typical power transient and a re-circulation pump trip as a flow transient, were also chosen for this analysis. It was found that the steady-state void distributions calculated by both the MATRA and MARS codes coincided well with the measured data in the range of thermodynamic qualities from 5% to 25%. The results of the transient calculations were also similar and were highly feasible. However, the computational aspects of the two codes were clearly different.

• Nuclear Engineering and Technology
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• 제31권3호
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• pp.314-327
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• 1999

A subchannel analysis code MATRA applicable to PWRs and ALWRs has been developed to be run on an IBM PC or HP WS based on the existing CDC CYBER mainframe version of COBRA-Rf-1. This MATRA code is a thermal-hydraulic analysis code based on the subchannel approach for calculating the enthalpy and How distribution in fuel assemblies and reactor cores for both steady-state and transient conditions. HATRA has been provided with an improved structure, various functions, and models to give more convenient user environment and to enhance the code accuracy. Among them, the pressure drop model has been improved to be applied to non-square-lattice rod arrays, and the models for the lateral transport between adjacent subchannels have been improved to enhance the accuracy in predicting two-phase flow phenomena. The predictions of MATRA were compared with the experimental data on the flow and enthalpy distribution in some sample rod-bundle cases to evaluate the performance of MATRA. All the results revealed that the predictions of MATRA were better than those of COBRA-IV-I.

• 한국통신학회논문지
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• 제30권6A호
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• pp.455-464
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• 2005

본 논문에서는 OFDMA, TDD 전송방식을 채택한 WiBro 시스템 표준의 resource 할당 알고리듬을 시뮬레이션을 통해 구현하였고, bandwidth efficiency를 보다 더 높이기 위한 알고리듬을 제안하였다. 제안하는 알고리듬은, 변조지수를 최대로 하여 채널 이득이 큰 subchannel부터 작은 subchannel 순서로 비트를 할당한다. 이 때 subchannel의 수를 증가시킴으로 인해 전송되어야 할 power가 증가하는데, 이 power가 available power를 넘는 순간에 할당되는 subchannel의 변조지수를 조정하여 available power를 넘지 않게 하는 방법이다. 기존의 greedy algorithm이나 WiBro 시스템과 비교하였을 때, 제안하는 알고리듬을 적용할 경우 power가 제한되어 있는 것으로 인해 사용되는 subchannel의 수가 현저히 작아져서 bandwidth efficiency 측면에서는 좋은 성능을 나타내게 한다. 하지만 channel attenuation이 큰 환경에서는 최대 throughput이 떨어지는 문제점이 발생하는데, 타 시스템에서 지원하는 만큼의 throughput을 보장하는 문제 역시 중요하기 때문에, 이를 극복하기 위해 이 경우에는 available power를 추가 할당하여 사용되는 subchannel수를 증가시켰다. 이 경우, 기존 시스템과 비슷한 throughput을 보장하면서 bandwidth 이득은 더 높게 얻을 수 있음을 확인하였다.

• 한국통신학회논문지
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• 제21권11호
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• pp.2923-2931
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• 1996

In this paper, the symbol error probability for orthogonal frequency division multiplexing (OFDM) in the multipath fading environment is obtained analytically. In the analysis, OFDM signals with and without the guard interval are considered, and the two-ray fading model is used for the multi-path fading channel. From the analysis results, it is found that the adjacent subchannel interfernce increases the symbol error rate when the guard interval is not employed or shorter than the length of the delay. It is also shown that the adjacent subchannel interference is a Gaussian random variable and its variance depends on the subchannel location and the number of subchannels. Finally, it is found that the variance of the subchannel interference also increases as the power of the signal increases for the OFDM with insufficient guard interval, yieldin an irreducible error at high signal to noise ratio.

• Nuclear Engineering and Technology
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• 제48권2호
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• pp.376-385
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• 2016

A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20-115% of the nominal flow rate and $60^{\circ}C$ (equivalent to Re ~ 37,100) at the inlet of the test rig. The pressure loss data in three measured subchannels were almost identical regardless of the subchannel locations. The flow rate at each type of subchannel was identified and the flow split factors were evaluated from the measured data. The predicted correlations and the computational fluid dynamics results agreed reasonably with the experimental data.

• Nuclear Engineering and Technology
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• 제27권4호
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• pp.491-502
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• 1995

핵연료집합체 부수로 유동장에 대한 상세한 정보에 기초해 교차류를 정확히 예측하는 것은 핵연료의 성능을 해석하는데 중요한 요소이다. 본 연구에서는 저-Reynolds 수 k-$\varepsilon$ 난류모형을 채택하여 인접한 두 부수로 사이에 발생하는 교차류를 해석하였다 또한, 2차유동을 정확히 모사하기 위해서 비등방성 대수응력모형을 사용하였다. 이상의 난류 모형은 유한요소법을 통해 해석되었으며 가용한 실험자료와 비교하여 검증하였다. 그리고, 부수로 유동장에 대한 수치해석 결과를 이용하여 횡방향 합력손실계수의 상관식을 구성하였다. 상관식은 교차류를 제공하는 부수로의 축방향 속도에 대한 교차류의 속도비, 제공받는 부수로의 Reynolds 수 그리고 Pitch-to-diameter의 함수로 구성되었다.

• Advances in Energy Research
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• 제5권2호
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• pp.91-105
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• 2017

Thermal hydraulic (TH) analysis of nuclear power reactors is utmost important. In this way, the numerical codes that preparing TH data in reactor core are essential. In this paper, a subchannel analysis of a Russian pressurized water reactor (WWER1000) core with enhanced numerical code is carried out. For this, in fluid domain, the mass, axial and lateral momentum and energy conservation equations for desired control volume are solved, numerically. In the solid domain, the cylindrical heat transfer equation for calculation of radial temperature profile in fuel, gap and clad with finite difference and finite element solvers are considered. The dependence of material properties to fuel burnup with Calza-Bini fuel-gap model is implemented. This model is coupled with Isotope Generation and Depletion Code (ORIGEN2.1). The possibility of central hole consideration in fuel pellet is another advantage of this work. In addition, subchannel to subchannel and subchannel to rod connection data in hexagonal fuel assembly geometry could be prepared, automatically. For a demonstration of code capability, the steady state TH analysis of a the WWER1000 core is compromised with Thermal-hydraulic analysis code (COBRA-EN). By thermal hydraulic parameters averaging Fuel Assembly-to-Fuel Assembly method, the one sixth (symmetry) of the Boushehr Nuclear Power Plant (BNPP) core with regular subchannels are modeled. Comparison between the results of the work and COBRA-EN demonstrates some advantages of the presented code. Using the code the thermal modeling of the fuel rods with considering the fission gas generation would be possible. In addition, this code is compatible with neutronic codes for coupling. This method is faster and more accurate for symmetrical simulation of the core with acceptable results.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3775-3786
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• 2023

Mixing Vane Grid (MVG) is one of the most important structures in fuel assembly due to its high performance in mixing the coolant and ultimately increasing Critical Heat Flux (CHF), which avoids the temperature rising suddenly of fuel rods. To evaluate the mixing performance of the MVG, a Total Diffusion Coefficient (TDC) mixing coefficient is defined in the subchannel analysis code. Conventionally, the TDC of the spacer grid is obtained from the combination of experiments and subchannel analysis. However, the processing of obtaining and determine a reasonable TDC is much challenging, it is affected by boundary conditions and MVG geometries. In is difficult to perform all the large and costing rod bundle tests. In this paper, the CFD method was applied in TDC analysis. A typical 5 × 5 MVG was simulated and validated to estimate the mixing performance of the MVG. The subchannel code was used to calculate the TDC. Firstly, the CFD method was validated from the aspect of pressure drop and lateral temperature distribution in the subchannels. Then the effect of boundary conditions including the inlet temperature, inlet velocities, heat flux ratio between hot and cold rods and the arrangement of hot and cold rods on MVG mixing and TDC were studied. The geometric effects on mixing are also carried out in this paper. The effect of vane pattern on mixing was investigated to determine which one is the best to represent the grid's mixing performance.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집E
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• pp.482-487
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• 2001

A CFD study was conducted to evaluate the nuclear fuel assembly coolant mixing that is promoted by the flow-mixing vanes on the grid spacer. Four mixing vanes (split vane, swirl vane, twisted vane, hybrid vane) were chosen in this study. A single subchannel of one grid span is modeled using the flow symmetry. The three mixing vanes other than swirl vane generate a large crossflow between the subchannels and a skewed elliptic swirling flow in the subchannel near the grid spacer. The swirl vane induces a circular swirling flow in the subchannel and a negligible crossflow. The split vane and the twisted vane were predicted to result in relatively larger pressure drop across the grid spacer. Since the average turbulent kinetic energy in the subchannel rapidly decreases to a fully developed level downstream of the spacer, turbulent mixing caused by the mixing vanes appears to be not as effective as swirling flow mixing in the subchannel. In summary, the CFD analysis represented the overall characteristics of coolant mixing well in a nuclear fuel assembly with the flow mixing vanes on the grid spacer. The CFD study is therefore quite useful for the development of an advanced flow-mixing vane.