• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2847-2858
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• 2021

When a nuclear reactor with rectangular fuel assemblies runs for a long time, impurities and debris may be taken into coolant channels, which may cause flow blockage, and the blocked fuel assemblies might be destroyed. Therefore, the purpose of this study is to perform a thermal-hydraulic analysis of a rectangular fuel assembly by STAR-CCM+, under the condition of one subchannel with 80% blockage ratio. A rectangular fuel assembly of the International Atomic Energy Agency (IAEA) 10 MW material test reactor (MTR) is chosen. In view of the gasket material taken into the coolant channel is close to the single side of the coolant channel, in the flow blockage accident of the Oak Ridge Research Reactor (ORRR), a new blockage category called single side blockage is attempted. The blockage positions include inlet, middle and outlet, and the blockage is set as a cuboid. It is found by simulations that the blockage redistributes the mass flow rate, and large vortices appear locally. The peak temperature of the cladding is maximum, when the blockage is located at the single side of the coolant channel inlet, and no boiling occurs in all blockage cases. Moreover, as the height of the blockage increases, the damage caused by the blockage increases slightly.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4296-4309
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• 2022

Mesh and turbulence model sensitivity analyses have been performed on computational fluid dynamics simulations executed with Hydra and ANSYS Fluent for a single CANadian Deuterium Uranium (CANDU) 37M nuclear fuel bundle placed within a standard pressure tube. The goal of this work was to perform a methodical analysis to objectively determine an appropriate mesh and to gauge the sensitivity of different turbulence models for CANDU subchannel flow under isothermal conditions. The boundary conditions and material properties are representative of normal operating conditions in a high-powered channel of the Darlington Nuclear Generating Station. Four meshes were generated with ANSYS Workbench Meshing, ranging from 22 to 84 million cells, and analyzed here to determine an appropriate level of mesh resolution and quality. Five turbulence models were compared in the turbulence model sensitivity analysis: standard k - ε, RNG k - ε, realizable k - ε, SST k - ω, and the Reynolds Stress Model. The intent of this work was to gain confidence in mesh generation and turbulence model selection of a single bundle to inform the decision making of subsequent investigations of an entire fuel channel containing a string of twelve bundles.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.3902-3909
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• 2021

SACOS series of subchannel analysis codes have been developed by XJTU-NuTheL for many years and are being used for the thermal-hydraulic safety analysis of various reactor cores. To achieve fine whole core pin-level analysis, the input preprocessing and parallel capabilities of the code have been developed in this study. Preprocessing is suitable for modeling rectangular and hexagonal assemblies with less error-prone input; parallelization is established based on the domain decomposition method with the hybrid of MPI and OpenMP. For domain decomposition, a more flexible method has been proposed which can determine the appropriate task division of the core domain according to the number of processors of the server. By performing the calculation time evaluation for the several PWR assembly problems, the code parallelization has been successfully verified with different number of processors. Subsequent analysis results for rectangular- and hexagonal-assembly core imply that the code can be used to model and perform pin-level core safety analysis with acceptable computational efficiency.

• 한국통신학회논문지
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• 제30권4C호
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• pp.200-206
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• 2005

적응 전송 기법은 변조방식, 부호율, 전력 등의 전송 매개변수를 채널의 상태에 적응시켜 무선 통신시스템의 성능을 향상시키는 기법이다. OFDM (Orthogonal frequency division multiplexing) 시스템에서는 이러한 적응기법을 부반송파별로 적용시킬 수 있다. 이 논문에서는 각 부채널의 상태에 따라 부반송파에 최적의 부호율을 적응시키는 적응부호율 OFDM 시스템을 고려한다. 성능 분석을 통해 적응부호율 OFDM 시스템이 비트오류율 $10^{-6}$에서 고정부호율 OFDM 시스템에 비해 $3\sim6$ dB의 신호 대 잡음비 이득 또는 $30\sim50\%$의 데이터 전송률 증가를 얻을 수 있음을 보인다.

• 대한전자공학회논문지TC
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• 제45권11호
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• pp.9-14
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• 2008

본 논문에서는 직교 주파수 분할 다중 접속 방식 (OFDMA, Orthogonal Frequency Division Multiple Access)에서 다중 사용자 다이버시티 이득을 증가시키기 위하여 사용자간 충돌 (collisions)을 피하도록 하는 스케줄링 기법을 제안한다. 제안하는 스케줄링 기법은 각 사용자간 채널 상태의 차이를 조사하여 최소 충돌 조건 (minimum collision criterion)을 만족시키도록 동작한다. 또한 제안된 스케줄링을 사용했을 때 OFDMA 시스템의 전송률을 분석하였는데, 이 분석을 통해 OFDMA 시스템의 전송률이 선택된 사용자들 간의 충돌 횟수에 따라 달라진다는 것을 확인할 수 있다. 모의실험을 통해 제안하는 스케줄링 기법이 사용자 간 충돌의 감소를 통하여 시스템의 전송률을 증대시킬 수 있음을 보인다. 또한 기존의 스케줄링 기법과의 비교를 통해, 제안된 스케를링 기법이 전송률 측면에서 더 우월하다는 것을 확인한다.

• 한국위성정보통신학회논문지
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• 제10권2호
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• pp.14-18
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• 2015

최근 전자전은 현대전의 핵심으로 자리매김하고 있으며 미래 전투체계인 네트워크 중심전에 따른 통신 생존의 중요성은 나날이 부각되고 있다. 본 논문에서는 GPS 재밍 등 군통신에서 사용되는 전자 방해책인 재밍 기술을 효과적으로 제거할 수 있는 항재밍 방안에 대해 제안하기 위해 부분대역 재밍 환경과 군통신에 널리 사용되는 IEEE 802.16 WiMAX 프로토콜을 고려하였다. 기존의 주파수 도약 방법과는 다른 부채널화를 통한 알고리즘을 제안하였으며, 제안한 알고리즘의 성능 확인을 위해 부분대역 재밍 파라미터에 따른 최대 채널용량에 해당하는 최대 부채널 개수를 확인할 수 있었다.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1071-1082
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• 2016

Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the invessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

• Nuclear Engineering and Technology
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• 제38권8호
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• pp.809-818
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• 2006

The existing experimental data related to the turbulent mixing factor in rod arrays is examined and a new definition of the turbulent mixing factor is introduced to take into account the turbulent mixing of fluids with various Prandtl numbers. The new definition of the mixing factor is based on the eddy diffusivity of energy. With this definition of the mixing factor, it was found that the geometrical parameter, ${\delta}_{ij}/D_h$ correlates the turbulent mixing data better than Sid, which has been used frequently in existing correlations. Based on the experimental data for a highly turbulent condition in square rod arrays, a correlation describing turbulent mixing dependent on the parameter ${\delta}_{ij}/D_h$ has been developed. The correlation is insensitive to the Re number and it takes into account the effect of the turbulent Prandtl number. The proposed correlation predicts a reasonable mixing even at a lower S/d ratio.

• 대한기계학회논문집B
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• 제40권5호
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• pp.329-337
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• 2016

연료집합체의 지지격자에 설치된 혼합날개는 난류 강화 기구로서 부수로 내부에서 선회류 또는 연료봉 간극사이에서 횡류를 발생시켜 대류열전달을 증진시키는 역할을 한다. 따라서 혼합날개의 기하학적인 형상 및 배열 형태는 혼합날개의 성능을 결정하는 중요한 인자이다. 본 연구에서는 OECD/NEA의 벤치마크 계산에서 활용된 분할 형태의 혼합날개가 장착된 $5{\times}5$ 연료집합체 내부에서의 유동분포 특성을 파악하기 위해 상용 전산유체역학 소프트웨어인 ANSYS CFX R.14를 사용하여 계산을 수행하였고, 계산결과를 MATiS-H 시험장치의 측정값과 비교하였다. 또한 분할 형태의 혼합날개 형상이 연료집합체 내부유동 형태에 미치는 영향에 대해 설명하였다.

• Nuclear Engineering and Technology
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• 제55권3호
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• pp.999-1008
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• 2023

Deformed fuel rods can cause a partial blockage of the flow area in a subchannel. Such flow blockage will influence the core coolant flow and further the core heat transfer during the reflooding phase and subsequent severe accidents. Nevertheless, most of the system analysis codes simulate the accident process based on the assumed flow blockage ratio, resulting in inconsistencies between simulated results and actual conditions. This paper aims to study the influence of flow blockage in severe accident scenario of the CAP1400 reactor. First, the flow blockage model of ISAA code is improved based on the FRTMB module. Then, the ISAA-FRTMB coupling system is adopted to model and calculate the QUENCH-LOCA-0 experiment. The correctness and validity of the flow blockage model are verified by comparing the peak cladding temperature. Finally, the DVI Line-SBLOCA accident is induced to analyze the influence of flow blockage on subsequent CAP1400 reactor core heat transfer and core degradation. From the results of the DVI Line-SBLOCA accident analysis, it can be concluded that the blockage ratio is in the range of 40%-60%, and the position of severe blockage is the same as that of cladding rupture. The blockage reduces the circulation area of the core coolant, which in turn impacts the heat exchange between the core and the coolant, leading to the early failure and collapse of some core assemblies and accelerating the core degradation process.