• 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.

• 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.

• Nuclear Engineering and Technology
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• 제28권1호
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• pp.56-71
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• 1996

지지격자를 갖는 5$\times$5 핵연료 봉다발부수로내에서 국부 수력특성인자들을 레이저 유속측정장치인 LDV(Laser Doppler Velocimeter)를 이용하여 측정하였다. 이 연구는 지지격자가 봉다발 난류유동구조에 미치는 영향에 관한 연구에 관점을 두었다. 축방향속도, 난류강도, 편이도, 편평도 등의 측정인자들을 측정하였다. 압력강하를 측정하여 지지격자의 손실계수와 봉다발의 마찰계수를 구하였다 실험결과로부터 활발한 난류혼합거리는 지지격자로부터 x/D$_{h}$=10까지이고, 강제 혼합거리는 지지격자로부터 x/D$_{h}$=20까지임이 관찰되었다. 지지격자하류에서의 축방향 난류강도 감쇄거동은 mesh격자나 screen을 통과하는 난류유동과 같은 경향을 보여주었다. 측정된 자료로부터 부수로해석 code에 입력함수로 적용할 수 있는 국부 난류혼합계수상관식을 구하였다. 국부 혼합계수분포 경향을 관찰하여보면 지지격자 근처에서 최대값을 보이고, 하류방향으로 진행하면서 안정된 값을 갖는다.

• 한국통신학회논문지
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• 제29권6A호
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• pp.623-630
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• 2004

본 논문에서는 하향 링크의 수신기 복잡도를 증가시키지 않으면서도 하향 링크 시스템 용량을 증가시키기 위해서, OFDM 시스템에서 SFBC(Space-frequency Block Code) 전송 다이버시터와 사전등화 기법을 적용한 적응 변조 방식을 제안하고 분석하였다. OFDM 시스템에서 SFBC 전송 다이버시터와 결합해서 적응 변조를 사용하면 시스템 용량이 크게 증가하게 된다. 하지만, 심각한 주파수 선택적 페이딩을 겪는 경우 심볼 간 간섭의 발생으로 인해 시스템 성능이 저하되게 된다. 이 경우, 본 논문에서는 부채널 간의 그룹핑을 통하여 인접 부채널 간 그룹별사전 등화기를 적용함으로써, 적응형 변조방식의 효율을 높일 수 있는 알고리듬을 제안하였다. 컴퓨터 시뮬레이션을 통하여 제안된 시스템이 시변 다중 경로 환경에서 기존의 기술보다 더 좋은 시스템 성능과 용량을 가짐을 확인하고, 심각한 주파수 선택적 페이딩을 겪는 채널 환경에서도 제안한 사전 등화기를 사용해 시스템 성능을 유지할 수 있음을 확인하였다.

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.499-507
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• 2020

Research reactors for radioisotope production, fuel and material testing and research activities are designed, constructed and operated based on the society's needs. In this study, neutronic and thermal hydraulic design of a high neutron flux research reactor core for radioisotope production is presented. Main parameters including core excess reactivity, reactivity variations, power and flux distribution during the cycle, axial and radial power peaking factors (PPF), Pu₂₃₉ production and minimum DNBR are calculated by nuclear deterministic codes. Core calculations performed by deterministic codes are validated with Monte Carlo code. Comparison of the neutronic parameters obtained from deterministic and Monte Carlo codes indicates good agreement. Finally, subchannel analysis performed for the hot channel to evaluate the maximum fuel and clad temperatures. The results show that the average thermal neutron flux at the beginning of cycle (BOC) is 1.0811 × 10¹⁴ n/㎠-s and at the end of cycle (EOC) is 1.229 × 10¹⁴ n/㎠-s. Total Plutonium (Pu₂₃₉) production at the EOC evaluated to be 0.9487 Kg with 83.64% grade when LEU (UO₂ with 3.7% enrichment) used as fuel. This designed reactor which uses LEU fuel and has high neutron flux and low plutonium production could be used for peaceful nuclear activities based on nuclear non-proliferation treaty concepts.

• 한국통신학회논문지
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• 제23권7호
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• pp.1707-1717
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• 1998

본 논문에서는 협대역 신호가 CDMA(Code Division Multiple Access)시스템과 동일한 주파수대역을 공유하는 할 때, 협대역 신호를 억제하여 CDMA시스템 성능을 향상시키는 방식에 대해서 연구하였다. 협대역 재머 신호를 억제하기 위해서 Multirate 기술을 이용한 polyphase 여파기를 제안하였으며, 현 IS-95의 47텝의 FIR 여파기와 함께 ASIC으로 구현하기 용이하여 순방향 채널에서 재머를 억제하는 방법으로 적합하였다. 재머 억제는 polyphase 여파기를 사용하여 2단계로 총 64개의 채널로 수신신호를 분리하여 각 부채널의 수신전력을 측정한 후 재머가 존재 하는 대역을 결정하고 결정된 부채널을 억제하여 BER율 향상시켰다. 선행측정 여파기를 사용한 결과[1]보다는 최대 O.8dB 정도가 성능저하가 있었으나, 부채널 수를 증가시켜 보다 섬세한 부채널 억제방식을 쓰면 보다 성능이 향상될 수 있고, 순방향 채널에서 재머 억제 방식으로 채택하면, 역방향에 대해서, 재머 억제 후 저하된 성능을 보상하기 위한 기술이 수월하다.

• Nuclear Engineering and Technology
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• 제47권5호
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• pp.523-533
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• 2015

Three-dimensional flow phenomena in a wire-wrapped 37-pin fuel assembly mock-up of a Japanese loop-type sodium-cooled fast reactor, Monju, were investigated with a numerical analysis using a general-purpose commercial computational fluid dynamics code, CFX. Complicated and vortical flow phenomena in the wire-wrapped 37-pin fuel assembly were captured by a Reynolds-averaged Navier-Stokes flow simulation using a shear stress transport turbulence model. The main purpose of the current study is to understand the three-dimensional complex flow phenomena in a wire-wrapped fuel assembly to support the license issue for the core design. Computational fluid dynamics results show good agreement with friction factor correlation models. The secondary flow in the corner and edge subchannels is much stronger than that in an interior subchannel. The axial velocity averaged in the corner and edge subchannels is higher than that averaged in the interior subchannels. Three-dimensional multiscale vortex structures start to be formed by an interaction between secondary flows around each wire-wrapped pin. Behavior of the large-scale vortex structures in the corner and edge subchannels is closely related to the relative position between the hexagonal duct wall and the helically wrapped wire spacer. The small-scale vortex is axially developed in the interior subchannels. Furthermore, a driving force on each wire spacer surface is closely related to the relative position between the hexagonal duct wall and the wire spacer.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.439-445
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• 2021

In view of the complex geometric structure of the rod bundle channel and the limitation of the current CHF visualization experiment technology, it is very difficult to obtain the rod bundle CHF mechanism directly through the phenomenon of the rod bundle CHF visualization experiment. In order to obtain the applicable CHF mechanism assumption for rod bundle channel, firstly, five most representative DNB type round tube CHF mechanistic models are obtained with evaluation and screening. Then these original round tube CHF mechanistic models based on inlet conditions are converted to local conditions and coupled with subchannel analysis code ATHAS. Based on 5 × 5 full-length rod bundle CHF experimental data independently developed by Nuclear Power Institute of China (NPIC), the applicability research of each model for CHF prediction performance in rod bundle channel is carried out, and the commonness and difference of each model are comparatively studied. The CHF mechanism assumption of superheated liquid layer depletion that is most likely to be applicable for the rod bundle channel is selected and two directions that need to be improved are given. This study provides a reference for the development of CHF mechanistic model in rod bundle channel.

• 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.

• 한국전자파학회논문지
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• 제12권2호
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• pp.268-277
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• 2001

본 논문에서는 IEEE 802.11a 무선 LAN에 적합한 OFDM 시스템에 비터비 복호를 사용하는 길쌈 부호와 단일 탭의 LMS 등화기를 적용하여 실내 무선 채널 환경에서 시스템 성능을 시뮬레이션을 통해 분석한다. 실내 무선 채널은 라이시안 페이딩 모델링하고, 부채널변조 방식으로는 QPSK와 16QAM을 사용한다. 직접 파 대 간섭파 전력비 K=5 dB 인 라이시안 페이딩 채널에서 길쌈 부호 및 비터비 복호를 사용하는 경우, 경판정에서 QPSK는 8.6 dB, 16QAM 은 19.2dB, 연판정에서 QPSK는 5.3dB, 16QPSK는 5.3dB, 16QAM은 9.8dB에서 $10^{-4}$의 BER을 만족하였다. 또한 16QAM/OEFM 방식에 단일 탭의 LMS 등화기를 사용하면 길쌈 부호만을 사용한 경우보다 경판정 비터비 복호의 경우 8.6dB,연판정의 경우에는 2dB의 성능이 향상됨을 알 수 있었다.