• Nuclear Engineering and Technology
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• v.53 no.12
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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.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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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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• v.28 no.1
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• pp.56-71
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• 1996

The local hydrulic characteristics in subchannels of 5$\times$5 nuclear fuel bundles with spacer grids were measured at upstream and downstream of the spacer grid for the investigation of the spacer grid effects on turbulent flow structure by using an LDV(Laser Doppler Velocimeter). The measured parameters are axial velocity and turbulent intensity, skewness factor, and flatness factor. Pressure drops were also measured to evaluate the loss coefficient for the spacer grid and the friction factor for rod bundles. From these data, it was found that the turbulent mixing and forced mixing occur up to $x/D^h=10$ and 20 from the spacer grid, respectively. The turbulence decay behind spacer grid behaves in the similar decay rate as turbulent flow through mesh grids or screens. Mixing factors useful in subchannel analysis code were correlated from the data and show the highest value near spacer grid and then have a stable values.

• The Journal of Korean Institute of Communications and Information Sciences
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• v.29 no.6A
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• pp.623-630
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• 2004

The adaptive modulation along with SFBC transmit diversity is a very effective method to increase the capacity of an OFDM system. However, severe performance degradation is resulted when inter-symbol interference is applied due to frequency-selective fading in mobile communications. In this paper, we have proposed and analyzed an OFDM system with SFBC transmit diversity and adaptive modulation scheme based on pre-equalization methods, in order to increase the data transmission rate in the downlink without much increase in system complexity. By introducing subchannel grouping and the pre-equalization method among adjacent subchannels, we could enhance the efficiency of the adaptive modulation a lot. By computer simulation, it has been proven that the proposed schemes show a better BER and throughput performance than the conventional schemes under severely time-varying multipath fading channel.

• Nuclear Engineering and Technology
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• v.52 no.3
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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.

• The Journal of Korean Institute of Communications and Information Sciences
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• v.23 no.7
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• pp.1707-1717
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• 1998

Polyphase filtering techniques is used to suppress the narrowband jammer signal such as USDC TDMA overlaying the band occupied by DS-CDMA system. In the proposed jammer suppression, the received signal is separated into 64 subchannels in two stages by polyphase filtering and the location of the narrowband jammer signal is determined by measuring each subchannel power and the contaminated subchannels are simply blocked. The $E_{b}/N_{0}$ 0/ improvement of the CDMA system from jammer suppession was outstanding. The $E_{b}/N_{0}$ degradation in comparison with a performance of no jammer is around 0.8dB in the worst case. The results are also compared with those of linear prediction jammer suppression. The implementation of the ployphase jammer suppression requires great amount of data processing and computation compared to linear predication filter. Thus it is more appropriate to implement with a ASIC rather than WITH several DSPs for user terminals of forward link.

• Nuclear Engineering and Technology
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• v.47 no.5
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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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• v.53 no.2
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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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• v.55 no.3
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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.

• The Journal of Korean Institute of Electromagnetic Engineering and Science
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• v.12 no.2
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• pp.268-277
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• 2001

In this paper, the system performance with the convolution code using a Viterbi decoding and the one tap LMS equalizer applied to the OFDM system, which is suitable for IEEE 802.1la wireless LAN in indoor wireless channel, is analyzed through computer simulation. Indoor wireless channel is modeled as Rician fading channel, and QPSK and 16QAM scheme are used for subchannel modulation. In Rician fading channel with the power ratio of the direct path signal to the scattered signals, K=5 dB, BER of $10^{-4}$ is satisfied if the SNRs of the QPSK/OFDM and the 16QAM/OFDM are 8.6 dB and 19.2 dB in hard decision and 5.3 dB and 9.8 dB in soft decision, respectively. Compared with convolution code scheme, it is observed that 16QAM/OFDM system with the one tap LMS equalizer has the performance improvement of 8.6 dB and 2 dB in hard decision and soft decision, respectively.