• Journal of electromagnetic engineering and science
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• v.11 no.2
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• pp.133-142
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• 2011

Coded cooperation that integrates channel coding in cooperative transmission has gained a great deal of interest in wireless relay networks. The performance analysis of coded cooperation protocol with multiple relays is investigated in this paper. We show that the diversity order achieved by the coded cooperation in a multi-relay wireless network is not only dependent on the number of cooperating relays but is also dependent on the code-rate of the system. We derive the code-rate bound, which is required to achieve the full diversity gain of the order of cooperating nodes. The code-rate required to achieve full diversity is a linearly decreasing function of the number of available relays in the network. We show that the instantaneous channel state information (CSI)-based relay selection can effectively alleviate this code-rate bound. Analysis shows that the coded cooperation with instantaneous CSI-based relay selection can achieve the full diversity, for an arbitrary number of relays, with a fixed code-rate. Finally, we develop tight upper bounds for the bit error rate (BER) and frame error rate (FER) of the relay selection based on coded cooperation under a Rayleigh fading environment. The analytical upper bounds are verified with simulation results.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.413-418
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• 1996

The main purpose of this paper is to develop the modified LTC code for accurate analysis of the boron concentration behavior of all components in the Nuclear Steam Supply System (NSSS). This is achieved by adapting a multi-cell mad to the existing Long Term Cooling (LTC) code. To verify the modified LTC, the simulated results were compared with the actual test results measured during YGN 4 initial criticality test. It was shown that the simulated results of this modified LTC were in good agreement with the actual test results. Also, the boron concentration behavior analysis were performed using the modified LTC code for both direct and indirect dilution/boration nude using YGN 3,4 design data. This modified LTC code can provide a valuable information in predicting boron concentration behavior during power maneuvering such as startup operation, shutdown operation and load follow operation. It is expected that the modified LTC can be applied to both on-line and off-line mode using Plant Computer System(PCS).

• Nuclear Engineering and Technology
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• v.43 no.2
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• pp.187-194
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• 2011

Korea Nuclear Fuel is developing the X-GEN fuel which shows high performance and robust reliability for the worldwide supply. However, the simplified code systems such as CESEC-III which were developed in 1970s are still used in the current Non-LOCA safety analysis of OPR1000 and APR1400 plants. Therefore, it is essential to secure an advanced safety analysis methodology to make the best use of the merits of X-GEN fuel. To accomplish this purpose, the $\b{i}$ntegrated $\b{s}$afety $\b{a}$nalysis $\b{m}$ethodology (iSAM), is developed by selecting the best-estimate thermal-hydraulic code RETRAN. iSAM possesses remarkable advantages, such as generality, integrity, and designer-friendly features. That is, iSAM can be applied to both OPR1000 and APR1400 plants and uses only one computer code, RETRAN, in the whole scope of the non-LOCA safety analyses. Also the iSAM adopts the unique and automatic initialization and run tool, $\b{a}$utomatic $\b{s}$teady-$\b{s}$tate $\b{i}$nitialization and $\b{s}$afety analysis too l (ASSIST), to enable unhandy designers to use the new design code RETRAN without difficulty. In this paper, a brief overview of the iSAM is given, and the results of applying the iSAM to typical non-LOCA transients being checked during the reload design are reported. The typical non-LOCA transients selected are the single control element assembly withdrawal (SCEAW) accident, the asymmetric steam generator transients (ASGT), the locked rotor (LR) accident, and bank CEA withdrawal (BCEAW) event. Comparison to current licensing results shows a close resemblance; thus, it reveals that the iSAM can be applied to the non-LOCA safety analysis of OPR1000 and APR1400 plants.

• Journal of the Korea Institute of Information and Communication Engineering
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• v.19 no.3
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• pp.683-688
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• 2015

Recently, various static analysis tools for software defect detection are becoming widely used in practice. However, there is little public information of the most frequent defects in commercial areas until now. In this paper, we analyze the defects found by CodeSonar, a static analysis tool that finds defects in C/C++, Java programs. So we report the most frequent defects by various aspects in Dongnam area, Korea.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.767-790
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• 2013

MAAP4 is a computer code that can simulate the response of a light water reactor power plant during severe accident sequences, including actions taken as part of accident management. The code quantitatively predicts the evolution of a severe accident starting from full power conditions given a set of system faults and initiating events through events such as core melt, reactor vessel failure, and containment failure. Furthermore, models are included in the code to represent the actions that could mitigate the accident by in-vessel cooling, external cooling of the reactor pressure vessel, or cooling the debris in containment. A key element tied to using a code like MAAP4 is an uncertainty analysis. The purpose of this paper is to present a MAAP4 based analysis to examine the sensitivity of a key parameter, in this case hydrogen production, to a set of model parameters that are related to a Level 2 PRA analysis. The Level 2 analysis examines those sequences that result in core melting and subsequent reactor pressure vessel failure and its impact on the containment. This paper identifies individual contributors and MAAP4 model parameters that statistically influence hydrogen production. Hydrogen generation was chosen because of its direct relationship to oxidation. With greater oxidation, more heat is added to the core region and relocation (core slump) should occur faster. This, in theory, would lead to shorter failure times and subsequent "hotter" debris pool on the containment floor.

• Nuclear Engineering and Technology
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• v.46 no.5
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• pp.641-654
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• 2014

A new computer code, named CORONA (Core Reliable Optimization and thermo-fluid Network Analysis), was developed for the core thermo-fluid analysis of a prismatic gas cooled reactor. The CORONA code is targeted for whole-core thermo-fluid analysis of a prismatic gas cooled reactor, with fast computation and reasonable accuracy. In order to achieve this target, the development of CORONA focused on (1) an efficient numerical method, (2) efficient grid generation, and (3) parallel computation. The key idea for the efficient numerical method of CORONA is to solve a three-dimensional solid heat conduction equation combined with one-dimensional fluid flow network equations. The typical difficulties in generating computational grids for a whole core analysis were overcome by using a basic unit cell concept. A fast calculation was finally achieved by a block-wise parallel computation method. The objective of the present paper is to summarize the motivation and strategy, numerical approaches, verification and validation, parallel computation, and perspective of the CORONA code.

• Nuclear Engineering and Technology
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• v.41 no.7
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• pp.953-968
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• 2009

It is important to accurately predict the temperature and density distributions in large stratified enclosures both for design optimization and accident analysis. Current reactor system analysis codes only provide lumped-volume based models that can give very approximate results. Previous scaling analysis has shown that stratified mixing processes in large stably stratified enclosures can be described using one-dimensional differential equations, with the vertical transport by jets modeled using integral techniques. This allows very large reductions in computational effort compared to three-dimensional CFD simulation. The BMIX++ (Berkeley mechanistic MIXing code in C++) code was developed to implement such ideas. This paper summarizes major models for the BMIX++ code, presents the two-plume mixing experiment simulation as one validation example, and describes the codes' application to the liquid salt buffer pool system in the AHTR (Advanced High Temperature Reactor) design. Three design options have been simulated and they exhibit significantly different stratification patterns. One of design options shows the mildest thermal stratification and is identified as the best design option. This application shows that the BMIX++ code has capability to provide the reactor designers with insights to understand complex mixing behavior with mechanistic methods. Similar analysis is possible for liquid-metal cooled reactors.

• Journal of the Korean Society for Aeronautical & Space Sciences
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• v.32 no.5
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• pp.1-9
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• 2004

The performance analysis code has been developed for vertical take-off and landing(VTOL) UAV which can be utilized as a trade analysis tool in the pre-conceptual design phase. The UAV requires VTOL capability and high speed cruise performance. The main logic of this performance analysis code is to estimate performance parameters of each mission segment by mission fuel weight iteration. The reliability of this performance analysis code is discussed by comparing the data of existing dual flight mode VTOL UAVs such as Boeing CRW and Bell Tilt Rotor.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.1-18
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• 2021

The criticality analysis of VVER-1000 mock-up benchmark experiments from the LR-0 research reactor operated by the Research Center Rez in the Czech Republic has been conducted with the MCS Monte Carlo code developed at the Computational Reactor Physics and Experiment laboratory of the Ulsan National Institute of Science and Technology. The main purpose of this work is to evaluate the newest ENDF/B-VIII.0 nuclear data library against the VVER-1000 mock-up integral experiments and to validate the criticality analysis capability of MCS for light water reactors with hexagonal fuel lattices. A preliminary code/code comparison between MCS and MCNP6 is first conducted to verify the suitability of MCS for the benchmark interpretation, then the validation against experimental data is performed with both ENDF/B-VII.1 and ENDF/B-VIII.0 libraries. The investigated experimental data comprises six experimental critical configurations and four experimental pin-by-pin power maps. The MCS and MCNP6 inputs used for the criticality analysis of the VVER-1000 mock-up are available as supplementary material of this article.

• Proceedings of the IEEK Conference
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• 2006.06a
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• pp.999-1000
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• 2006

The LDPC Code is focusing a powerful FEC(Forward Error Correction) codes for 4G Mobile Communication system. LDPC codes are used minimizing channel errors by modeling AWGN Channel as VDSL system. The performance of LDPC code is better than that of turbo code in long code word on iterative decoding algorithm. LDPC code are encoded by sparse parity check matrix. there are decoding algorithms for a LDPC code, Bit Flipping, Message passing, Sum-Product. Because LDPC Codes use low density parity bit, mathematical complexity is low and relating processing time becomes shorten.