• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.18 no.3
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• pp.202-210
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• 2006

In the nuclear power plant, emergency core coolant system (ECCS) is furnished at reactor coolant system (RCS) in order to cool down high temperature water in case of emergency. However, in this coolant system, thermal stratification phenomenon can be occurred due to coolant leaking in the check valve. The thermal stratification produces excessive thormal stresses at the pipe wall so as to yield thermal fatigue crack (TFC) accident. In the present study, when the turbulence penetration occurs in the branch pipe, the maximum temperature differences of fluid at the pipe cross-sections of the T-branch with thermal stratification are examine.

• Journal of Semiconductor Engineering
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• v.1 no.2
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• pp.74-80
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• 2020

Due to its superior error correcting performance with affordable hardware complexity, the Polar code becomes one of the most important error correction codes (ECCs) and now intensively examined to check its applicability in various fields. However, Successive Cancellation (SC) decoding that brings the advanced Successive Cancellation List (SCL) decoding suffers from the long latency due to the nature of serial processing limiting the practical implementation. To mitigate this problem, many decoding architectures, mainly divided into pruning and parallel decoding, are presented in previous manuscripts. In this paper, we compare the recent SC decoding architectures and analyze them using a tree structure.

• Journal of the Institute of Electronics Engineers of Korea CI
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• v.46 no.3
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• pp.123-129
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• 2009

It is important to locate nodes in the research of wireless sensor network. In this paper, we propose a method that estimates the positions of nodes by using adjacent node information and signal strength in wireless sensor network. With this method, we can find positions of nodes easily because we use Information that nodes have. And we can make a map for all the nodes because we can measure a relative position for an node whose position is not known based on anchor nodes whose positions are already known. In addition, we can confirm whether nodes are placed appropriately. We confirmed that we can locate positions of unknown nodes with small error through verifying the proposed method.

• Journal of the Korean Institute of Intelligent Systems
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• v.18 no.5
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• pp.611-616
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• 2008

Error correcting codes (ECCs) are commonly used to protect against the soft errors. Single error correcting and double error detecting (SEC-DED) codes are generally used for this purpose. The proposed approach in this paper selectively reduced power consumption, delay, and area in single-error correcting, double error-detecting checker circuits that perform memory error correction. The multi-objective genetic algorithm is employed to solve the non -linear optimization problem. The proposed method allows that user can choose one of different non-dominated solutions depending on which consideration is important among them. Because we use multi-objective genetic algorithm, we can find various dominated solutions. Therefore, we can choose the ECC according to the important factor of the power, delay and area. The method is applied to odd-column weight Hsiao code which is well- known ECC code and experiments were performed to show the performance of the proposed method.

• Journal of The Korean Society of Agricultural Engineers
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• v.47 no.5
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• pp.27-39
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• 2005

In the recent design of high ductile fiber-reinforced ECC (engineered cementitious composite), optimizing both processing and mechanical properties for specific applications is critical. This study presents an innovative method to develop new class ECCs, which possess the different fluid properties to facilitate diverse types of processing (i.e., self-consolidating or shotcrete processing) while maintaining ductile hardened properties. In the material design concept, we employ a parallel control of fresh and hardened properties by using micromechanics and cement rheology. Control of colloidal interaction between the particles is regarded as a key factor to allow the performance of the specific processing. To determine how to control the particle interactions and the viscosity of cement suspension, we first introduce two chemical admixtures including a highly charged polyelectrolyte and a non-ionic polymer. Optimized mixing steps and dosages we, then, obtained within the solid concentration predetermined based on micromechanical principle. Test results indicate that the rheological properties altered by this approach were revealed to be highly effective in obtaining the desired function of the fresh ECC, allowing us to readily achieve hardened properties, represented by pseudo strain-hardening behavior in uniaxial tension.

• Nuclear Engineering and Technology
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• v.32 no.4
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• pp.316-327
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• 2000

The Emergency Core Cooling System in current nuclear power plants typically has a considerable number of complex functions and largely cumbersome operator interfaces. Functions for initiation, switch-over between various phases of operation, interlocks, monitoring, and alarming are usually performed by relays and analog comparator logic which are difficult to maintain and test. To improve problems of an analog based ECC (Emergency Core Cooling) System, the trip computer for ECCS based on Dynamic Safety System (DSS) is implemented. The DSS is a computer based reactor protection system that has fail-safe nature and performs a dynamic self-testing. The most important feature of the DSS is the introduction of test signal that send the system into a tripped state. The test signals are interleaved with the plant signals to produce an output which switches between a tripped and health state. The dynamic operation is a key feature of the failsafe design of the system. In this work, a possible implementation of the DSS using PLC is presented for a CANDU Reactor. ECC System of the CANDU Reactor is selected as the reference system.

• Nuclear Engineering and Technology
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• v.12 no.3
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• pp.153-162
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• 1980

The investigation of the fuel cladding temperature behavior and heat transfer mechanism during the reflooding phase of a LOCA plays an important role in performance evaluation of ECCS and safety analysis of water reactors. Reflooding experiments were performed with horizontal and vertical flow channels to investigate the effect of coolant flow channel orientation on rewetting process. Emphasis was mainly placed on the CANDU reactor which has horizontal pressure tubes in core, and the results were compared with those of vertical channel. Also to investigate the rewetting process visually, the experiments by using a rod in annulus and a quartz tube heated outside were performed. It can be concluded that the rewetting velocity in horizontal flow channel is clearly affected by flow stratification, however, the average rewetting velocity is similar to those in vertical flow channel for same conditions.

• Nuclear Engineering and Technology
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• v.41 no.6
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• pp.753-764
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• 2009

JAEA started the LSTF experiments in 1985 for the fourth stage of the ROSA Program (ROSA-IV) for the LWR thermal-hydraulic safety research to identify and investigate the thermal-hydraulic phenomena and to confirm the effectiveness of ECCS during small-break LOCAs and operational transients. The LSTF experiments are underway for the ROSA-V Program and the OECD/NEA ROSA Project that intends to resolve issues in thermal-hydraulic analyses relevant to LWR safety. Six types of the LSTF experiments have been done for both the system integral and separate-effect experiments among international members from 14 countries. Results of four experiments for the ROSA Project are briefly presented with analysis by a best-estimate (BE) code and a computational fluid dynamics (CFD) code to illustrate the capability of the LSTF and codes to simulate the thermal-hydraulic phenomena that may appear during SBLOCAs and transients. The thermal-hydraulic phenomena dealt with are coolant mixing and temperature stratification, water hammer up to high system pressure, natural circulation under high core power condition, and non-condensable gas effect during asymmetric SG depressurization as an AM action.

• Nuclear Engineering and Technology
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• v.13 no.1
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• pp.12-21
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• 1981

An extension work of the previous paper on the rewetting velocity model is presented. Application of the rewetting velocity model presented elsewhere requires a priori values of ø. In the absence of ø values, film boiling heat transfer coefficient ( $h_{df}$ ) and fog-film length (ι) data are needed. To provide these informations, a modified Bromley's correlation is first derived and used to obtain $h_{df}$ values at higher pressure conditions. In addition, the analysis of the precooling parameters, such as ø and ι is further extended using much more expansive PWR FLECHT data. Thus, the applicable range of the rewetting velocity model is further expanded in this work.

• Nuclear Engineering and Technology
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• v.41 no.9
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• pp.1157-1170
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• 2009

A variety of Generation III/III+ reactor designs featuring enhanced safety and improved economics are being proposed by nuclear power industries around the world to solve the future energy supply shortfall. Nanofluid coolants showing an improved thermal performance are being considered as a new key technology to secure nuclear safety and economics. However, it should be noted that there is a lack of comprehensible design works to apply nanofluids to Generation III+ reactor designs. In this work, the review of accident scenarios that consider expected nanofluid mechanisms is carried out to seek detailed application spots. The Axiomatic Design (AD) theory is then applied to systemize the design of nanofluid-engineered nuclear safety systems such as Emergency Core Cooling System (ECCS) and External Reactor Vessel Cooling System (ERVCS). The various couplings between Gen-III/III+ nuclear safety features and nanofluids are investigated and they try to be reduced from the perspective of the AD in terms of prevention/mitigation of severe accidents. This study contributes to the establishment of a standard communication protocol in the design of nanofluid-engineered nuclear safety systems.