• 한국지능시스템학회논문지
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• 제11권9호
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• pp.844-848
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• 2001

This paper introduces unmaned fully automation systems, which are applied for the CSTR(Continuously Stirred Tank Reactor) and SBR (Sequencing Batch Reactor) wastewater treatment system. The pilot plant is constructed in the country side which is little far from a main city. So networks and wireless modules are employed for the data transmission. The SBR plant has a local control and the remote monitoring system which is contained communication parts which consist of ADSL (Asymmetric Digital Subscriber Line) network and CDMA (Code Division Multiple Access) Wireless module. Remote control and monitoring systems are constructed at laboratory in a metropolis.

• Membrane and Water Treatment
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• 제2권2호
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• pp.105-119
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• 2011

This study presents the tests of an Immersed Membrane BioReactor (IMBR) equipped with a draft tube and focuses on the influence of hydrodynamic conditions on membrane fouling in a pilot-scale using a hollow fiber membrane module of ZW-10 under ambient conditions. In this system, the cross-flow velocities across the membrane surface were induced by a cylindrical draft-tube. The relationship between cross-flow velocity and aeration strength and the influence of the cross-flow on fouling rate (under various hydrodynamic conditions) were investigated using Multi-Dimension Scaling (MDS) analysis. MDS technique is especially suitable for samples with many variables and has relatively few observations, as the data about Membrane Bio-Reactor (MBR) often is. Observations and variables are analyzed simultaneously. According to the results, a specialized form of MDS, CoPlot enables presentation of the results in a two dimensional space and when plotting variables ratio (output/input) rather than original data the efficient units can be visualized clearly. The results indicate that: (i) aeration plays an important role in IMBR performance; (ii) implementing the MDS approach with reference to the variables ratio is consequently useful to characterize performance changes for data classification.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2547-2555
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• 2021

The frequency of reactor coolant leakage is expected to increase over the lifetime of a nuclear power plant owing to degradation mechanisms, such as flow-acceleration corrosion and stress corrosion cracking. When loss of coolant accidents (LOCAs) occur, several parameters change rapidly depending on the size and location of the cracks. In this study, leak flow during LOCAs is predicted using a deep fuzzy neural network (DFNN) model. The DFNN model is based on fuzzy neural network (FNN) modules and has a structure where the FNN modules are sequentially connected. Because the DFNN model is based on the FNN modules, the performance factors are the number of FNN modules and the parameters of the FNN module. These parameters are determined by a least-squares method combined with a genetic algorithm; the number of FNN modules is determined automatically by cross checking a fitness function using the verification dataset output to prevent an overfitting problem. To acquire the data of LOCAs, an optimized power reactor-1000 was simulated using a modular accident analysis program code. The predicted results of the DFNN model are found to be superior to those predicted in previous works. The leak flow prediction results obtained in this study will be useful to check the core integrity in nuclear power plant during LOCAs. This information is also expected to reduce the workload of the operators.

• Journal of Welding and Joining
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• 제30권1호
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• pp.51-58
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• 2012

To assess hydraulic connections between sub-components of the International Thermonuclear Experimental Reactor (ITER) diagnostic port plug, the laser welding and ablation cutting process were investigated in order to be applied the remote handling maintenance. In this study, laser ablation cutting, which vaporizes a small amount of solid material directly into gas by focusing a laser beam of high density energy, is adopted in order to overcome the limitation of the normal laser cutting technology that the head should be placed as close to the work piece as possible to blow out melt metal at a distance. Complete cutting of a work piece is obtained by repetitive multi-passes of the laser beam. The welding and cutting process were tested on the sample work pieces and finally on a prototype of a hydraulic connection module for remote handling. The results showed that this process can be a promising candidate for hydraulic connections by remote handling.

• 전기전자학회논문지
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• 제3권1호
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• pp.101-108
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• 1999

본 논문은 금속파편이 발생시 금속파편의 위치를 추정하기위한 위치 추정 알고리즘을 제안한다. 이 추정 알고리즘은 자동화된 알람 구별부와 위치 추정부로 구성된다. 알람 구별부는 충격 신호인지 거짓 충격 신호인지를 구별하는 기능을 수행하며, 위치 추정부는 충격신호로부터 충격신호의 시작점을 찾아내어 위치를 추정한다. 이 알고리즘의 타당성을 검증하기위해 원자로 Mock-up 상에서 실험한 결과 위치 추정값 과 실제 충격값이 약 10 % 이내의 오차율을 보였다.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 1998년도 제13차 학술회의논문집
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• pp.295-300
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• 1998

In this paper, a real time diagnostic algorithm fur estimating the impact location by loose parts is proposed. It is composed of two modules such as the alarm discrimination module (ADM) and the impact-location estimation module(IEM). ADM decides whether the detected signal that triggers the alarm is the impact signal by loose parts or the noise signal. When the decision from ADM is concluded as the impact signal, the beginning time of burst-type signal, which the impact signal has usually such a form in time domain, provides the necessary data fur IEM. IEM by use of the arrival time method estimates the impact location of loose parts. The overall results of the estimated impact location are displayed on a computer monitor by the graphical mode and numerical data composed of the impact point, and thereby a plant operator can recognize easily the status of the impact event. This algorithm can perform the diagnosis process automatically and hence the operator's burden and the possible operator's error due to lack of expert knowledge of impact signals can be reduced remarkably. In order to validate the application of this method, the test experiment with a mock-up (flat board and reactor) system is performed. The experimental results show the efficiency of this algorithm even under high level noise and potential application to Loose Part Monitoring System (LPMS) for improving diagnosis capability in nuclear power plants.

• 시스템엔지니어링학술지
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• 제17권1호
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• pp.21-34
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• 2021

The use of Field-Programmable Gate Arrays (FPGAs) in the development of safety-related Human-Machine Interface (HMI) systems has gained much momentum in nuclear applications. Recently, one of the application areas for the Advanced Power Reactor 1400 (APR1400) is in the development of the advanced Component Interface Module (CIM) of the Engineered Safety Features Actuation System (ESFAS). Using systems engineering approach, we have developed a new FPGA-based advanced CIM software. The first step of our software development process involves the Preliminary Hazard Analysis (PHA) based on the previous CIM design. In this paper, we describe the qualitative approach used in performing the preliminary hazard analysis. The paper presents the methodology for applying a modified Hazard and Operability (HAZOP) procedure for the conduct of PHA which resulted in a qualitative risk-ranking scheme that informed the decisions for the safety criteria in the requirements specification phase. The qualitative approach provided the justification for design changes during the advanced CIM software development process.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4431-4446
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• 2023

A Multiphysics coupling framework, MPCORE, has been developed to analyze safety parameters using the best estimate codes. The framework contains neutron kinetics (NK), thermal hydraulics (TH), and fuel performance (FP) codes to analyze fuel burnup, radial power distribution, and coolant temperature (T_bc). Shuffling and rotation capabilities have been verified on the Watts Bar reactor for three cycles. This study focuses on two coupling approaches for TH and FP modules. The one-way coupling approach involves coupling the FP code with the NK code, providing no data to the TH modules but getting T_bc as boundary condition from TH module. The two-way coupling approach exchanges information from FP to TH modules, so that the simplified heat conduction solver of the TH module is not used. The power profile in both approaches does not differ significantly, but there is an impact on coolant and cladding parameters. The one-way coupling approach tends to over-predict the cladding hydrogen concentration (CHC). This research highlights the difference between one-way and two-way coupling on critical boron concentration, T_bc, CHC, oxide surface temperature, and pellet centerline temperature. Overall, MPCORE framework with two-way coupling provides a more accurate and reliable analysis of safety parameters for nuclear reactors.

• 한국물환경학회지
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• 제21권4호
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• pp.337-346
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• 2005

An MSBR using a membrane for not only filtration but also aeration (MA-MSBR) was designed to reduce membrane fouling and to enhance water quality, and compared with an MSBR using a membrane for only filtration (BA-MSBR). COD removal efficiency of the MA-MSBR was similar to that of the BA-MSBR, but membrane performance of the MA-MSBR was better than that of the BA-MSBR. The MA-MSBR had more small particles in mixed liquor, so the specific cake resistance of flocs in the MA-MSBR was higher than that in the BA-MSBR. However, in the aerobic reaction step of the MA-MSBR, air went through membrane pores and out of the membrane surface, so cake layers on the membrane surface and a portion of organics adsorbed on membrane pores could be removed periodically. Therefore, cake resistance, $R_c$, and fouling resistance by adsorption and blocking, $R_f$, for the MA-MSBR increased more slowly than those for the BA-MSBR. Additionally, in order to compare the energy efficiency for two MSBRs, oxygen transfer efficiency and power to supply air into the reactor by a membrane module and a bubble stone diffuser were measured using deionized water. From these measurements, the transferred oxygen amount per unit energy was calculated, resulting that of MA-MSBR was slightly higher than that of BA-MSBR.

• Nuclear Engineering and Technology
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• 제51권7호
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• pp.1738-1748
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• 2019

Previous research has shown that the inclined condition has an impact on the natural circulation (natural circulation) mode operation of Floating Nuclear Power Plant (FNPP) mounted on the movable marine platform. Due to its compact structure, small volume, strong maneuverability, the Integral Pressurized Water Reactor (IPWR) is adopted as marine reactor in general. The OTSGs of IPWR are symmetrically arranged in the annular region between the reactor vessel and core support barrel in this paper. Therefore, many parallel natural circulation loops are built between the core and the OTSGs primary side when the main pump is stopped. and the inclined condition would lead to discrepancies of the natural circulation drive head among the OTSGs in different locations. In addition, the flow rate and temperature nonuniform distribution of the core caused by inclined condition are coupled with the thermal hydraulics parameters maldistribution caused by OTSG group operating mode on low power operation. By means of the RELAP5 codes were modified by adding module calculating the effect of inclined, heaving and rolling condition, the simulation model of IPWR in inclined condition was built. Using the models developed, the influences on natural circulation operation by inclined angle and OTSG position, the transitions between forced circulation (forced circulation) and natural circulation and the effect on natural circulation operation by different OTSG grouping situations in inclined condition were analyzed. It was observed that a larger inclined angle results the temperature of the core outlet is too high and the OTSG superheat steam is insufficient in natural circulation mode operation. In general, the inclined angle is smaller unless the hull is destroyed seriously or the platform overturn in the ocean. In consequence, the results indicated that the IPWR in the movable marine platform in natural circulation mode operation is safety. Selecting an appropriate average temperature setting value or operating the uplifted OTSG group individually is able to reduce the influence on natural circulation flow of IPWR by inclined condition.