• 대한인간공학회:학술대회논문집
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• 대한인간공학회 1996년도 추계학술대회논문집
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• pp.47-55
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• 1996

This study developed a comprehensive and easily applicable nuclear reactor control system evaluation method using reactor operators behavioral and mental workload database. A proposed control panel design cycle consists of the 5 steps: (1) finding out inconvenient, erroneous, and mentally stressful factors for the proposed design through evaluative experiments, (2) drafting improved design alternatives considering detective factors found out in the step (1), (3) comparative experiements for the design alternatives, (4) selecting a best design alternative, (5) returning to the step (1) and repeating the design cycle. Reactor operators behavioral and mental workload database collected from evaluative experiments in the step (1) and comparative experiments in the step (3) of the design cycle have a key roll in finding out defective factors and yielding the criteria for selection of the proposed reactor control systems. The behavioral database was designed to include the major informations about reactor operators' control behaviors: beginning time of operations, involved displays, classification of observational behaviors, dehaviors, decisions, involved control devices, classification of control behaviors, communications, emotional status, opinions for man-machine interface, and system event log. The database for mental workload scored from various physiological variables-EEG, EOG, ECG, and respir- ation pattern-was developed to indicate the most stressful situation during reactor control operations and to give hints for defective design factors. An experimental test for the evaluation method applied to the Compact Nuclear Simulator (CNS) installed in Korea Atomic Energy Research Institute (KAERI) suggested that some defective design factors of analog indicators should be improved and that automatization of power control to a target level would give relaxation to the subject operators in stressful situation.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2009년도 춘계학술대회 논문집
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• pp.389-390
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• 2009

• 한국연소학회:학술대회논문집
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• 한국연소학회 2015년도 제51회 KOSCO SYMPOSIUM 초록집
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• pp.93-96
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• 2015

Commercial coal gasifiers typically use entrained flow type reactors, but have unique features in terms of reactor shape, gasifying agent, coal feeding type, ash/slag discharge, and reaction stages. The MHI gasifier is characterized as air-blow dry-feed entrained reactor, which incorporates a short combustion stage at the bottom and a tall gasification stage above. This study investigates the flow and reaction characteristics inside a MHI gasifier by using computational fluid dynamics (CFD) in order to understand its design and operation features. For its pilot-scale system at 200 ton/day capacity, the distribution of coal and air supply between the two reaction stages was varied. It was found that the syngas composition and carbon conversion rate were not significantly influenced by the changes in the distribution of coal and air supply. However, the temperature, velocity and flow pattern changed sensitively to the changes in the distribution of coal and air supply. The results suggest that one key factor to determine the operational ranges of coal and air supply would be the temperature and flow pattern along the narrower wall between the two reaction stages.

• 유기물자원화
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• 제11권4호
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• pp.79-89
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• 2003

본 연구는 생활폐기물을 호기성분해 개념에서 유기물을 분해 안정화시킬 때 적정한 환경인자들을 도출, 검토하기 위해서 가정에서 배출되는 일반생활폐기물을 각각의 비율에 따라 섞은 후 함수율을 55%, 60%로 조절하여 분해 안정화정도를 평가하였다. VS량의 변화는 6번 반응기를 제외하고는 모든 반응기에서 초기에 투입된 음식물량이 모두 분해가 된 것으로 판단되었고, 1, 2, 3, 4번 반응기는 비슷한 분해정도를 보여 비닐/플라스틱의 첨가가 통기개량제로서의 역할을 한 것으로 사료되었다. 반응기별 온도는 모든 반응기의 최고 온도가 약 $57{\sim}59^{\circ}C$정도로 측정되었고, VS량의 변화와 비슷하게 1, 2, 3, 4번 반응기가 최고온도에서 지속기간이 5, 6번 반응기보다는 약 2~4일정도 더 지속되어 분해반응이 더 잘 일어난 것으로 판단되었으며, $CO_2$ 발생량도 온도 변화와 비슷한 경향을 나타내고 있다. 함수율 변화는 미생물의 산화열로 증발된 수분의 원활한 배출이 이루어지지 않은 관계로 수분 감소율은 모든 반응기에서 낮게 측정되었다. pH는 음식물쓰레기로 인하여 초기에는 약산성이었으나, 반응이 진행됨에 따라 pH가 상승하여 일반적인 경향과 비슷하게 측정되었다. 전기전도도와 C/N비는 전체적으로 소폭증가하는 경향을 보였다. 침하량과 무게감소율도 위의 다른 인자들과 비슷한 경향을 보였다. 즉, 1, 3번 반응기와 2, 4번 반응기가 더 큰 침하율과 무게감소율을 보였다.

• Nuclear Engineering and Technology
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• 제49권3호
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• pp.476-483
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• 2017

The decay heat that is produced by nuclear reactor spent fuel must be cooled in a spent fuel storage pool. A wickless heat pipe or a vertical two-phase closed thermosyphon (TPCT) is used to remove this decay heat. The objective of this research is to investigate the thermal performance of a prototype model for a large-scale vertical TPCT as a passive cooling system for a nuclear research reactor spent fuel storage pool. An experimental investigation and numerical simulation using RELAP5/MOD 3.2 were used to investigate the TPCT thermal performance. The effects of the initial pressure, filling ratio, and heat load were analyzed. Demineralized water was used as the TPCT working fluid. The cooled water was circulated in the water jacket as a cooling system. The experimental results show that the best thermal performance was obtained at a thermal resistance of $0.22^{\circ}C/W$, the lowest initial pressure, a filling ratio of 60%, and a high evaporator heat load. The simulation model that was experimentally validated showed a pattern and trend line similar to those of the experiment and can be used to predict the heat transfer phenomena of TPCT with varying inputs.

• 한국압력기기공학회 논문집
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• 제13권1호
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• pp.40-47
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• 2017

In this paper, we investigate the feasibility of a strategy of failure detection and identification. The point of proposed strategy includes a pattern extraction approach for failure identification using Auto-Associative Kernel Regression (AAKR). We consider a simulation data concerning 605 signals of a Generic Pressurized Water Reactor(GPWR). In the application, the reconstructions are provided by a set of AAKR models, whose input signals have been selected by Correlation Analysis(CA) for the identification of the groups. The failure pattern is extracted by analyzing the residuals of observations and reconstructions. We present the possibility of extraction of patterns for six failure.

• 전기학회논문지P
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• 제58권4호
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• pp.397-403
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• 2009

In this paper, an exciter current control of a synchronous generator for ships using a compound type digital automatic voltage regulator (DVAR) in order to provide a constant output voltage of the generator is presented. The compound type DAVR is composed of a controller part to adjust output voltage and an power source unit to supply power to the exciter. The controller part, which generates the PWM switching pattern via the PI controller, drives a power MOSFET for bypass to limit the SG's exciter current. The power source unit part is parallel connected to an output terminal of the generator through a reactor and a power CT. The residual magnetic flux of SG provides exciter current to the exciter through the reactor during the initial running or no load state and load current supplies field current to the exciter through the power CT during loading state. This paper confirmed an experiment to verify the validity of compound type DAVR system for controlling output voltage of synchronous generator.

• Nuclear Engineering and Technology
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• 제37권6호
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• pp.575-586
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• 2005

In a CANDU (CANada Deuterium Uranium) reactor, fuel channel integrity depends on the coolability of the moderator as an ultimate heat sink under transient conditions such as a loss of coolant accident (LOCA) with coincident loss of emergency core cooling (LOECC), as well as normal operating conditions. This study presents assessments of moderator thermal-hydraulic characteristics in the normal operating conditions and one transient condition for CANDU-6 reactors, using a general purpose three-dimensional computational fluid dynamics code. First, an optimized calculation scheme is obtained by many-sided comparisons of the predicted results with the related experimental data, and by evaluating the fluid flow and temperature distributions. Then, using the optimized scheme, analyses of real CANDU-6 in normal operating conditions and the transition condition have been performed. The present model successfully predicted the experimental results and also reasonably assessed the thermal-hydraulic characteristics of a real CANDU-6 with 380 fuel channels. A flow regime map with major parameters representing the flow pattern inside a calandria vessel has also proposed to be used as operational and/or regulatory guidelines.

• 대한전기학회논문지
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• 제43권5호
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• pp.802-808
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• 1994

The purpose of this paper is to describe an application of plasma polymerized thin film as an electron beam resist. Plasma polymerized thin film was prepared using an interelectrode capacitively coupled gas-flow-type reactor, and chosen methylmethacrylate(MMA)and methylmethacrylate-tetrameth-yltin(MMA-TMT) as a monomer. This thin films were also delineated by the electron-beam apparatus with an acceleration voltage of 30kV and an expose dose ranging from 20 to 900$\mu$C/cmS02T. The delineated pattern in the resist was developed with the same reactor which is used for polymerization using an argon as etching gas. The growth rate and etching rate of the thin film is increased with increasing of discharge power. Thin films by plasma polymerization show polymerization rate of 30~45($\pm$3) A/min, and etching rate of 440($\pm$30) A/min during Ar plasma etching at discharge power of 100W. In apparently lower than that of conventional PMMA, but the plasma-etching rate of PP(MMA-TMT) was higher than that of PPMMA.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.907-915
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• 2024

Various core designs with multi-batch fuel management (FM) are proposed and optimized for an innovative small modular reactor (iSMR), focusing on enhancing the inherent safety and neutronic performance. To achieve soluble-boron-free (SBF) operation, cylindrical centrally-shielded burnable absorbers (CSBAs) are utilized, reducing the burnup reactivity swing in both two- and three-batch FMs. All 69 fuel assemblies (FAs) are loaded with 2-cylindrical CSBA. Furthermore, the neutron economy is improved by deploying a truly-optimized PWR (TOP) lattice with a smaller fuel radius, optimized for neutron moderation under the SBF condition. The fuel shuffling and CSBA loading patterns are proposed for both 2- and 3-batch FM with the aim to lower the core leakage and achieve favorable power profiles. Numerical results show that both FM configurations achieve a small reactivity swing of about 1000 pcm and the power distributions are within the design criteria. The average discharge burnup in the two-batch core is comparable to three-batch commercial PWR like APR-1400. The proposed checker-board CR pattern with extended fingers effectively assures cold shutdown in the two-batch FM scenario, while in the three-batch FM, three N-1 scenarios are failed. The whole evaluation process is conducted using Monte Carlo Serpent 2 code in conjunction with ENDF/B-VII.1 nuclear library.