• 공업화학
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• 제10권3호
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• pp.467-476
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• 1999

공기에 의한 n-butane의 산화로부터 무수마레인산을 합성하는 고정층 촉매 반응기의 거동을 조사하였다. 대류-확산-반응기구로 묘사되는 고정층 촉매반응기의 거동은 Langmuir-Hinshelwood형의 반응속도식 및 비정상상태 이차원 유사균일상 모델을 적용 조사하였다. 예측모델은 Sharma의 pilot-plant 실험 결과인 단일층 반응기의 축방향 온도 및 수율 분포에 대한 최적적합을 통한 최적매개변수 추정에 의해 구성하였다. 또한 예측모델은 단일층 반응기와 통일한 수율 및 전화율을 생성시킬 수 있도록 모사된 불균열활성의 이중층 반응기가 열점에서 $8.96^{\circ}C$ 낮은 온도 상승을 일으켰다. 단일과 이중층 반응기의 가능한 조업조건 (냉매온도, 반응물의 농도, 온도 및 유량)변화에 대한 매개변수 감응도를 조사한 결과 동일한 조업 조건하에서 이중층 반응기가 단일층 반응기에 비해 더 넓은 조업범위는 물론 전화율 및 수율이 다소 높게 나타났다.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2789-2802
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• 2020

Small Modular Reactors (SMRs) are attracting wide attention due to their outstanding performance, extensive studies have been carried out for lead-based fast reactors (LFRs) that cooled with Lead or Lead-bismuth (LBE), and small modular natural circulation LFR is one of the promising candidates for SMRs and LFRs development. One of the challenges for the design small modular natural circulation LFR is to master the natural circulation thermal-hydraulic performance in the reactor primary circuit, while the natural circulation characteristics is a coupled thermal-hydraulic problem of the core thermal power, the primary loop layout and the operating state of secondary cooling system etc. Thus, accurate predicting the natural circulation LFRs thermal-hydraulic features are highly required for conducting reactor operating condition evaluate and Thermal hydraulic design optimization. In this study, a thermal-hydraulic analysis code is developed for small modular natural circulation LFRs, which is based on several mathematical models for natural circulation originally. A small modular natural circulation LBE cooled fast reactor named URANUS developed by Korea is chosen to assess the code's capability. Comparisons are performed to demonstrate the accuracy of the code by the calculation results of MARS, and the key thermal-hydraulic parameters agree fairly well with the MARS ones. As a typical application case, steady-state analyses were conducted to have an assessment of thermal-hydraulic behavior under nominal condition, and several parameters affecting natural circulation were evaluated. What's more, two characteristics parameters that used to analyze natural circulation LFRs natural circulation capacity were established. The analyses show that the core thermal power, thermal center difference and flow resistance is the main factors affecting the reactor natural circulation. Improving the core thermal power, increasing the thermal center difference and decreasing the flow resistance can significantly increase the reactor mass flow rate. Characteristics parameters can be used to quickly evaluate the natural circulation capacity of natural circulation LFR under normal operating conditions.

• 동력기계공학회지
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• 제11권3호
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• pp.3-8
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• 2007

The emergency AC power supply system of the nuclear power plant is designed to supply the power to the nuclear reactor at the emergency operating condition. The safety function of the diesel generator at the nuclear power plant is to supply AC electric power to the plant safety system whenever the preferred AC power supply is unavailable. The reliable operation of onsite emergency diesel generator should be ensured by a conditioning monitoring system designed to maintain and monitor and forecast the reliability level of diesel generator. To do this kind of diesel generator condition monitoring we reviewed several operating factors and history of the wolsong unit 3 diesel generator and selected the proper conditioning monitoring operating factors.

• 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.

• 제어로봇시스템학회:학술대회논문집
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• 제어로봇시스템학회 2002년도 ICCAS
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• pp.69.5-69
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• 2002

1. Introduction 2. System Description 3. Hybrid Modeling of SM Reactor 4. Simulation Results 5. Optimization of Operating Condition 6. Conculsion

• Nuclear Engineering and Technology
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• 제19권3호
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• pp.180-185
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• 1987

월성로형 핵연료봉의 정상상태 노내거동 분석용 전산코드인 KAFEPA를 개발하였다. KAFEPA 전산코드는 같은 목적하에서 AECL에 의해 개발되었던 ELESIM 전산코드에 대응하지만, KAPEPA 전산코드는 ELESIM에 비해 보다 이론적이고 정확하게 예측하는 계산모형들, 즉, 핵분열기체방출 모형, 노내고밀화모형 및 중성자속 감소 계산모형들을 내포하고 있다. KAFEPA 전산코드는 핵분열생성물 기체 방출에 대한 22개 노내 실험자료에 그 예측치를 비교함으로써 검증되었다. KAFEPA 전산코드에 의한 예측치는 상기 실험자료와 잘 일치하였다.

• 에너지공학
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• 제25권1호
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• pp.15-22
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• 2016

월성 2,3,4호기의 비정상 운전 중 열수송계통의 과도변화해석이 수행되었다. 중수로에 대한 캐나다의 규제문서인 AECB R-77 요구조건에 대한 만족성을 평가하였다. 해석 결과 여러 비정상 운전시 과도변화에 의한 원자로 모관의 최고압력값은 ASME 코드의 제한치 이내로 만족되었다. 고압시 보호장치인 LRV의 영향은 미미한 것으로 나타났다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2015년도 제46회 하계학술대회
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• pp.1114-1115
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• 2015

This paper describes the design of toroid-type HTS DC reactor magnet. Target operating current and inductance of the HTS DC reactor are 1,500 A and 400 mH, respectively. The HTS DC reactors were designed through electromagnetic analysis and 3D CAD program. And, we analyze the operating performance of the Double Pancake Coil module for the 1,500 A, 400 mH HTS DC reactor magnet under the liquid nitrogen condition.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회B
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• pp.2344-2347
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• 2007

To generate negative ion, a small dielectric barrier discharge (DBD) plasma reactor was used in this study and operated by high AC voltage. With increasing of voltage, we can get more negative ions. However unfortunately, if the input voltage is too high, it will also cause formation of ozone which is very harmful to human being health. So the work of finding out the best condition of Voltage and frequency was carried out firstly. After several times of measurement, operating at 20 kHz frequency is the best condition generating high ion concentration without ozone. For the purpose of finding out the best reactor structure, two types of surface dielectric barrier discharge (DBD) reactors were examined to produce negative oxygen ions at the conditions of 20 kHz frequency. The results indicated that the surface DBD reactor with several small tips showed better characteristics for generation of negative oxygen ions at the same condition.

• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.85-92
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• 2016

The operating temperature of VHTR components is much higher than that of conventional PWR due to high core outlet temperature of VHTR. Material requirements and technical issues of VHTR reactor components which are mainly dominated by high temperature service condition were discussed. The codification effort for high temperature material and design methodology are explained. The design class for VHTR components are classified as class A or B according to the recent ASME high temperature reactor design code. A separation of thermal boundary and pressure boundary is used for VHTR components as an elevated design solution. Key design characteristics for reactor pressure vessel, control rod, reactor internals, graphite reflector, circulator and intermediate heat exchanger were analysed. Thermo-mechanical analysis of the process heat exchanger, which was manufactured for test, is presented as an analysis example.