• 한국수소및신에너지학회논문집
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• 제13권2호
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• pp.119-126
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• 2002

The methanol partial oxidation using commercial $CuO/ZnO/Al_2O_3$ catalysts in a plug flow reactor was studied in the temperature range of $200{\sim}250^{\circ}C$ at atmospheric pressure, It was achieved the high activities by Cu-based catalysts and the selectivity of $CO_2$/$H_2$ was 100% when $O_2$ was fully convened. The reactivity changes and their hysteresis with increasing/decreasing temperatures were observed due to the chemical state differences between the oxidation and the reduction on the Cu surface, It was suggested as the two-step reaction: the complete oxidation and the following steam reforming for methanol, which was indicated by the distributions of final products vs. the residence time. In addition, the complete oxidation step was shown to be extremely fast and the total reaction rate can be controlled by the steam reforming reaction.

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.1022-1035
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• 2016

A double-tube once-through steam generator (DOTSG) consisting of an outer straight tube and an inner helical tube is studied in this work. First, the structure of the DOTSG is optimized by considering two different objective functions. The tube length and the total pressure drop are considered as the first and second objective functions, respectively. Because the DOTSG is divided into the subcooled, boiling, and superheated sections according to the different secondary fluid states, the pitches in the three sections are defined as the optimization variables. A multi-objective optimization model is established and solved by particle swarm optimization. The optimization pitch is small in the subcooled region and superheated region, and large in the boiling region. Considering the availability of the optimum structure at power levels below 100% full power, we propose a new operating scheme that can fix the boundaries between the three heat-transfer sections. The operation scheme is proposed on the basis of data for full power, and the operation parameters are calculated at low power level. The primary inlet and outlet temperatures, as well as flow rate and secondary outlet temperature are changed according to the operation procedure.

• 한국압력기기공학회 논문집
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• 제19권2호
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• pp.61-68
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• 2023

The phenomenon of fretting wear due to the flow-induced vibration in steam generator (SG) tube is a significant degradation mechanism in nuclear power plants. Fretting wear in SG tube is primarily attributed to the friction and impact forces between the SG tube and the tube support structures, experienced during nuclear power plants operation. While the Archard model has generally been used for the prediction of fretting wear in SG tube, it is limited by its linear nature. In this study, we introduced an "Impact Shear Work-rate" (ISW) model, which takes into account the combined effects of impact and sliding. The ISW model was evaluated using existing experimental data on fretting wear in SG tube and was compared against the Archard model. The prediction results using the ISW model were more accurate than those using the Archard model, particularly for impact forces.

• 한국압력기기공학회 논문집
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• 제19권2호
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• pp.109-116
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• 2023

This study presents a CRUD modelling to simulate the thermal resistance behavior of CRUD, deposited on the surface of a cladding tube of a fuel assembly. When heat produced from fuels transfers to a coolant through a cladding tube, the CRUD acting as an additional thermal resistance is expressed as two layers, i.e., a solid oxide layer and an imaginary fluid layer, which are added to the experimental tube's heat structure of the MARS-KS input data. The validation calculation for the experiments performed in UNIST-DISNY facility showed that the center and surface temperatures of the cladding tube increased as the porosity and the steam amount inside pores of the CRUD got higher. In addition, the temperature gradient in the imaginary fluid layer was calculated to be larger than that in the solid oxide part, indicating that the steam amount inside the layer acted more largely as thermal resistance. It was also evaluated through sensitivity calculations that the cladding tube temperature was more sensitive to the CRUD porosity and the steam amount in pores than to the inlet flow rate of the coolant.

• 한국추진공학회지
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• 제21권1호
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• pp.72-83
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• 2017

본 연구에서는 유도탄 사출관 내부의 수치모사를 통해 이상 유동에 대한 열 유체역학적 분석을 수행하였다. 고정된 해석영역에서 계산이 진행되었고 증발이 완료된 물을 냉각제로 사용하였다. 고온의 공기와 냉각제간의 상호작용 및 유동장을 해석하기 위해, Realizable $k-{\varepsilon}$ 난류 모델과 VOF (Volume Of Fluid) 모델을 선정하고 냉각제 유량에 따른 수치 해석을 진행하였다. 해석결과, 사출관 상부 압력은 냉각제 유량에 따라 비선형적으로 증가하였다. 그리고 내부에서의 유동 진행 과정과 온도분포, 냉각제분포가 밀접한 연관이 있음을 확인하였다. 사출관 하부의 초기 온도는 냉각제량의 증가에 비례하여 감소하지만, 특정시간 이후 경향이 역전되면서 오히려 온도의 상승을 유발하였다. 또한, 혼합가스의 순환유동에 의해 초기의 온도변화가 요동하는 경향도 확인되었다.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 2001년도 가을 학술발표회 논문집
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• pp.109-114
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• 2001

In this study, the physical and dynamic characteristics of Bottom-ash produced at steam power plants were compared and considered. The comparative objects were Bottom-ash in which a lot of powder contained and that in which less than that relatively contained. The difference in quantity of powder showed different effect on the character of flow. This study was undertaken on the use of Bottom-ash as a fine aggregate, and showed the durability and the character of strength according to each rate.

• 한국안전학회지
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• 제35권4호
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• pp.84-91
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• 2020

The Korean nuclear industry developed the SPACE (Safety and Performance Analysis Code for nuclear power plants) code and this code adpots two-phase flows, two-fluid, three-field models which are comprised of gas, continuous liquid and droplet fields and has a capability to simulate three-dimensional model. According to the revised law by the Nuclear Safety and Security Commission (NSSC) in Korea, the multiple failure accidents that must be considered for accident management plan of nuclear power plant was determined based on the lessons learned from the Fukushima accident. Generally, to improve the reliability of the calculation results of a safety analysis code, verification work for separate and integral effect experiments is required. In this reason, the goal of this work is to verify calculation capability of SPACE code for multiple failure accident. For this purpose, it was selected the experiment which was conducted to simulate a Multiple Steam Generator Tube Rupture(MSGTR) accident with Passive Auxiliary Feedwater System(PAFS) operation by Korea Atomic Energy Research Institute (KAERI) and focused that the comparison between the experiment results and code calculation results to verify the performance of the SPACE code. The MSGR accident has a unique feature of the penetration of the barrier between the Reactor Coolant System (RCS) and the secondary system resulting from multiple failure of steam generator U-tubes. The PAFS is one of the advanced safety features with passive cooling system to replace a conventional active auxiliary feedwater system. This system is passively capable of condensing steam generated in steam generator and feeding the condensed water to the steam generator by gravity. As the results of overall system transient response using SPACE code showed similar trends with the experimental results such as the system pressure, mass flow rate, and collapsed water level in component. In conclusion, it could be concluded that the SPACE code has sufficient capability to simulate a MSGTR accident.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2699-2708
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• 2020

Helical-coil steam generator (HCSG) technology is a major design candidate for small modular reactors due to its compactness and capability to produce superheated steam with high generation efficiency. In this paper, we investigate the feasibility of the passively autonomous power maneuvering by coupling the 3-D transient multi-physics of a soluble-boron-free (SBF) core with a time-dependent HCSG model. The predictor corrector quasi-static method was used to reduce the cost of the transient 3-D neutronic solution. In the numerical system simulations, the feedwater flow rate to the secondary of the HCSGs is adjusted to extract the demanded power from the primary loop. This varies the coolant temperature at the inlet of the SBF core, which governs the passively autonomous power maneuvering due to the strongly negative coolant reactivity feedback. Here, we simulate a 100-50-100 load-follow operation with a 5%/minute power ramping speed to investigate the feasibility of the passively autonomous load-follow in a 450 MW_th SBF PWR. In addition, the passively autonomous frequency control operation is investigated. The various system models are coupled, and they are solved by an in-house Fortran-95 code. The results of this work demonstrate constant steam temperature in the secondary side and limited variation of the primary coolant temperature. Meanwhile, the variations of the core axial shape index and the core power peaking are sufficiently small.

• 설비공학논문집
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• 제8권2호
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• pp.187-197
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• 1996

A computer program for thermal design analysis has been developed to predict the operating characteristics and performance of an absorption heat pump to recover $30{\sim}40^{\circ}C$ of waste hot water. The effects of heat transfer area of the system components, temperature and mass flow rate of heat transfer medium, and solution circulation rate on the system performance are investigated in detail. The results obtained indicate that the COP is increased with a decrease in the temperature of driving steam and with an increase in the temperature of waste hot water while the COP is little affected by the variation of a hot water temperature. It is also found that the heating output is increased with an increase in the temperature of waste hot water and driving steam as well as with a decrease in the temperature of hot water. The simulation results are also compared with the experimental results for a periodic operation of the system and obtained a satisfactory agreement.

• Corrosion Science and Technology
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• 제8권3호
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• pp.126-131
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• 2009

The interaction between wear and corrosion can significantly increase total material losses in water chemistry environment. The corrosive wear tests of a PWR steam generator tube material (Alloy 690) against the anti vibration bar material (409 SS) were performed at room temperature. The tests were performed in alkaline water chemistry conditions. NaOH solution was selected for test condition to investigate the corrosive wear effect of steam generator tube material in alkaline pH condition without other factors. The flow induced vibration can caused tube damage and the corrosion can be occurred by water chemistry. The test results showed that, in the alkaline solution at pH 13.9, the corrosion current density was increased about ten times than that in the distilled water. And wear rate at pH 13.9 was increased about ten times from that at neutral condition. However, the wear rate was decreased with time. The decrease would be attributed to the change in roughness of specimen or sub-layer of the worn surface with time. From microstructure observation, severe abrasive shape and several wear debris were found. From those results, it could infer that the oxide film on Alloy 690 changed to easily breakable one in the alkaline water, and then abrasion with corrosion became the main wear mechanism.