• Nuclear Engineering and Technology
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• 제53권5호
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• pp.1556-1568
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• 2021

The plate-type fuel assembly adopted in nuclear research reactor suffers from complicated effect induced by non-uniform irradiation, which might affect stress conditions, mechanical behaviors and thermal-hydraulic performance of the fuel assembly. This paper is the Part II work of a two-part study devoted to analyzing the complex unique mechanical deformation and thermal-hydraulic characteristics for the typical plate-type fuel assembly under irradiation effect, which is on the basis of developed and verified numerical thermal-fluid-structure coupling methodology under irradiation in Part I of this work. The mechanical deformation, thermal-hydraulic performance and Mises stress have been analyzed for the typical plate-type fuel assembly consisting of support plates under non-uniform irradiation. It was interesting to observe that: the plate-type fuel assembly including the fuel plates and support plates tended to bend towards the location with maximum fission rate; the hot spots in the fuel foil appeared at the location with maximum thickness increment; the maximum Mises stress of fuel foil was located at the adjacent location with the maximum plate thickness increment et al.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.193-198
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• 1996

In this study the RETRAN-03 code was modified to analyze the thermal-hydraulic transients under three-dimensional ship motions for the application to the future marine reactors. First Japanese nuclear ship MUTSU reactor have been analyzed under various ship motions to verify this code. As results, typical thermal-hydraulic characteristics of marine reactors such as flow rate oscillations and S/G water level oscillations are successfully simulated at various conditions.

• 한국초전도저온공학회:학술대회논문집
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• 한국초전도저온공학회 2003년도 학술대회 논문집
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• pp.167-170
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• 2003

The thermal and hydraulic characteristics of the cooling passage for HTS power cable should be carefully investigated to get the highly reliable and economic operation. In this study, the pressure drop and temperature change of the coolant flowing the corrugated passage was estimated using the simple one-dimensional approach, and the temperature distribution in the radial direction was calculated. The results show the dependency of the mass flow rate, thermal conductivities on the cooling performance of the HTS cable.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2534-2546
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• 2023

Gas-cooled space reactor, which adopts He-Xe gas mixture as working fluid, is a better choice for megawatt power generation. In this paper, thermal-hydraulic characteristics of He-Xe gas mixture in 2×2 rod bundle wrapped with helical wires is numerically investigated. The velocity, pressure and temperature distribution of the coolant are obtained and analyzed. The results show that the existence of helical wires forms the vortexes and changes the velocity and temperature distribution. Hot spots are found at the contact corners between helical wires and fuel rods. The highest temperature of the hot spots reach 1600K, while the mainstream temperature is less than 400K. The helical wire structure increases the friction pressure drop by 20%-50%. The effect extent varies with the pitch and the number of helical wires. The helical wire structure leads to the reduction of Nusselt number. Comparing thermal-hydraulic performance ratios (THPR) of different structures, the THPR values are all less than 1. It means that gas-cooled space reactor adopting helical wires could not strengthen the core heat removal performance. This work provides the thermal-hydraulic design basis for He-Xe gas cooled space nuclear reactor.

• 드라이브 ㆍ 컨트롤
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• 제18권3호
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• pp.8-15
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• 2021

A rear axle steering (RAS) system is attached to the rear of medium and large commercial vehicles that transport large cargo. The existing RAS systems are driven by electro-hydraulic actuator (EHA), and most commercialized EHAs consist of electric motors, hydraulic pumps, relief valves, prefill valves and cylinders. The prefill valve required for such EHAs is a type of check valve with extremely low cracking pressure that should not allow RAS to have noise or vibration, and the prefill valve prevents system negative pressure as well as unstable operation. Most papers on this topic rely on experiments to predict valve performance, and theoretically detailed modeling of valves or pipelines is performed, but it is very rare to evaluate hydraulic vibration characteristics by analysing everything from hydraulic pumps to valves comprehensively. In this study, we proposed an experimental circuit that can predict the performance of the prefill valve. The study also analysed the pressure-flow pulsation that is transmitted to the valve through the pipeline, and how the transmitted pressure-flow pulsation affects the valve vibration.

• 대한기계학회논문집B
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• 제28권11호
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• pp.1322-1331
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• 2004

This paper presents studies on the design of water hydraulic system and components to replace oil with tap-water as the pressure transmission medium in hydraulic systems. In order to improve the performance of water hydraulic system, the thermal and hydraulic properties of tap-water are first investigated. Based on these characteristics, the design parameters, such as the clearances of the moving parts, the cross-sectional area of pipes and relative roughness, are proposed so that the performance of water hydraulic system is the same as that of oil. In addition, the operating ranges, which show the possibility of using water hydraulic system, are examined.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.19-26
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• 2020

Small nuclear reactor features higher power capacity, longer operation life than conventional power sources. It could be an ideal alternative of existing power source applied for special equipment for terrestrial or underwater missions. In this paper, a 25kWe heat pipe cooled reactor power source applied for multiple use is preliminary designed. Based on the design, a thermal-hydraulic analysis code for heat pipe cooled reactor is developed to analyze steady and transient performance of the designed nuclear reactor. For reactor design, UN fuel with 65% enrichment and potassium heat pipes are adopted in the reactor core. Tungsten and LiH are adopted as radiation shield on both sides of the reactor core. The reactor is controlled by 6 control drums with B₄C neutron absorbers. Thermoelectric generator (TEG) converts fission heat into electricity. Cooling water removes waste heat out of the reactor. The thermal-hydraulic characteristics of heat pipes are simulated using thermal resistance network method. Thermal parameters of steady and transient conditions, such as the temperature distribution of every key components are obtained. Then the postulated reactor accidents for heat pipe cooled reactor, including power variation, single heat pipe failure and cooling channel blockage, are analyzed and evaluated. Results show that all the designed parameters satisfy the safety requirements. This work could provide reference to the design and application of the heat pipe cooled nuclear power source.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3140-3153
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• 2022

A numerical feasibility study was conducted to investigate the thermal-hydraulic characteristics of a steam generator with corrugated plates for a small modular reactor. Accordingly, a one-dimensional thermal-hydraulic analysis code was developed based on the existing state-of-the-art thermal-hydraulic models and correlations for corrugated plate heat exchangers. Subsequently, the pressure loss, heat transfer, and instability characteristics of the steam generator with corrugated plates were investigated according to the chevron angle and mass flux. Additionally, the characteristics of rectangular and disk-type corrugated plate steam generators with equivalent heat transfer areas were analyzed. The steam generator with disk-type corrugated plates exhibited better performance in terms of pressure loss and heat transfer rate than the rectangular type. In addition, when the mass flux decreased from the onset of boiling points, reverse gradients of the total pressure change were observed in both types. Thus, it was confirmed that Ledinegg instability could occur in the steam generator with corrugated plates. However, it was dependent on the chevron angle, and the optimal chevron angle to minimize instability was 45° under the conditions of the present analysis.

• 한국신재생에너지학회:학술대회논문집
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• 한국신재생에너지학회 2005년도 춘계학술대회
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• pp.697-700
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• 2005

Aquifer Thermal Energy Storage (ATES) system can be very cost-effective and renewable energy sources, depending on site-specific parameters and load characteristics. In order to develop an ATES system which has certain hydrogeological characteristics, understanding of the thermo hydraulic processes of an aquifer is necessary for a proper design of an aquifer heat storage system under given conditions. The thermo hydraulic transfer for heat storage is simulated using FEFLOW according to two sets of pumping and waste water reinjection scenarios of heat pump operation in a two layered confined aquifer. In the first set of model, the movement of the thermal front and groundwater level are simulated by changing the locations of injection and pumping well in seasonal cycle. However, in the second set of model the simulation is performed in the state of fixing the locations of pumping and injection well. After 365 days simulation period, the temperature distribution is dominated by injected water temperature and the distance from injection well. The small temperature change is appears on the surface compared to other slices of depth because the first layer has very low porosity and the transfer of thermal energy are sensitive at the porosity of each layer. The groundwater levels and temperature changes in injection and pumping wells are monitored to validate the effectiveness of the used heat pump operation method and the thermal interference between wells is analyzed.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.666-671
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• 1998

Computational fluid dynamics(CFD) code FLUENT[11 was used to simulate the thermal hydraulic processes occuring in conceptual design of the accelerator-driven subcritical reactor(ADSR) liquid lead target. The purpose of the analysis is to investigate the thermal hydraulic characteristics of liquid lead as ADSR target material with various target geometries and injection locations of proton beam. In the calculation analysis, the local temperature of the liquid lead target rises to the boiling temperature very rapidly When the proton beam is injected from the bottom of the target system, the duration time to reach the boiling temperature is longer and the temperature distribution is flatter than other cases.