• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.677-690
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• 2009

A previous study demonstrated that the two-row fuel assembly has much more favorable neutron-physical and thermal-hydraulic behavior than the conventional one-row fuel assemblies. Based on the newly developed two-row fuel assembly, an SCWR core is proposed and analyzed. The performance of the proposed core is investigated with 3-D coupled neutron-physical and thermal-hydraulic calculations. During the coupling procedure, the thermal-hydraulic behavior is analyzed using a sub-channel analysis code and the neutron-physical performance is computed with a 3-D diffusion code. This paper presents the main results achieved thus far related to the distribution of some neutronic and thermal-hydraulic parameters. It shows that with adjustment of the coolant and moderator mass flow in different assemblies, promising neutron-physical and thermal-hydraulic behavior of the SCWR core is achieved. A sensitivity study of the heat transfer correlation is also performed. Since the pin power in fuel assemblies can be non-uniform, a sub-channel analysis is necessary in order to investigate the detailed distribution of thermal-hydraulic parameters in the hottest fuel assembly. The sub-channel analysis is performed based on the bundle averaged parameters obtained with the core analysis. With the sub-channel analysis approach, more precise evaluation of the hot channel factor and maximum cladding surface temperature can be achieved. The difference in the results obtained with both the sub-channel analysis and the fuel assembly homogenized method confirms the importance of the sub-channel analysis.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.641-642
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• 2005

• Journal of computational fluids engineering
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• v.7 no.2
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• pp.43-52
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• 2002

This paper presents the numerical methodology of ATHOS3 code for thermal hydraulic analysis of steam generators in nuclear power plant. Topics include porous media approach, governing equations, physical models and correlations for solid-to-fluid interaction and heat transfer, and numerical solution scheme. The ATHOS3 code is applied to the thermal hydraulic analysis of steam generator in the Korea Kori Unit-1 nuclear power plant and the computed results are presented

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.596-597
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• 2005

• Nuclear Engineering and Technology
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• v.48 no.2
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• pp.376-385
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• 2016

A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20-115% of the nominal flow rate and $60^{\circ}C$ (equivalent to Re ~ 37,100) at the inlet of the test rig. The pressure loss data in three measured subchannels were almost identical regardless of the subchannel locations. The flow rate at each type of subchannel was identified and the flow split factors were evaluated from the measured data. The predicted correlations and the computational fluid dynamics results agreed reasonably with the experimental data.

• Design & Manufacturing
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• v.12 no.1
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• pp.41-46
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• 2018

This study proves the causes of cylinder and piston jam by scratches which is the fatal problem of hydraulic breaker through the thermal analysis and thermal-structural coupled field analysis. The trouble from the scratch is a complex problem which can be caused by manufacturing process (this is an internal factor) and the users mistake or contamination in the hydraulic circuit (these are an external factor). Hence, it's not easy to investigate the causes, also hard to prevent the recurrence. In this reason, hydraulic breaker manufacturers are trying to improve the manufacturing process such as machining, heat treatment, grinding, cleaning, also to prevent the contamination in hydraulic circuit and to remove the remains. It's being managed thoroughly by manufacturers. This study shows the effect of the temperature rise by the frictional heat generated when the piston hits the tool on the hydraulic oil while the hydraulic breaker is operating, also the temperature distribution when it starts to affect main components of hydraulic breaker. The stress and the amount of deformation also could be found through thermal-structural coupled field analysis. It proved that the stress and deformation are proportionally increased according to the temperature rise in hit area, and it affects the cylinder and the viscosity of hydraulic oil inside the cylinder when it heats up beyond the certain temperature.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.313-314
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• 2004

• Nuclear Engineering and Technology
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• v.52 no.12
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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.

• Proceedings of the KSME Conference
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• 2001.06e
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• pp.463-468
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• 2001

This paper presents the methodology for thermal hydraulic analysis of Pressurized Water Reactor (PWR) steam generators. Topics include porous media approach, governing equations, physical models and correlations for solid-to-fluid interaction and heat transfer and numerical solution scheme. Some details about the ATHOS3 code currently used widely for thermal hydraulic analysis of PWR steam generators in the industry are presented. The ATHOS3 code is applied to the thermal hydraulic analysis of steam generator in the Korea YGN 3&4 nuclear power plant and the computed results are presented.

• Journal of the Korean Society of Mechanical Technology
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• v.13 no.4
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• pp.85-91
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• 2011

Experimental analysis has been carried out to investigate thermal characteristics of hydraulic system in special vehicles. Hydraulic system performance is largely influenced by oil temperature, and there are considerable performance decline and malfunctions in the system for high temperature conditions caused by heavy load and continuous operation. Transient oil temperature and pressure variation are analyzed and heat generation rates in the several main system parts are compared for various flow rates. With the start of system operation oil temperature gradually increases, and viscosity deceases by about 70% as temperature increases from $20^{\circ}C$ to $80^{\circ}C$. Operation pressure in the hydraulic system decreases with oil temperature, and its variation rate becomes less steep as oil temperature increases. Heat generation rate in hydraulic pump also depends on the oil temperature, and it reaches maximum near $50^{\circ}C$. These results in this study can be applied to optimal design of efficient hydraulic system in special vehicles.