A numerical-experimental analysis of qualification assessment for mono-block plasma-facing component with flat heat sink

Lim, Ji Hwan; Park, Minkyu

doi:10.1088/2631-8695/ac3b2a

Cited by 15 publications

(7 citation statements)

References 27 publications

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“…A solenoid valve with on-off function is installed under the accumulator to discharge all used water out of the loop after the experiment. Detailed information on the experimental equipment or loop design process can be found in previous papers [39][40][41]. Table 1 summarizes the ranges of the system parameters and uncertainties of the measured variables of the sub-cooled flow boiling experiment conducted in this study.…”

Section: Experimental Methods 21 Test Loopmentioning

confidence: 99%

Enhanced single-phase heat transfer performance via vortex secondary flow in hypervapotron configuration

Lim¹,

Park²

2021

Phys. Scr.

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As the hypervapotron (HV) heat sink is used to cool many areas inside the fusion tokamak, it is essential to understand its heat transfer performance to calculate the thermal efficiency of the power generation system. Therefore, in this study, the single-phase (SP) heat transfer performance of HV heat sink was evaluated through sub-cooled flow boiling experiments under one-side high-heat load conditions. When vapor is generated inside the heat sink, flow instability and a potential risk of reaching the critical heat flux are created. Therefore, in commercial power plants, cooling systems tend to operate in the SP regime. System parameters that can be adjusted in the power generation system include the system pressure, mass flow rate, and subcooling, and the effect of these three parameters on the heat transfer performance in the SP regime was analyzed. It was experimentally observed that the mass flow rate was the most influential variable. The prediction performance of the SP forced convection heat transfer correlations of the existing conventiaonl channel were evaluated. The results revealed that they tended to under-predict the heat transfer performance of the HV heat sink. In addition, the same trends were found when the forced convection heat transfer correlation of the curved channel was evaluated. The reasons for the former and the latter are that the heat transfer enhancement effect by the vortex flow occurring between the fins of the HV heat sink is not reflected in the correlations, and the vortex effect of the HV heat sink is not expressed as a variable. Therefore, a new vortex forced convection heat transfer correlation was developed through the newly defined Dean number of the HV heat sink. The developed correlation recorded an average error rate of 0.48%.

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Section: Experimental Methods 21 Test Loopmentioning

confidence: 99%

Enhanced single-phase heat transfer performance via vortex secondary flow in hypervapotron configuration

Lim¹,

Park²

2021

Phys. Scr.

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“…Additionally, an AlN plate with a very low HTC was installed on top of the FeCrAl heater to prevent heat loss over the channel. Detailed information on the heater development process and material selection can be found in a previous papers [10,20,23].…”

Section: One-side Joule Heating Assemblymentioning

confidence: 99%

“…A rechecking of the 5% of outlier data left after the development process confirmed that most were the result of incorrect machining of the channel's thermocouple holes or measurement errors by other measuring equipment. More detailed information about Python code can be found from previous papers [10,20,23,56].…”

Section: Python Code Developmentmentioning

confidence: 99%

Study on onset of nucleate boiling with screw tube for fusion reactor application

Lim

Park

2022

Nucl. Fusion

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The onset of nucleate boiling (ONB) is the point at which the heat transfer mechanism in fluids changes and is one of the thermo-hydraulic factors that must be considered when establishing a cooling system operation strategy. Because the high heat flux of several MW/m2, which is loaded within a tokamak, is applied under a one-side heating condition, it is necessary to determine a correlative relation that can predict ONB under special heating conditions. In this study, the ONB of a one-side-heated screw tube was experimentally analyzed via a subcooled flow boiling experiment. The helical nut structure of the screw tube flow path wall allows for improved heat transfer performance relative to smooth tubes, providing a screw tube with a 53.98% higher ONB than a smooth tube. The effects of the system parameters on the ONB heat flux were analyzed based on the changes in the heat transfer mechanism, with the results indicating that the flow rate and degree of subcooling are proportional to the ONB heat flux because increasing these factors improves the forced convection heat transfer and increases the condensation rate, respectively. However, it was observed that the liquid surface tension and latent heat decrease as the pressure increases, leading to a decrease in the ONB heat flux. An evaluation of the predictive performance of existing ONB correlations revealed that most have high error rates because they were developed based on ONB experiments on micro-channels or smooth tubes and not under one-side high heat load conditions. To address this, we used dimensional analysis based on Python code to develop new ONB correlations that reflect the influence of system parameters.

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“…Here, SSE (error sum of squares) is the sum of squares of the difference between the values that are not explained by the regression model and the values predicted by the regression model; SSR (regression sum of squares) is the sum of squares of the mean differences between the values predicted by the regression model and the observations; and SST (total sum of squares) is the total variation or the sum of SSE and SSR. SST, SSR and SSE can be obtained from (15), (16), and (17), respectively [34][35][36].…”

Section: Assessment Of Existing Ofi Correlationsmentioning

confidence: 99%

Experimental study of thermal limits with one-side heated smooth channel for plasma facing component safety

Lim

Park

2021

Eng. Res. Express

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In order to stably operate the equipment inside the tokamak, which is loaded with a heat flux of several MW/m2 under the one-side heating condition, it is necessary to thoroughly prepare for various thermal engineering limits that may occur under the high heat flux load condition. In this study, we have experimentally explored critical heat flux (CHF) and onset of flow instability (OFI), which are considered potential threats in a DEMO fusion power plant. Specifically, the effect of system parameters on CHF was investigated. The results indicate that with an increase in subcooling and mass flux, the CHF increased, as it induced a faster bubble condensation near the CHF. As the system pressure increased, the CHF also increased. This is because the bubble size reduction effect was dominant in the pressure range of 1–10 bar. Most of the existing CHF correlations could evaluate the CHF with reasonable accuracy of within 25%; especially, the Boscary CHF correlation yielded the highest accuracy with an average error of 12%. Similar to CHF, OFI, which is a measure of the sudden fluctuations in the system pressure caused by a large amount of vapor generated due to the high heat flux, tended to increase as the subcooling, mass flow rate, and system pressure increased. Most of the existing OFI correlations yielded large error rates (more than 135%) as these correlations were primarily developed for micro-channels. Therefore, in this study, a new OFI correlation was developed using a Python code, in combination with an artificial intelligence (AI) regression method. The developed correlation can be used in the cooling system design of tokamaks, which involve a high-heat load condition on one-side of the reactor.

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A numerical-experimental analysis of qualification assessment for mono-block plasma-facing component with flat heat sink

Cited by 15 publications

References 27 publications

Enhanced single-phase heat transfer performance via vortex secondary flow in hypervapotron configuration

Enhanced single-phase heat transfer performance via vortex secondary flow in hypervapotron configuration

Study on onset of nucleate boiling with screw tube for fusion reactor application

Experimental study of thermal limits with one-side heated smooth channel for plasma facing component safety

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