Two-phase pressure drop analysis of swirl tube in one-side high heat load condition for plasma facing component application

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.

Section: Experimental Methods 21 Experimental Loopmentioning

confidence: 99%

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

2021

Eng. Res. Express

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“…This series of processes creates a secondary flow on the wall and promotes heat transfer. Bowring [7] named the secondary flow effect due to the accelerated bubble departure frequency and the collapse of the thermal boundary layer in the FDB regime as agitation heat flux. He derived the equation for the agitation heat flux at the wall, as shown in equation (3) under the assumption that the volume of the bulk fluid flowing into the wall can be treated as a slug equal to the volume of the departed bubble, where N a denotes the active nucleation site density, f the bubble departure frequency, and V b the volume when the bubble departs.…”

Section: Flat Heat Sink Test Sectionmentioning

confidence: 99%

“…In particular, the vertical targets of the divertor, which removes helium ash and facilitates plasma operation, are loaded with an ultra-high heat flux of up to 20 MW m −2 [5]. If such a large energy is transferred to the fluid inside the channel, severe boiling phenomenon occurs, the differential pressure increases [6], and there is a risk of inducing flow instability [7]. However, among the engineering issues caused by the one-side high heat load condition, the concentration of heat flux at the top of the cooling channel can not only lower the critical heat flux, which is the maximum heat flux that the cooling channel can withstand, but also cause material damage such as cracks [8].…”

Section: Introductionmentioning

confidence: 99%

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

2021

Eng. Res. Express

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The heat removal capacity of a flat heat sink was studied using subcooled flow boiling experiments, to address the thermal peaking problem. Based on the Bowring criteria, the boiling curve is divided into a partially developed nucleate boiling regime (PDB) and fully developed nucleate boiling regime (FDB), and the existing heat transfer correlations for each flow regime are evaluated. In the PDB regime, the Baburajan correlation exhibited the highest prediction rate with an average error rate of 10.92%; however, the FDB regime heat transfer correlations exhibited high error rates at very high heat flux conditions. Therefore, the authors developed a new FDB correlation using the artificial intelligence technique by correlating the bubble agitation effect, which is a mechanism of the FDB regime, and then, evaluated the qualification assessment on this basis. The mono-block plasma-facing component with a flat heat sink was found to meet all criteria (except #1.2, shutdown plasma ratcheting), and to succeed in the actual fabrication.

“…Various types of cooling channel structures that can withstand ultrahigh temperatures have been proposed as candidate solutions. The most representative channel structures are the swirl tube [8], hypervapotron channel [9], and screw tube. Screw tubes contain cooling channels with helical nut structures on the flow path wall, providing excellent cooling performance and critical heat flux (CHF) performance through the generation of secondary flow near the flow path wall [10].…”

Section: Introductionmentioning

confidence: 99%

“…Most previous thermo-hydraulic performance evaluation studies of screw tubes assessed CHF performance to determine whether the tube could withstand the ultra-high heat flux within a tokamak. However, the cooling system of a power plant operating under subcooled flow conditions is designed to function in an SP regime, as the vapor produced within a channel under a two-phase regime is a potential hazard in terms of flow instability [18], increased two-phase pressure drop [8,10], and CHF. Figure 2 shows the three flow regimes that occur inside a channel under the subcooled flow boiling and one-side heating conditions.…”

Section: Introductionmentioning

confidence: 99%

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

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.