• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2771-2788
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• 2020

Experiments are conducted to study bubble flow behavior during the instability of subcooled boiling under uniform and non-uniform transverse heating. The non-uniform heat distribution introduces nonuniform bubble generation and condensation rates on the heated surface, which is different from the uniform heating. These bubble generation and condensation characteristics introduce a non-uniform local pressure distribution in the transverse direction, which creates an extra non-uniform pressure on the flowing bubbles. Therefore, different bubble flow behavior can be observed between uniform and non-uniform heating conditions. In the uniform heating, bubble velocity fluctuations are low, and the bubbles travel straight along the axial direction. In the non-uniform heating, more fluctuation in the bubble velocity occurs at low mass flow rate and high subcooled inlet temperatures, and reverse flow is observed. Additionally, the bubbles show a zigzag trajectory when they pass through the channel, which indicates the existence of cross flow in the transverse direction.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.1126-1135
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• 2022

The MARS code has been assessed for the prediction of onset of flow instability (OFI) in a vertical channel. For assessment, we built an experiment database that consists of experiments under various geometry and thermal-hydraulic condition. It covers pressure from 0.12 to 1.73 MPa; heat flux from 0.67 to 3.48 MW/m²; inlet sub-cooling from 39 to 166 ℃; hydraulic diameters between 2.37 and 6.45 mm of rectangular channels and pipes. It was shown that the MARS code can predict the OFI mass flux for pipes reasonably well. However, it could not predict the OFI in a rectangular channel well with a mean absolute percentage error of 8.77%. In the cases of rectangular channels, the error tends to depend on the hydraulic diameter. Because the OFI is directly related to the subcooled boiling in a flow channel, we suggest a modified subcooled boiling model for better prediction of OFI in a rectangular channel; the net vapor generation (NVG) model and the modified wall evaporation model were modified so that the effect of hydraulic diameter and heat flux can be accurately considered. The assessment of the modified model shows the prediction of OFI mass flux for rectangular channels is greatly improved.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3635-3642
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• 2021

Natural circulation systems (NCSs) are extensively applied in nuclear power plants because of their simplicity and inherent safety features. For some passive natural circulation systems in floating nuclear power plants (FNPPs), the ocean is commonly used as the heat sink. Condensation induced water hammer (CIWH) events may appear as the steam directly contacts the subcooled seawater, which seriously threatens the safe operation and integrity of the NCSs. Nevertheless, the research on the formation mechanisms of CIWH is insufficient, especially in NCSs. In this paper, the characteristics of flow rate and fluid temperature are emphatically analyzed. Then the formation types of CIWH are identified by visualization method. The experimental results reveal that due to the different size and formation periods of steam slugs, the flow rate presents continuous and irregular oscillation. The fluid in the horizontal hot pipe section near the water tank is always subcooled due to the reverse flow phenomenon. Moreover, the transition from stratified flow to slug flow can cause CIWH and enhance flow instability. Three types of formation mechanisms of CIWH, including the Kelvin-Helmholtz instability, the interaction of solitary wave and interface wave, and the pressure wave induced by CIWH, are obtained by identifying 67 CIWH events.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1851-1859
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• 2022

Experimental research on flow characteristics of open natural circulation system was performed, to figure out the mechanism of the open natural circulation behaviors. The influence factors, such as the heating power, the inlet subcooled and the level of cooling tank on the flow characteristics of the system were examined. It was shown that within the scope of the experimental conditions, there are five flow types: single-phase stable flow, flash and geyser coexisting unstable flow, flash stable flow, flash unstable flow, and flash and boiling coexisting unstable flow. The geyser flow in flash and geyser coexisting unstable flow is different from classic geysers flow. The flow oscillation period and amplitude of the former are more regular, is a newly discovered flow pattern. By drawing the flow instability boundary diagram and sorting out the flow types, it is found that the two-phase unstable flow is mainly characterized by boiling and flash, which determine the behavior of open natural circulation respectively or jointly. Moreover, compared with full liquid level system, non-full liquid level system is more prone to boiling phenomenon, and the range of heat flux density and undercooling degree corresponding to unstable flow is larger.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2276-2296
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• 2022

This study validates the applicability of the CUPID code for simulating subcooled wall boiling under high-pressure conditions against number of DEBORA tests. In addition, a new numerical technique in which the interfacial momentum non-drag forces are calculated at the cell faces rather than the center is presented. This method reduced the numerical instability often triggered by calculating these terms at the cell center. Simulation results showed good agreement against the experimental data except for the bubble sizes in the bulk. Thus, a new model to calculate the Sauter mean diameter is proposed. Next, the effect of the relationship between the bubble departure diameter (D_dep) and the nucleation site density (N) on the performance of the Wall Heat Flux Partitioning (WHFP) model is investigated. Three correlations for D_dep and two for N are grouped into six combinations. Results by the different combinations show that despite the significant difference in the calculated D_dep, most combinations reasonably predict vapor distribution and liquid temperature. Analysis of the axial propagations of wall boiling parameters shows that the N term stabilizes the inconsistences in D_dep values by following a behavior reflective of D_dep to keep the total energy balance. Moreover, ratio of the heat flux components vary widely along the flow depending on the combinations. These results suggest that separate validation of D_dep correlations may be insufficient since its performance relies on the accompanying N correlations.