• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.82-87
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• 1996

To find the effects of pressure on critical heat flux (CHF) for the conditions of low pressures (especially up to 10 bar) and low mass flux (~300 kg/$m^2$s), a series of experiments have been accomplished by using uniformly heated Inconel-625 tube. The experimental ranges are as follows: pressure (from 1.2 to 8 bar). mass velocities (from 100 to 250 kg/$m^2$s) and the inlet subcooling ($\Delta$h$_{i}$ ＝ 350 kJ/kg). According to the experimental data, it is found that the CHF is nearly independent of the pressure and increases with mass flux. From the results of the CHF correlation assessment for this experimental data, we could find somewhat different tendency of CHF behavior from every other CHF prediction correlation and table.ation and table.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.386-391
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• 1997

It is important to understand correctly a CHF under low flow condition for the purpose of enhancing the reactor safety and performance in the LWRs. The CHF experiments have been carried out for an internally heated vertical annulus in RCS loop facility. The experimental conditions cover ranges of pressure from 1.82 to 12.08 MPa, mass flux from 300 to 550kg/$m^2$. s and inlet subcooling of 210kJ/kg. The CHF data decrease with increasing pressure at high value of mass flux. For mass flux of about 300kg/$m^2$. s, the CHF rue little influenced by pressure. The CHF data are correlated well by using the dimensionless heat flux and dimensionless mass flux for a fixed inlet subcooling except the data group of 12.08 MPa. It seems that the Doerffer correlation and Katto correlation overestimate the CHF for low pressure and lower value of mass flux within this experimental ranges. The Bowling correlation gives a better prediction than the other two correlations.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.834-841
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• 2022

System thermal-hydraulic (STH) code is adopted for nuclear safety analysis. The critical flow model (CFM) is significant for the accuracy of STH simulation. To overcome the defects of current CFMs (low precision or long calculation time), a CFM based on a genetic neural network (GNN) has been developed in this work. To build a powerful model, besides the critical mass flux, the critical pressure and critical quality were also considered in this model, which was seldom considered before. Comparing with the traditional homogeneous equilibrium model (HEM) and the Moody model, the GNN model can predict the critical mass flux with a higher accuracy (approximately 80% of results are within the ±20% error limit); comparing with the Leung model and the Shannak model for critical pressure prediction, the GNN model achieved the best results (more than 80% prediction results within the ±20% error limit). For the critical quality, similar precision is achieved. The GNN-based CFM in this work is meaningful for the STH code CFM development.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3962-3973
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• 2022

Several critical heat flux (CHF) correlations including the look-up table in the MARS code have been assessed for the prediction of CHF in a downward-flow narrow rectangular channel. For the assessment, we built an experiment database that covers pressures between 1.01 and 39.0 bar, gap sizes between 1.09 and 6.53 mm, mass fluxes up to 25,772 kg/m²s, and under one-sided and two-sided heating conditions. The results of the assessment showed that the Kaminaga correlation has the best overall prediction compared to others. However, because the correlation uses global variables, such as inlet and outlet subcooling and total heat transfer area, it is difficult to use in a system code. A new CHF correlation is then proposed by replacing the global variables in the Kaminaga correlation with local ones and adding correction factors to consider the effect of gap size, mass flux, and the number of heating walls. Additional correction factor is added to consider the effect of inlet subcooling. It is shown that the new one is better than the Kaminaga correlation and it is easy to implement to any system code.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.2094-2099
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• 2004

An experimental study on heat transfer characteristics near the critical pressure has been performed with an internally-heated vertical annular channel cooled by R-134a fluid. Two series of tests have been completed: (a) steady-state critical heat flux (CHF) and (b) heat transfer tests for pressure reduction transients through the critical pressure. In the present experimental range, the steady-state CHF decreases with the increase of the system pressure For a fixed inlet mass flux and subcooling, the CHF falls sharply at about 3.8 MPa and shows a trend toward converging to zero as the pressure approaches the critical point of 4.059 MPa. The CHF phenomenon near the critical pressure does not lead to an abrupt temperature rise of the heated wall because the CHF occurred at remarkably low power levels. In the pressure reduction transient experiments, as soon as the pressure passed through the critical pressure, the wall temperatures rise rapidly up to a very high value due to the occurrence of the departure from nucleate boiling. The wall temperature reaches a maximum at the saturation point of the outlet temperature, then tends to decrease gradually.

• Nuclear Engineering and Technology
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• v.36 no.5
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• pp.403-414
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• 2004

An experimental study of heat transfer characteristics near the critical pressure has been performed with an internally-heated vertical annular channel cooled by R-134a fluid. Two series of tests have been completed: (a) steady-state critical heat flux (CHF) tests, and (b) heat transfer tests for pressure reduction transients through the critical pressure. In the present experimental range, the steady-state CHF decreases with increase of the system pressure for fixed inlet mass flux and subcooling. The CHF falls sharply at about 3.8 MPa and shows a trend towards converging to zero as the pressure approaches the critical point of 4.059 MPa. The CHF phenomenon near the critical pressure does not lead to an abrupt temperature rise of the heated wall, because the CHF occurs at remarkably low power levels. In the pressure reduction transients, as soon as the pressure passes below the critical pressure from the supercritical pressure, the wall temperatures rise rapidly up to very high values due to the departure from nucleate boiling. The wall temperature reaches a maximum at the saturation point of the outlet temperature, and then tends to decrease gradually.

• Fire Science and Engineering
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• v.24 no.1
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• pp.122-127
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• 2010

The combustion characteristics of phenol foam were analysed using variable external irradiation level (20, 25, 35, 50, and $70kW/m^2$) and in the mixture gas of oxygen/nitrogen. The oxygen index were carried out from the oxygen index tester (KS M ISO 4589-2) and ignition time, critical heat flux, and mass loss rate were carried out from the mass loss calorimeter (ISO 5660-1). As the results of this study, the critical heat flux and average mass loss rate were $28.99kW/m^2$ and $0.56{\sim}1.77g/m^2s$ respectively at the variable external irradiation level. And the limited oxygen index were 45.1% in mixture gas of oxygen/nitrogen. In conclusion, we knew that phenol foam had the best performance than other foam materials in fire safety from all data of this study.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.2396-2401
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• 2007

Freon CHF experiments are carried out to investigate the CHF enhancements by mixing vane shapes of spacer grids in nuclear fuel assembly. The experiments were performed for a wide range mass flux, 50$\sim}$3000 kg/$m^2s$. Three kinds of spacer grids in 5${\times}$5 rod bundles are tested: no mixing vane grids, hybrid mixing vane grids, and split mixing vane grids. The CHF performances are compared along with the data belong to the PWR operating conditions based on a water equivalence through a fluid-to-fluid modeling method. The average of the data in this range is 16.4% for 37 data of hybrid vane grid and 12.5% for 24 data of split vane. In the lower mass flux, however, the split vane grid shows slightly higher performance than the hybrid vane grid.

• Journal of Mechanical Science and Technology
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• v.15 no.7
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• pp.921-930
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• 2001

A new theoretical critical heat flux (CHF) model was developed for the forced convective flow boiling at high pressure, high mass velocity, and low quality. The present model for an intermittent vapor blanket was basically derived from the sublayer dryout theory without including any empirical constant. The vapor blanket velocity was estimated by an axial force balance, and the thickness of vapor blanket was determined by a radial force balance for the Marangoni force and lift force. Based on the comparison of the predicted CHF with the experimental data taken from previous studies, the present CHF model showed satisfactory results with reasonable accuracy.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.528-533
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• 1997

A CANFLEX(CANdu FLEXible fuelling) 43-element bundle has developed for a CANDU-6 reactor as an alternative of 37-element fuel bundle. The design has two diameter elements (11.5 and 13.5㎜) to reduce maximum element power rating and buttons to enhance the critical heat flux(CHF), compared with the standard 37-element bundle. The freon CHF experiments have performed for two series of CANFLEX bundles with and without buttons with a modelling fluid as refrigerant H-l34a and axial uniform heat flux condition. Evaluating the effects of buttons of CANFLEX bundle on CHF and Critical Channel Power(CCP) with the experimental results, it is shown that the buttons enhance CCP as well as CHF. All the CHF's for both the CANFLEX bundles are occurred at the end of fuel channel with the high dryout quality conditions. The CHF enhancement ratio are increased with increase of dryout quality for all flow conditions and also with increase of mass flux only lot high pressure conditions. It indicates that the button is a useful design lot CANDU operating condition because most CHF flow conditions for CANDU fuel bundle are ranged to high dryout quality conditions.