• Nuclear Engineering and Technology
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• v.34 no.4
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• pp.269-285
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• 2002

Water heat transfer experiments were carried out in a uniformly heated annulus with a wide range of pressure conditions. The local heat transfer coefficients for saturated water (low boiling have been measured just before the occurrence of the critical heat flux (CHF) along the length of the heated section. The trends of the measured heat transfer coefficients were quite different from the conventional understanding for the heat transfer of saturated flow boiling. This discrepancy was explained from the nucleate boiling in the liquid film of annular flow under high heat flux conditions. The well-known correlations were compared with the measured heat transfer coefficients. The Shah and Kandlikar correlations gave better prediction than the Chen correlation. However, the modified Chen correlation proposed in the present work showed the best agreement with the present data among correlations examined .

• Transactions of the Korean Society of Mechanical Engineers B
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• v.36 no.8
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• pp.775-782
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• 2012

To investigate pool boiling heat transfer in a vertical annulus with closed bottoms, the length of an outer tube was varied between 0.3 and 0.6 m. For the test, a heated tube of 0.2-m length and 19.1-mm diameter and water at atmospheric pressure were used. To elucidate the effects of the outer tube length on heat transfer, the results for the annulus were compared with data for a single unrestricted tube. The increase in the outer tube length resulted in an increase or decrease in heat transfer depending on the gap size. This tendency is mainly attributed to the difference in the intensity of liquid agitation.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.325-330
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• 1996

The existing CHF correlations for internally heated concentric annulus channels are assessed using KAIST CHF database for uniformly heated vertical annuli. Six annulus correlations (Jannsen-Kervinen. Barnett, Levitan-Lantsman, Kumamaru et al., Doerffer et al., and Bobkov et at.) are chosen for assessment based on literature survey and Groeneveld et al.'s CHF table for round tube is also assessed for comparison. Among the above correlations, two are inlet-condition type and others local conditions type. To make the comparison meaningful, the local-condition-type correlations are assessed in two ways: direct substitution method (DSM) and heat balance condition method (HBM). Totally 1174 data are classified into 10 groups based on pressure and mass flux conditions and correlations are assessed to each group separately. Prediction capability of each correlation depends on the data group and none shows the best prediction over the entire group. In overall, the correlations by Doerffer et al. and Jannsen et al. appear to be the best, but Barnett or Levitan-Lantsman correlations also show reasonable prediction for most groups. However, the low-pressure, ]ow flow CHFs are not well predicted by any correlations. The CHF table for round tubes overpredicts the CHF in annuli at fixed local conditions.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.37 no.4
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• pp.377-385
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• 2013

To identify the effects of an outflow area on pool boiling heat transfer in a vertical annulus, three different flow restrictors were studied experimentally. For the test, a heated tube of smooth stainless steel and water at atmospheric pressure were used. Both annuli with open and closed bottoms were considered. To validate the effects of the outflow area on the heat transfer, the results of the annulus with the restrictor were compared with the data for the plain annulus without the restrictor. The reduction of the outflow area ultimately results in a decrease in the heat transfer. As the outflow area is very small, a slight increase in heat transfer is also observed. The major cause of this tendency is explained as the difference in the intensity of liquid agitation cause by the movement of coalesced bubbles. It is identified that the convective flow, pulsating flow, and evaporative mechanism are considered as the important mechanisms.

• 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.

• Journal of computational fluids engineering
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• v.4 no.2
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• pp.1-8
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• 1999

Natural convection in a horizontal annulus with the inner cylinder heated by the application of a constant heat flux and the isothermally cooled outer cylinder is considered, and the transition of flows and the bifurcation phenomenon are numerically investigated for air with Pr=0.7. The zero initial condition always induces a crescent-sheped eddy flow. A bicellular flow in which the fluid descends along the vertical central plane of the annulus can be obtained at high Rayleigh number by introducing artificial numerical disturbances. Dual solutions are found above a certain critical Rayleigh number. Hysteresis phenomena have not been observed.

• Nuclear Engineering and Technology
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• v.34 no.4
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• pp.382-395
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• 2002

An experimental study on transient critical heat flux (CHF) under flow coastdown has been performed for the water flow in a non-uniformly heated vertical annulus under low flow and a wide range of pressure conditions. The objectives of this study are to systematically investigate the effect of the flow transient on the CHF and to compare the transient CHF with steady-state CHF The transient CHF experiments have been performed for three kinds of flow transient modes based on the coastdown data of a nuclear power plant reactor coolant pump. At the same inlet subcooling, system pressure and heat flux, the effect of the initial mass flux on the critical mass flux can be negligible. However, the effect of the initial mass flux on the time-to- CHF becomes large as the heat flux decreases. The critical mass flux has the largest value for slow flow reduction rate. There is a pressure effect on the ratio of the transient CHF data to steady-state CHF data. Except under low system pressure conditions, the flow transient CHF was revealed to be conservative compared with the steady-state CHF data. Bowling CHF correlation and thermal hydraulic system code MARS show promising results for the prediction of CHF occurrence .

• 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.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.35 no.3
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• pp.237-244
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• 2011

An experimental study was carried out to investigate the effect of the inclination angle on the nucleate pool boiling of saturated water at atmospheric pressure. We considered an annulus with a gap of 5 mm and a bottom opening. The inner tube of the annulus was heated, and the outer diameter and the length of the tube were 25.4 mm and 500 mm, respectively. The inclination angle was varied from horizontal to vertical. The results were compared to those for an annulus with a larger gap and a single tube. In the small-gap annulus, the effect of the inclination angle on the heat transfer was not significant. However, an early onset of the critical heat flux was observed at 80 kW/$m^2$ when the annulus was horizontal. Liquid agitation and bubble coalescence were considered to be the major heat-transfer mechanisms.

• 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.