• Nuclear Engineering and Technology
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• 제12권3호
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• pp.153-162
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• 1980

냉각재상실사고의 재관수 단계중 연료봉 피복재의 온도거동 및 열전달 기구를 파악하는 것은 비상노심냉각계통 및 원자로의 안전성해석에 중요하다. 냉각재유동채널의 방위가 rewetting과정에 미치는 영향을 연구하기 위하여 수직 및 수경 유동채널을 이용한 실험을 수행하였으며, 노심이 수평압력관으로 구성되어 있는 CANDU원자로에 관한 실험을 중점적으로 수행하여 그 결과를 수직채널의 결과와 비교 하였다. 또한 rewetting현상을 육안관찰가기 위해 환상형 테스트부 및 외부에서 가열되는 석영관을 사용하였다. 실험결과로써 수평채널에서의 rewetting 속도는 유동의 층상 현상에 크게 영향을 받으나 그 평균값은 수직채널리 경우와 큰차이없음을 알 수 있었다.

• 한국해양공학회지
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• 제25권5호
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• pp.15-20
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• 2011

This study carried out a series of experiments involving impact tests (Drop Weight type & Charpy type with a standard specimen and newly designed I-type specimen), hardness tests, and fracture surface observations of French-made roll shell steel (F), abnormal roll shell steel (M), reheated roll shell steel (R), and S25C steel under heat treatment conditiAn experimental study was carried out to study the solid-liquid mixture upward hydraulic transport of solid particles in vertical and inclined annuli with a rotating inner cylinder. The lift forces acting on a fluidized particle play a central role in many important applications such as the removal of drill cuttings in horizontal drill holes, sand transport in fractured reservoirs, sediment transport, the cleaning of particles from surfaces, etc. In this study a clear acrylic pipe was used to observe the movement of solid particles. Annular velocities varied from 0.4 to 1.2 m/s. The effect of the annulus inclination and drill pipe rotation on the carrying capacity of a drilling fluid, particle rising velocity, and pressure drop in a slim hole annulus were measured for fully-developed flows of water and aqueous solutions of CMC (sodium carboxymethyl cellulose) and bentonite. The rotation of the inner cylinder was efficient at carrying particles to some degree. For a higher particle volume concentration, the hydraulic pressure loss of the mixture flow increased because of the friction between the wall and solids or between solids.

• 태양에너지
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• 제7권2호
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• pp.86-97
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• 1987

Natural convection in the annulus between a horizontal conducting tube and a cylinder with spacers has been studied by 2-dimensional numerical method with finite difference techniques. The effects of Rayleigh number, conductivities of conducting tube and spacer, and position of spacers were studied analytically. In case of vertical spacers, the maximum local Nusselt number appears at ${\theta}{\approx}50^{\circ}$ in a conducting tube and ${\theta}{\approx}30^{\circ}$ in an outer cylinder, The local Nusselt numbers show positive values on the lower spacer, but negative values on the surface of the upper spacer. In case of horizontal spacers, the flow over the spacer is more active than that of under the spacer as the Rayleigh number increases. The maximum local Nusselt appeares at ${\theta}=180^{\circ}$ in a conducting tube and at ${\theta}=0^{\circ}$ in an outer cylinder. The local Nusselt numbers show positive values on the upward surface, but negative values on the downward surface of spacer. As the dimensionless conductivity increases, the mean Nusselt number remarkably increases at $K_w/K_f<48$ and show almost even at $K_w/K_f{\ge}48$. The mean Nusselt number of a conducting tube with vertical spacers is 5.12 percent less and with horizontal spacers is 11.33 percent less than that of a conducting tube without spacer at $Ra=10^4$, Pr = 0.7 and $K_w/K_f=48$.

• 대한기계학회논문집B
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• 제29권2호
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• pp.239-246
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• 2005

Effects of subcooling on pool boiling heat transfer in vertical annuli with closed bottoms have been investigated experimentally. For the test, a tube of 19.1mm diameter and the water at atmospheric pressure have been used. Three annular gaps of 7.05, 18.15, and 28.20 have been tested in the subcooled water and results of the annuli are compared with the data of a single unrestricted tube. The increase in pool subcooling results in much change in heat transfer coefficients. At highly subcooled regions, heat transfer coefficients for the annuli are much larger than those of a single tube. As the heat flux increases and subcooling decrease, a deterioration of heat transfer coefficients is observed at the annulus of 7.05mm gap. Single-phase natural convection and liquid agitation are the governing mechanisms for the single tube while liquid agitation and bubble coalescence are the major factors at the bottom closed annuli.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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• pp.201-207
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• 2001

We modeled the liquid film dryout(LFD) process for both tube and annulus which have uniformly heated vertical channels. We set phenomenological initial conditions in the model. The initial void fraction on the onset of the annular flow location is derived from the physical chum-to-annular flow criterion with the help of the drift-flux-model. The initial thermodynamic-equilibrium-quality is calculated by iteration with the flow quality to find the onset of the annular-flow location. Present model tends to predict very well at the lower exit quality but under-estimates at the higher exit quality. We found that the prediction error of the present model is gradually bigger as the inlet subcooling approaches near the saturation. We obtained excellent results for both tube and annulus channels as the mean of 0.97 and root-mean-square error of 11% for the number of 3883 experimental data on tubes, and of 0.96 and of 12% for 593 on annuli. The present model extended the applicable range to the relatively low exit quality region than previous LFD models.

• Nuclear Engineering and Technology
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• 제31권1호
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• pp.29-48
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• 1999

Local two-phase flow parameters of subcooled flow boiling in inclined annulus were measured to investigate the effect of inclination on the internal flow structure. Two-conductivity probe technique was applied to measure local gas phasic parameters, including void fraction, vapor bubble frequency, chord length, vapor bubble velocity and interfacial area concentration. Local liquid velocity was measured by Pilot tube. Experiments were conducted for three angles of inclination; 0$^{\circ}$(vertical), 30$^{\circ}$, 60$^{\circ}$. The system pressure was maintained at atmospheric pressure. The range of average void fraction was up to 10% and the average liquid superficial velocities were less than 1.3 m/sec. The results of experiments showed that the distributions of two-phase How parameters were influenced by the angle of channel inclination. Especially, the void fraction and chord length distributions were strongly affected by the increase of inclination angle, and flow pattern transition to slug flow was observed depending on the How conditions. The profiles of vapor velocity, liquid velocity and interfacial area concentration were found to be affected by the non-symmetric bubble size distribution in inclined channel. Using the measured distributions of local phasic parameters, an analysis for predicting average void fraction was performed based on the drift flux model and flowing volumetric concentration. And it was demonstrated that the average void fraction can be more appropriately presented in terms of flowing volumetric concentration.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 춘계학술대회
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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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• 제36권5호
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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.

• 대한기계학회논문집B
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• 제26권11호
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• pp.1513-1520
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• 2002

Effect of Convex Surface Curvature on the Onset of Nucleate Boiling of Subcooled Fluid Flow in Vertical Concentric Annuli An experimental study has been carried out to investigate the effect of the transverse convex surface curvature of core tubes on heat transfer in concentric annular tubes. Water is used as the working fluid. Three annuli having a different radius of the inner cores, Ri=3.18mm, 6.35mm, and 12.70mm with a fixed ratio of Ri/Ro=0.5 are used over a range of the Reynolds number between about 40,000 and 80,000. The inner cores are made of smooth stainless steel tubes and heated electrically to provide constant heat fluxes throughout the whole length of each test section. Experimental result shows that heat flux values on the onset of nucleate boiling of the smaller inner diameter model is much higher than that of the larger size test model.

• Nuclear Engineering and Technology
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• 제31권5호
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• pp.486-497
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• 1999

The local heat transfer coefficient is experimentally investigated for the reflux condensation in a countercurrent flow between the steam-air mixture and the condensate, A single vertical tube has a geometry which is a length of 2.4m, inner diameter of 16.56mm and outer diameter of 19.05mm and is made of stainless steel. Air is used as a noncondensible gas. The secondary side has a shape of annulus around vertical tube and the lost heat by primary condensation is transferred to the coolant water. The local temperatures are measured at 11 locations in the vertical direction and each location has 3 measurement points in the radial direction, which are installed at the tube center, at the outer wall and at the coolant side. In three different pressures, the 27 sets of data are obtained in the range of inlet steam flow rate 1.348∼3.282kg/hr, of inlet air mass fraction 11.8∼55.0%. The investigation of the flooding is preceded to find the upper limit of the reflux condensation. Onset of flooding is lower than that of Wallis' correlation. The local heat transfer coefficient increases as the increase of inlet steam flow rate and decreases as the increase of inlet air mass fraction. As an increase of the system pressure, the active condensing region is contracted and the heat transfer capability in this region is magnified. The empirical correlation is developed by 165 data of the local heat transfer. As a result, the Jacob number and film Reynolds number are dominant parameters to govern the local heat transfer coefficient. The rms error is 17.7% between the results by the experiment and by the correlation.