• Nuclear Engineering and Technology
• /
• 제55권10호
• /
• pp.3754-3767
• /
• 2023

Turbulent jet discharges in subcooled water pools are essential for safety systems in nuclear power plants, specifically in the pressure suppression pool of boiling water reactors and In-containment Refueling Water Storage Tank of advanced pressurized water reactors. The gas and liquid flow in these systems is investigated using multiphase flow analysis. This field has been extensively examined using a combination of experiments, theoretical models, and Computational Fluid Dynamics (CFD) simulations. ANSYS CFX offers two approaches to model multiphase flow behavior. The non-homogeneous Eulerian-Eulerian Model has been used in this work; it computes global information and is more convenient to study interpenetrated fluids. This study utilized the Large Eddy Simulation Model as the turbulence model, as it is better suited for non-stationary and buoyant flows. The CFD results of this study were validated with experimental data and theoretical results previously obtained. The figures of merit dimensionless penetration length and the dimensionless buoyancy length show good agreement with the experimental measurements. Correlations for these variables were obtained as a function of dimensionless numbers to give generality using only initial boundary conditions. CFD numerical model developed in this research has the capability to simulate the behavior of non-condensable gases discharged in water.

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

• 한국가스학회지
• /
• 제21권4호
• /
• pp.92-102
• /
• 2017

피셔-트롭쉬 합성반응은 CO와 H2의 혼합가스로 이루어진 합성가스를 부가가치가 높은 탄화수소 제품으로 변환시킨다. 본 논문에서는 저온 피셔-트롭쉬 합성반응과 단일, 다중 마이크로채널 반응기에 패킹시킨 촉매를 기반으로 강화된 반응조건의 열전달을 고려하여 전산유체역학 기반의 시뮬레이션을 진행하고 분석하였다. 단일채널모델을 통하여 CO 전환률이 ~65% 이상, $C_{5+}$ 선택도가 ~74% 이상을 달성하면서도 Co 기반의 super-active 촉매를 통해 GHSV를 $30000hr^{-1}$을 달성할 수 있음을 보였다. 다중 마이크로채널 반응기모델에서는 열전달 시뮬레이션을 동시에 해석하여, 3가지의 다른 반응기구조에 대해서, 직교류 wall boiling 냉매를 사용시 ${\Delta}T_{max}$가 23 K였으며 평행유동 subcooled 냉매와 평행유동 wall boiling 냉매의 경우 각각 15 K와 13 K의 ${\Delta}T_{max}$를 보였다. 반응기 전체적으로 498 - 521 K에서 온도제어가 가능했으며 계산된 사슬성장 가능성은 저온 피셔-트롭쉬 합성에 적합한 것으로 보인다.

• 한국안전학회지
• /
• 제32권2호
• /
• pp.147-152
• /
• 2017

The external reactor vessel cooling (ERVC) is well known strategy to mitigate a severe accident at which nuclear fuel inside the reactor vessel is molten. In order to compare the heat removal capacity of ERVC between the nuclear reactor designs quantitatively, numerical method is often used. However, the study for ERVC using computational fluid dynamics (CFD) is still quite scarce. As a validation study on the numerical prediction for ERVC using CFD, the subcooled boiling flow and natural circulation of coolant at the ULPU-V experiment was simulated. The commercially available CFD software ANSYS-CFX was used. Shear stress transport (SST) model and RPI model were used for turbulence closure and wall-boiling, respectively. The averaged flow velocities in the downcomer and the baffle entry under the reactor vessel lower plenum are in good agreement with the available experimental data and recent computational results. Steam generated from the heated wall condenses rapidly and coolant flows maintains single-phase flow until coolant boils again by flashing process due to the decrease of saturation temperature induced by higher elevation. Hence, the flow rate of coolant natural circulation does not vary significantly with the change of heat flux applied at the reactor vessel, which is also consistent with the previous literatures.

• Journal of Advanced Marine Engineering and Technology
• /
• 제10권2호
• /
• pp.156-164
• /
• 1986

In the case of boiling on high temperature wall, vapor film covers fully or parcially the surface. This phenomenon, film boiling or transition boiling, is very important in the surface heat treatment of metal, design of cryogenic heat exchanger and emergency cooling of nuclear reactor. Mainly supposed hydraulic-thermal accidents in nuclear reactor are LCCA (Loss of Coolant Accident) and PCM (Power-Cooling Mismatch). Recently, world-wide studies on reflooding of high temperature rod bundles after the occurrence of the above accidents focus attention on wall temperature history and required time in transient cooling process, wall superheat at rewet point, heat flux-wall superheat relationship beyond the transition boiling region, and two-phase flow state near the surface. It is considered that the further systematical study in this field will be in need in spite of the previous results in ref. (2), (3), (4). The paper is the study about the fast transient cooling process following the wall temperature excursion under the CHF (Critical Heat Flux) condition in a forced convective subcooled boiling system. The test section is a vertically arranged concentric annulus of 800 mm long and 10 mm hydraulic diameter. The inner tube, SUS 304 of 400 mm long, 8 mm I.D, and 7 mm O.D., is heated uniformly by the low voltage AC power. The wall temperature measurements were performed at the axial distance from the inlet of the heating tube, z=390 mm. 6 chromel- alumel thermocouples of 76 .mu.m were press fitted to the inner surface of the heating tube periphery. To investigate the heat transfer characteristics during the fast transient cooling process, the outer surface (fluid side) temperature and the surface heat flux are computed from the measured inner surface temperature history by means of a numerical method for inverse problems of transient heat conduction. Present cooling (boiling) curve is sufficiently compared with the previous results.

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2001년도 춘계학술대회논문집D
• /
• pp.208-213
• /
• 2001

The minimum heat flux conditions are experimentally investigated for the subcooled liquid film flow on the horizontal plate. The experimental results show that the minimum heat flux point temperature becomes higher with the increase of the velocity and the subcooling of the liquid film flow. However, the effect of distance from the leading edge of the heat transfer plate on the minimum heat flux is almost negligible. Also, the experimental results show that the propagation velocity of wetting front increase with increasing the velocity and the subcooling of the liquid film flow.

• Journal of Advanced Marine Engineering and Technology
• /
• 제11권1호
• /
• pp.72-79
• /
• 1987

When the wall temperature is very high, a stable vapor film covers the heat transfer surface. The vapor film creates a strong thermal resistance when heat is transferred to the liquid though it. This phenomenon, called "film boiling" is very important in the heat treatment of metals, the design of cryogenic heat exchangers, and the emergency cooling of nuclear reactors. In the practical engineering problems of the transient cooling process of a high temperature wall, the wall temperature history, the variation of the heat transfer coefficients, and the wall superheat at the rewetting points, are the main areas of concern. These three areas are influenced in a complex fashion such factors as the initial wall temperature, the physical properties of both the wall and the coolant, the fluid temperature, and the flow state. Therefore many kinds of specialized experiments are necessary in the creation of precise thermal design. The object of this study is to investigate the heat transfer characteristics in the transient cooling process of a high temperature wall. The slow transient cooling experiment was carried out with a copper block of high thermal capacity. The block was 240 mm high and 79 mm O.D.. The coolant flowed throuogh the center of a 10 mm diameter channel in the copper block. In the copper block, three sheathed thermocouples were placed in a line perpendicular to the flow. These thermocouples were used to take measurements of the temperature histories of the copper block.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
• /
• pp.241-246
• /
• 1996

An analytical model to calculate rate of vapor generation due to heterogeneous wall nucleation in flashing flow is developed. In the present model, an important parameter of the vapor generation term, i.e. nucleation site density is calculated by integrating its probability distribution function with respect to active cavity radius. The limits of integration are minimum and maximum active cavity radii, and these are formulated using an active cavity model for nucleate boiling. This formulation, therefore. can statistically account for the effect of surface specific thermo-physical and geometric conditions on the vapor generation rate and flashing inception. For verifying the adequacy of the present model, steady state two-fluid and the bubble transport equations are solved with applicable constitutive equations. The applicable region of the bubble transport equation is also extended to churn-turbulent flow regime to predict interfacial area concentration at high void fraction. Predicted results in terms of axial pressure and void fraction profiles along the channels are compared with experimental data of Super Moby Dick and BNL Reasonable agreements have been achieved and this shows the applicability of the present model to flashing flow analysis.

• 설비공학논문집
• /
• 제13권10호
• /
• pp.960-965
• /
• 2001

The minimum heat flux conditions are experimentally investigated for the subcooled liquid film flow on the horizontal plate. The experimental results show that the minimum heat flux point temperature becomes higher with the increase of the velocity and the subcooling of the liquid film flow. However, the effect of distance from the leading edge of the heat transfer plate on the minimum heat flux is almost negligible. Also, the experimental results show that the propagation velocity of wetting front increases with increasing the velocity and the subcooling of the liquid film flow.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
• /
• pp.108-113
• /
• 1996

Several analytical models for the departure from nucleate boiling (DNB) phenomenon have been developed during the last decade. Among these, Chang & Lee's model based on a bubble crowding mechanism is remarkable in the fundamental features characterized as the formulation of mass, energy, and momentum balance equation at thermal-hydraulic conditions leading to the DNB. However, Bricard and Souyri remarked that the assumption of stagnant bubbly layer at the DNB condition is questionable and the signs on the axial projections of the momentum fluxes at the core/bubbly layer interface in the momentum balance equations are erroneous. From this remark, Chang & Lee's model has been re-examined and modified by correcting the erroneous treatments in the momentum balance equations and removing the spurious assumptions. The revised model predicts well the extensive DNB data of water in uniformly heated tubes at low qualities and shows more accurate prediction compared with the original model.