Heat transfer, momentum losses and flow mixing in a 61-tube bundle with wire-wrap

Friction factors and heat transfer coefficients were obtained in the laminar and turbulent regions for a 61-tube wire-wrapped hexagonal bundle in a water flow loop. Circumferential static pressure and temperature profiles of tubes, and the flow patterns produced by injection of dye at the periphery of the bundle revealed a strong local effect of the wire-wrap. The increase in heat and momentum transfer resulting from the wire-wrap was more pronounced in the laminar region than in the turbulent region. Correlations for the friction factors and Nusselt numbers were developed from the data and compared with the literature.     

CFD investigations of helical wire-wrap fuel pin bundle and its comparison with straight wire bundle

Computational Fluid Dynamics (CFD) investigations of a fast reactor fuel pin bundle wrapped with helical and straight spacer wires have been carried out and the advantages of using helical spacer wire have been assessed. The flow and temperature distributions in the fuel pin bundle are obtained by solving the statistically averaged 3-Dimensional conservation equations of mass, momentum and energy along with high Reynolds number k-ε turbulence model using a customized CFD code CFDEXPERT. It is seen that due to the helical wire-wrap spacer, the coolant sodium not only flows in axial direction in the fuel pin bundle but also in a transverse direction. This transverse flow enhances mixing of coolant among the sub channels and due to this, the friction factor and heat transfer coefficient of the coolant increase. Estimation of friction factor, Nusselt number, sodium temperature uniformity at the outlet of the bundle and clad hot spot factor which are measures of the extent of coolant mixing and non-homogeneity in heat transfer coefficient around fuel pin are paid critical attention. It is seen that the friction factor and Nusselt number are higher (by 25% and 15% respectively) for the helical wire wrap pin bundle compared to straight wire bundle. It is seen that for 217 fuel pin bundle the maximum clad temperature is 750 K for straight wire case and the same for helical wire is 720 K due to the presence of transverse flow. The maximum temperature occurs at the location of the gap between pin and wire. The ΔT between the bulk sodium in the central sub-channel and peripheral sub-channel is 30 K for straight wire and the same for helical wire is 18 K due to the presence of secondary transverse flow which makes the outlet temperature more uniform. The hotspot factor and the hot channel factors predicted by CFD simulation are 10% lower than that used in conventional safety analysis indicating the conservatism in the safety analysis.     

Turbulent flow simulation in a wire-wrap rod bundle of an LMFBR

The pressure drop and heat transfer characteristics of wire-wrapped 19-pin rod bundles in a nuclear reactor subassembly of liquid metal cooled fast breeder reactor (LMFBR) have been investigated through three-dimensional turbulent flow simulations. The predicted results of eddy viscosity based turbulence models (k-?, k-ω) and the Reynolds stress model are compared with those of experimental correlations for friction factor and Nusselt number. The Re is varied between 50,000 and 150,000 and the ratio of helical pitch of wire wrap to the rod diameter is varied from 15 to 45. All the three turbulence models considered yield similar results. The friction factor increases with reduction in the wire-wrap pitch while the heat transfer coefficient remains almost unaltered. However, reduction in the wire-wrap pitch also enhances the transverse flow velocity in the cross-sectional plane as well as the local turbulence intensity, thereby improving the thermal mixing of coolant. Consequently, the presence of wire wrap reduces temperature variation within each section of the subassembly. The associated reduction in differential thermal expansion of rods is expected to improve the structural integrity of the fuel subassembly.     

Axial static pressure variations in inner and side subchannels of a 61-tube wire-wrap bundle

Measurements of axial distribution of the static pressure in an inner and side subchannel of a 61 wire-wrap tube bundle obtained with water at atmospheric conditions are presented. The wire wrap configuration is different from those used by previous workers and more representative of a bundle for the blanket of a Gas Cooled Fast Reactor. The data display axial static pressure variations which are attributed to the interchannel cross flow induced by the wire-wrap configuration. The static pressure drop over one wire pitch agrees well with the bundle pressure drop based on a bundle average Reynolds number and a friction factor f = 0.436 Re^−0.263 (Re > 2000). The experimental data obtained with water provide a useful benchmark to model and check the accuracy of thermal-hydraulic codes used for the analysis of subchannel flow distribution and pressure drop in wire wrap tube bundle cooled with one-phase fluid.The nodal subchannel code COBRA-IV was modeled by adjusting the forced cross-flow function to match the measured axial static pressure distribution in an inner and side subchannel. Some discrepancy remained in the static pressure profile in the side channel attributed to the flow distortion at the bundle exit.     

A review of heat exchanger tube bundle vibrations in two-phase cross-flow

Flow-induced vibration is an important concern to the designers of heat exchangers subjected to high flows of gases or liquids. Two-phase cross-flow occurs in industrial heat exchangers, such as nuclear steam generators, condensers, and boilers, etc. Under certain flow regimes and fluid velocities, the fluid forces result in tube vibration and damage due to fretting and fatigue. Prediction of these forces requires an understanding of the flow regimes found in heat exchanger tube bundles. Excessive vibrations under normal operating conditions can lead to tube failure.

Relatively little information exists on two-phase vibration. This is not surprising as single-phase flow induced vibration; a simpler topic is not yet fully understood. Vibration in two-phase is much more complex because it depends upon two-phase flow regime, i.e. characteristics of two-phase mixture and involves an important consideration, which is the void fraction. The effect of characteristics of two-phase mixture on flow-induced vibration is still largely unknown. Two-phase flow experiments are much more expensive and difficult to carry out as they usually require pressurized loops with the ability to produce two-phase mixtures. Although convenient from an experimental point of view, air–water mixture if used as a simulation fluid, is quite different from high-pressure steam–water. A reasonable compromise between experimental convenience and simulation of steam–water two-phase flow is desired.

This paper reviews known models and experimental research on two-phase cross-flow induced vibration in tube bundles. Despite the considerable differences in the models, there is some agreement in the general conclusions. The effect of tube bundle geometry, random turbulence excitations, hydrodynamic mass and damping ratio on tube response has also been reviewed. Fluid–structure interaction, void fraction modeling/measurements and finally Tubular Exchanger Manufacturers Association (TEMA) considerations have also been highlighted.     

Supercritical-water heat transfer in a vertical bare tube

This paper presents selected results on heat transfer to supercritical water flowing upward in a 4-m-long vertical bare tube. Supercritical-water heat-transfer data were obtained at pressures of about 24 MPa, mass fluxes of 200-1500 kg/m² s, heat fluxes up to 884 kW/m² and inlet temperatures from 320 to 350 °C for several combinations of wall and bulk-fluid temperatures that were below, at or above the pseudocritical temperature.In general, the experiments confirmed that there are three heat-transfer regimes for forced-convective heat transfer to water flowing inside tubes at supercritical pressures: (1) normal heat-transfer regime characterized in general with heat transfer coefficients (HTCs) similar to those of subcritical convective heat transfer far from critical or pseudocritical regions, which are calculated according to the Dittus-Boelter type correlations; (2) deteriorated heat-transfer regime with lower values of the HTC and hence higher values of wall temperature within some part of a test section compared to those of the normal heat-transfer regime; and (3) improved heat-transfer regime with higher values of HTC and hence lower values of wall temperature within some part of a test section compared to those of the normal heat-transfer regime.This new heat-transfer dataset is applicable as a reference dataset for future comparison with supercritical-water bundle data and for a verification of scaling parameters between water and modeling fluids.Also, these HTC data were compared to those calculated with the original Dittus-Boelter and Bishop et al. correlations. The comparison showed that the Bishop et al. correlation, which uses the cross-section average Prandtl number, represents HTC profiles more correctly along the heated length of the tube than the Dittus-Boelter correlation. In general, the Bishop et al. correlation shows a fair agreement with the experimental HTCs outside the pseudocritical region, however, overpredicts by about 25% the experimental HTCs within the pseudocritical region. The Dittus-Boelter correlation can also predict the experimental HTCs outside the pseudocritical region, but deviates significantly from the experimental data within the pseudocritical region. It should be noted that both these correlations cannot be used for a prediction of HTCs within the deteriorated heat-transfer regime.     

The hexagonal bundle heat transfer and fluid flow experiment agathe hex

Problems of heat transfer and fluid flow in gas-cooled reactor fuel elements have been studied at the Swiss Federal Institute for Reactor Research (EIR) for 14 years. Since 1967, the activities have been directed toward gas-cooled fast breeder reactors (GCFRs). The aim of analytical and experimental studies has been to develop analytical models and comprehensive computer codes for the prediction of temperature and pressure distributions in GCFR fuel element configurations. The models developed at EIR are based on the results of specific experiments. Full-scale experiments in actual geometry are being carried out to verify the computer codes for a wide range of parameters. This paper describes the heat transfer loop and the test sections designed to verify GCFR thermohydraulic design codes.     

Single-phase convective heat transfer enhancement by spacer grids in a rod bundle

The spacer grids within a fuel assembly of a nuclear reactor core disrupt and re-establish the momentum and thermal boundary layers so that they enhance the local heat transfer within and downstream of the spacer grids. An experimental study in a 6×6 rod bundle has been performed to investigate the effects of spacer grids on the single-phase convective heat transfer enhancement. The experimental data showed that the Reynolds number has a significant impact on the heat transfer enhancement only when the Reynolds numbers are lower than about 10,000. The conventional correlations showed poor predictions of the heat transfer enhancement by spacer grids at low Reynolds numbers; in particular, the maximum heat transfer rate at the top end of the spacer grids was significantly overestimated. Furthermore, the conventional correlations did not properly account for the effects of the Reynolds numbers on the heat transfer enhancement. Therefore, more systematic experiments should be performed using various spacer grids with large blockage ratios at low Reynolds numbers, considering an early phase of the reflood conditions.     

Pool boiling heat transfer enhancement on porous surface tube

The passive residual heat removal exchanger (PRHR HX),which is a key equipment of the passive residual heat removal system,is installed in an elevated pool.Its heat transfer performance affects security and economics of the reactor,and boiling heat transfer in the liquid surrounding the exchanger occurs when the liquid saturation temperature exceeded.The smooth tubes,which are widely used as heat transfer tubes in PRHR HX,can be replaced by some enhanced tubes to improve the boiling heat transfer capability.In this paper,the pool boiling heat transfer characteristics of smooth tube and a machined porous surface tube are investigated by using high-pressure steam condensing inside tube as heating source.Compared with smooth tube,the porous surface tube considerably enhances the boiling heat transfer,and shortens the time significantly before reaching the liquid saturation temperature.Its boiling heat transfer coefficient increases from 68% to 75%,and the wall superheat decreases by 1.5oC.Combining effect of condensation inside tube with boiling outside tube,the axial wall temperatures of heat transfer tube are neither uniform nor linear distribution.Based on these investigations,enhance mechanism of the porous surface tube is analyzed.     

Effects of tube inclination on pool boiling heat transfer

Effects of tube inclination angles on nucleate pool boiling heat transfer of water at atmospheric pressure have been investigated experimentally. Experiments were performed for seven angles (0, 15, 30, 45, 60, 75, and 90°) with two tubes (12.7 and 19.1 mm in diameter) of 540 mm in length. Through the study, it can be concluded that tube inclination gives much change on nucleate pool boiling heat transfer. When the tube is near the horizontal and the vertical positions, the maximum and minimum of heat transfer coefficients are expected, respectively. The decrease in bubble slug formation on the tube surface and easy liquid access to the surface are thought to be the causes for the enhanced heat transfer.     

Simulation of heat transfer of supercritical water in obstacle-bearing vertical tube

The heat transfer coefficient is very low at bulk temperatures higher than the pseudo-critical point,because the supercritical pressure leads to a vapor-like fluid.In this paper,the heat transfer downstream an obstacle-bearing vertical tube is simulated by the CFD code of Fluent 6.1,using an adaptive grid in the supercritical condition.The reliable results are obtained by the RNG k-ε model using the enhanced wall treatment.The blockage ratio and local temperature of obstacle affect greatly the heat transfer enhancement,and the resultant influence region and decay trend are compared with the existing equations.     

Investigation of pre- and post-dryout heat transfer of steam-water two-phase flow in a rod bundle

Pre- and post-dryout heat transfer experiments were performed for steam-water two-phase flow in a 5 × 5 rod bundle under conditions of total mass fluxes from 80 to 320 kg/m²s, inlet qualities from 0.1 to 0.8, heat fluxes from 3 to 26 W/cm² and a pressure of 3 MPa. Heater rod surface temperatures or heat transfer coefficients predicted by several correlations were compared with experimental data with emphasis on the applicability of the correlations to the present experimental conditions which were pertinent to thermal-hydraulic conditions during a LOCA in a nuclear reactor. The Chen and Biorge et al. correlations underestimated heat transfer coefficients in the pre-dryout region. The Varone-Rohsenow prediction which accounted for the thermal nonequilibrium effect, calculated heater rod surface temperatures relatively well in the post-dryout region over the whole region of the present experimental conditions. The Dittus-Boelter and Groeneveld correlations predicted heater rod surface temperatures relatively well in the post-dryout region under high total mass flux conditions, but underestimated considerably under low total mass flux conditions.     

Local configuration factors for radiant interchange between cylindrical surfaces in rod bundle geometry

The aim of this study is to obtain configuration factor expressions for the diffuse interchange of radiant energy between two parallel cylindrical rods by contour integral technique. The contour integral approach for diffuse configuration factors has been derived by Sparrow (1963). It replaces the usual area integrals by more tractable contour (i.e., line) integrals, which simplifies analytical calculation of configuration factors. The advantages of this representation are associated with the reduced order of the integrals (i.e., double reduced to single, quadruple reduced to double) which must be evaluated to calculate the angle factor. For the numerical or analytical evaluation of configuration factors, the reduction in the order of the integrals has a great practical utility. In the literature, contour integral technique has been stated to appreciably be the more precise method than the area integration method.The cases studied are for elemental ring areas of (i) two cylindrical rods/tubes, (ii) two cylindrical rods with interference from a third rod and (iii) cylindrical rod within a cylindrical enclosure. The results of present configuration factor expressions meant for the ring elements of finite length geometries have been compared with values using exact algebraic expressions available in the literature for infinite length, analytical configuration factor expressions for equal finite lengths and with numerical results available in the literature. Contour integral technique has been observed to be a simple method that yields realistic results. This study has application in the analysis of fin-tube radiators, boiler tubes, nuclear fuel rod bundle, etc.     

Simple heat transfer model for laminar film condensation in a vertical tube

A new physical model for estimating the liquid film thickness and condensation heat transfer coefficient in a vertical tube, considering the effects of gravity, liquid viscosity, and vapor flow in the core region, is proposed. In particular, for calculating the velocity profile in the liquid film, the liquid is assumed to be in Couette flow forced by the interfacial velocity at the liquid-vapor interface. The interfacial velocity is calculated using an empirical power-law velocity profile. The film thickness and heat transfer coefficient from the new model are compared with existing experimental data and the original Nusslet condensation theory. The new model describes the liquid film thinning effect due to the vapor shear flow and predicts the condensation heat transfer coefficient from experiments reasonably well.     

Numerical simulation of heat transfer deterioration phenomenon in supercritical water through vertical tube

In this study, a numerical investigation of heat transfer deterioration (HTD) in supercritical water flowing through vertical tube is performed by using six low-Reynolds number turbulence models. All low-Reynolds models can be extended to reproduce the effect of buoyancy force on heat transfer and show the occurrence of localized HTD. However, most k–ε models seriously over-predict the deterioration and do not reproduce the subsequent recovery of heat transfer. The V2F and SST models perform better than other models in predicting the onset of deterioration due to strong buoyancy force. The SST model is able to quantitatively reproduce the two heat transfer deterioration phenomena with low mass flux which have been found in the present study.     

Numerical study on flow and heat transfer characteristics in the rod bundle channels under super critical pressure condition

In this study, the 3D flow and heat transfer characteristics in rod bundle channels of the super critical water-cooled reactor were numerically investigated using CFX codes. Different turbulent models were evaluated and the flow and heat transfer characteristics in different typical channels were obtained. The effect of pitch-to-diameter ratio (P/D) on the distributions of surface temperature and heat transfer coefficient (HTC) was analysed. For typical quadrilateral channel, it was found that HTC increases with P/D first and then decreases significantly when P/D is <1.4. There exists a “flat region” at the maximum value when P/D is 1.4. If P/D is larger than 1.4, heat transfer deterioration (HTD) occurs as main stream enthalpy is quite small. Furthermore, the HTD under low mass flow rate and the non-uniformity of circumferential temperature were also discussed.     

An investigation on heat transfer characteristics of different pressure steam-water in vertical upward tube

Within the range of pressure from 9 to 30 MPa, mass velocity from 600 to 1200 kg/(m² s), and heat flux at inner wall from 200 to 600 kW/m², experiments have been performed to investigate the heat transfer characteristics of steam-water two-phase flow in vertical upward tube. The outer diameter of the tube is 32 mm, and the wall thickness is 3 mm. Based on results, it was found that Dryout is the main mechanism of the heat transfer deterioration in the sub-critical pressure region. Near the critical pressure, when the heat transfer deterioration occurs, the steam quality of water is lower than that in the sub-critical pressure region, so that DNB is the main mechanism in this pressure region. At supercritical pressure, the heat transfer performance in circular channel is improved and enhanced. Heat transfer deterioration phenomenon is observed when the fluid bulk temperature approaches to the pseudo-critical value. Nusselt correlation of the forced-convection heat transfer in supercritical pressure region has been provided, which can be used to predict heat transfer coefficient of the vertical upward flow in tube.