Experimental investigation of heat transfer for supercritical pressure water flowing in vertical annular channels

An experiment has recently been completed at Xi’an Jiaotong University (XJTU) to obtain wall-temperature measurements at supercritical pressures with upward flow of water inside vertical annuli. Two annular test sections were constructed with annular gaps of 4 and 6 mm, respectively, and an internal heater of 8 mm outer diameter. Experimental-parameter ranges covered pressures of 23-28 MPa, mass fluxes of 350-1000 kg/m²/s, heat fluxes of 200-1000 kW/m², and bulk inlet temperatures up to 400 °C. Depending on the flow conditions and heat fluxes, two distinctive heat transfer regimes, referring to as the normal heat transfer and deteriorated heat transfer, have been observed. At similar flow conditions, the heat transfer coefficients for the 6 mm gap annular channel are larger than those for the 4 mm gap annular channel. A strong effect of spiral spacer on heat transfer has been observed with a drastic reduction in wall temperature at locations downstream of the device in the annuli. Two tube-data-based correlations have been assessed against the experimental heat transfer results. The Jackson correlation agrees with the experimental trends and overpredicts slightly the heat transfer coefficients. The Dittus-Boelter correlation is applicable only for the normal heat transfer region but not for the deteriorated heat transfer region.     

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.     

Investigation of forced convection heat transfer of supercritical pressure water in a vertically upward internally ribbed tube

In the present paper, the forced convection heat transfer characteristics of water in a vertically upward internally ribbed tube at supercritical pressures were investigated experimentally. The six-head internally ribbed tube is made of SA-213T12 steel with an outer diameter of 31.8 mm and a wall thickness of 6 mm and the mean inside diameter of the tube is measured to be 17.6 mm. The experimental parameters were as follows. The pressure at the inlet of the test section varied from 25.0 to 29.0 MPa, and the mass flux was from 800 to 1200 kg/(m² s), and the inside wall heat flux ranged from 260 to 660 kW/m². According to experimental data, the effects of heat flux and pressure on heat transfer of supercritical pressure water in the vertically upward internally ribbed tube were analyzed, and the characteristics and mechanisms of heat transfer enhancement, and also that of heat transfer deterioration, were also discussed in the so-called large specific heat region. The drastic changes in thermophysical properties near the pseudocritical points, especially the sudden rise in the specific heat of water at supercritical pressures, may result in the occurrence of the heat transfer enhancement, while the covering of the heat transfer surface by fluids lighter and hotter than the bulk fluid makes the heat transfer deteriorated eventually and explains how this lighter fluid layer forms. It was found that the heat transfer characteristics of water at supercritical pressures were greatly different from the single-phase convection heat transfer at subcritical pressures. There are three heat transfer modes of water at supercritical pressures: (1) normal heat transfer, (2) deteriorated heat transfer with low HTC but high wall temperatures in comparison to the normal heat transfer, and (3) enhanced heat transfer with high HTC and low wall temperatures in comparison to the normal heat transfer. It was also found that the heat transfer deterioration at supercritical pressures was similar to the DNB at subcritical pressures.     

Investigations on post-dryout heat transfer in bilaterally heated annular channels

Post-dryout heat transfer in bilaterally heated vertical narrow annular channels with 1.0, 1.5 and 2.0 mm gap size has been experimentally investigated with deionized water under the condition of pressure ranging from 1.38 to 5.9 MPa and low mass flow rate from 42.9 to 150.2 kg/m²s. The experimental data was compared with well known empirical correlations including Groeneveld, Mattson, etc., and none of them gave an ideal prediction. Theoretical investigations were also carried out on post-dryout heat transfer in annular channels. Based on analysis of heat exchange processes arising among the droplets, the vapor and two tube walls of annular channel, a non-equilibrium mechanistic heat transfer model was developed. Comparison indicated that the present model prediction showed a good agreement with our experimental data. Theoretical calculation result showed that the forced convective heat transfer between the heated wall and vapor dominate the overall heat transfer. The heat transfer caused by the droplets direct contact to the wall and the interfacial convection/evaporation of droplets in superheated vapors also had an indispensable contribution. The radiation heat transfer would be neglected because of its small contribution (less than 0.11%) to the total heat transfer.     

两流程2×2棒束超临界水传热实验研究

为支持欧盟2×2燃料组件入堆考核实验,开展了2×2棒束内超临界水传热实验研究。设计了两流程实验段,两流程之间通过矩形金属框隔离。2×2加热棒束通过定位格架安装在金属框中,超临界水通过第1流程向下流动然后进入第2流程冷却加热棒,第1流程通过金属框吸收第2流程流体的热量而温度升高。在2×2棒束横截面上存在明显的周向温度分布不均匀性。系统分析了实验参数包括质量流速、热流密度和实验压力对棒束平均换热系数的影响。结果表明,系统参数对棒束传热的影响规律与对单管和环管的一致。     

Analysis, verification, and benchmarking of the transient thermal hydraulic ITERTHA code for the design of ITER divertor

An analytical study for the International Thermonuclear Experimental Reactor Thermal Hydraulic Analysis code (ITERTHA) is carried out for a copper divertor with a 5 mm tungsten tile. The influence of the incident heat flux, swirl-tape insertion in cooling channels as well as the coolant flow velocity on the divertor thermal response is analyzed and discussed. The ITERTHA code results are verified by the commercial finite element code, COSMOS. The heat transfer coefficients at the nodes located on the cooling channel-wall are determined outside COSMOS code by the same methodology used in ITERTHA. A good agreement is achieved under different incident heat fluxes. The ITERTHA code is also benchmarked against the thermal-hydraulic calculation of the outer divertor of the Fusion Ignition Research Experiment, FIRE for an incident heat flux of 20 MW/m² and coolant flow velocity of 10 m/s in a cooling channel of 8 mm diameter with swirl-tape inserts of 2 ratio and 1.5 mm thickness. The results show excellent agreement for both steady and transient states and prove the successful implementation of both the hydraulic and heated diameters of the swirl-tape channels in the used heat transfer correlations.     

当量直径对超临界水流动传热特性的影响

以CFD商业软件FLUENT为计算平台,对圆管和圆环通道内超临界水流动传热特性进行数值模拟。通过对几种湍流模型的对比,选取在超临界条件下适用性相对较好的SST模型进行计算,分别比较不同热力当量直径和不同水力当量直径下圆管与圆环通道加热面壁温、边界层温度及速度的分布,研究热力当量直径和水力当量直径对超临界水流动传热特性的影响。结果表明,正常传热工况下,水力当量直径对超临界水流动传热特性有很大影响,而热力当量直径几乎无影响。圆环通道内流动传热关系式可基于圆管进行拟合,超临界水流动传热特性的其他影响因素还需进一步研究。     

Experimental investigation on heat transfer from a heated rod with a helically wrapped wire inside a square vertical channel to water at supercritical pressures

A supercritical water heat transfer test section has been built at Xi’an Jiaotong University to study the heat transfer from a 10 mm rod inside a square vertical channel with a wire-wrapped helically around it as a spacer. The test section is 1.5 m long and the wire pitch 200 mm. Experimental conditions included pressures of 23–25 MPa, mass fluxes of 500–1200 kg/m² s, heat fluxes of 200–800 kW/m², and inlet temperatures of 300–400 °C. Wall temperatures were measured with thermocouples at various positions near the rod surface. The experimental Nusselt numbers were compared with those calculated by empirical correlations for smooth tubes. The Jackson correlation showed better agreement with the test data compared with the Dittus-Boelter correlation but overpredicted the Nusselt numbers almost within the whole range of experimental conditions. Both correlations cannot predict the heat transfer accurately when deterioration occurred at low mass flux and relatively high heat flux in the pseudocritical region. Comparison of experimental data at two different supercritical pressures showed that the heat transfer was more enhanced at the lower supercritical pressure but the deterioration was more likely to occur at the higher pressure, meaning increased safety. Based on a comparison with an identical channel without the helical wrapped wire, it was found that the wire spacer does not enhance the heat transfer significantly under normal heat transfer conditions, but it contributes to the improvement of the heat transfer in the pseudocritical region and to a downstream shift of the onset of the deterioration. The Jackson buoyancy criterion is found to be valid and works well in predicting the onset of heat transfer deterioration occurring in the experiments without wire.     

Development of supercritical water heat-transfer correlation for vertical bare tubes

This paper presents an analysis of heat-transfer to supercritical water in bare vertical tubes. A large set of experimental data, obtained in Russia, was analyzed and a new heat-transfer correlation for supercritical water was developed. This experimental dataset was obtained within conditions similar to those in supercritical water-cooled nuclear reactor (SCWR) concepts.The experimental dataset was obtained in supercritical water flowing upward in a 4-m long vertical bare tube with 10-mm ID. The data were collected at pressures of about 24 MPa, inlet temperatures from 320 to 350 °C, values of mass flux ranged from 200 to 1500 kg/m² s and heat fluxes up to 1250 kW/m² for several combinations of wall and bulk-fluid temperatures that were below, at, or above the pseudocritical temperature.A dimensional analysis was conducted using the Buckingham Π-theorem to derive the general form of an empirical supercritical water heat-transfer correlation for the Nusselt number, which was finalized based on the experimental data obtained at the normal and improved heat-transfer regimes. Also, experimental heat transfer coefficient (HTC) values at the normal and improved heat-transfer regimes were compared with those calculated according to several correlations from the open literature, with CFD code and with those of the proposed correlation.The comparison showed that the Dittus-Boelter correlation significantly overestimates experimental HTC values within the pseudocritical range. The Bishop et al. and Jackson correlations tended also to deviate substantially from the experimental data within the pseudocritical range. The Swenson et al. correlation provided a better fit for the experimental data than the previous three correlations at low mass flux (∼500 kg/m² s), but tends to overpredict the experimental data within the entrance region and does not follow up closely the experimental data at higher mass fluxes. Also, HTC and wall temperature values calculated with the FLUENT CFD code might deviate significantly from the experimental data, for example, the k-? model (wall function). However, the k-? model (low Reynolds numbers) shows better fit within some flow conditions.Nevertheless, the proposed correlation showed the best fit for the experimental data within a wide range of flow conditions. This correlation has an uncertainty of about ±25% for calculated HTC values and about ±15% for calculated wall temperature. A final verification of the proposed correlation was conducted through a comparison with other datasets. It was determined that the proposed correlation closely represents the experimental data and follows trends closely, even within the pseudocritical range. Finally, a recent study determined that in the supercritical region, the proposed correlation showed the best prediction of the data for all three sub-regions investigated.Therefore, the proposed correlation can be used for HTC calculations in SCW heat exchangers, for preliminary HTC calculations in SCWR fuel bundles as a conservative approach, for future comparison with other datasets and for the verification of computer codes and scaling parameters between water and modelling fluids.     

Supercritical water heat transfer in vertical tubes: A look-up table

A new procedure for the prediction of wall temperatures in vertical tubes has been established in the framework of the development of the High Performance Light Water Reactor (HPLWR). The prediction of the wall temperature is accomplished by a look-up table for heat transfer in supercritical water. The look-up table lists the wall temperatures as a function of mass flux, heat flux, pressure, tube diameter, and bulk enthalpy. Based on an extended literature survey, experimental data for different conditions of upward flows in vertical smooth tubes are selected. To exclude data which exhibited local, buoyancy driven effects, a criterion of Jackson for deterioration of heat transfer is used to remove these data from further processing. An interpolation method is applied to assemble the tabulated grid points, based on published correlations for heat transfer in supercritical water. The look-up table covers a mass flux range of 700–3500 kg/m² s, a heat flux range of 300–1600 kW/m², a pressure range of 22.5–25 MPa, a diameter range of 8–20 mm and a bulk enthalpy range of 1200–2700 kJ/kg. Extreme combinations which required extrapolation of the data are excluded. The accuracy of the table in the vicinity of the pseudo-critical point is significantly higher than published correlations.     

Twisted-tape-induced swirl flow heat transfer and pressure drop in a short circular tube under velocities controlled

The twisted-tape-induced swirl flow heat transfer due to exponentially increasing heat inputs with various exponential periods (Q = Q₀ exp(t/τ), τ = 7, 14 and 23 s) and the twisted-tape-induced pressure drop were systematically measured with mass velocities, G, ranging from 4022 to 15,140 kg/m² s by an experimental water loop flow. Measurements were made on a 59.2 mm effective length which was spot-welded two potential taps on the outer surface of a 6 mm inner diameter, a 69.6 mm heated length and a 0.4 mm thickness of platinum circular test tube. The twisted tapes with twist ratios, y [=H/d = (pitch of 180° rotation)/d], of 2.39, 3.39 and 4.45 were used in this work. The relation between the swirl velocity and the pump input frequency and that between the fanning friction factor and Reynolds number (Re_d = 2.04 × 10⁴ to 9.96 × 10⁴) were clarified. The twisted-tape-induced swirl flow heat transfers with y = 2.39, 3.39 and 4.45 were compared with the values calculated by our correlation of the turbulent heat transfer for the empty tube and other worker's one for the circular tube with the twisted-tape insert. The influence of y and Reynolds numbers based on swirl velocity, Re_sw, on the twisted-tape-induced swirl flow heat transfer was investigated into details and the widely and precisely predictable correlation of the twisted-tape-induced swirl flow heat transfer was derived based on the experimental data. The correlation can describe for the twisted-tape-induced swirl flow heat transfer for the wide ranges of twist ratios (y = 2.39-4.45), mass velocities (G = 4022-15140 kg/m² s) and Reynolds numbers based on swirl velocity (Re_sw = 2.88 × 10⁴ to 1.22 × 10⁵) within −10 to +30% difference.     

Experimental investigations on single-phase heat transfer enhancement with longitudinal vortices in narrow rectangular channel

Four pairs of rectangular block as longitudinal vortex generators (LVG) were mounted periodically in a narrow rectangular channel to investigate fluid flow and convective heat transfer respectively in the narrow rectangular channel with LVG and without LVG. Both the channels have the same narrow gap (d) = 3 mm, the same hydraulic diameter (D_h) = 5.58 mm and the same length to diameter ratio (L/D_h) = 80.65. The experiments were performed with the channels oriented uprightly and uniform heat fluxes applied at the one side of the heating plate and single-phase water was used as test fluid. The parameters that were varied during the experiments included the mass flow rate, inlet liquid temperature, system pressure, and heat flux.In each of the experiments conducted, the temperature of both the liquid and the wall was measured at various locations along the flow direction. Based on the measured temperatures and the overall energy balance across the test section, the heat transfer coefficients for single-phase forced convection have been calculated. At the same time, in these experiments, the single-phase pressure drop across the channels was also measured. The correlations have been developed for mean Nusselt numbers and friction factors. Additionally, the visual experiments of infrared thermo-image recording the temperature on the outer wall of the heating plate have been conducted for validating the effects of LV.In these experimental investigations, both laminar regime and turbulent regime were under the thermo-hydraulic developing conditions, laminar-to-turbulent transition occurred in advance with the help of LV when Reynolds numbers vary between 310 and 4220. In laminar regime, LV causes heat transfer enhancement of about 100.9% and flow resistance increase of only 11.4%. And in turbulent regime, LV causes heat transfer enhancement of above 87.1% and flow resistance increase of 100.3%. As a result, LV can obviously enhance heat transfer of single-phase water, and increase flow resistance mildly.     

超临界压力下水在方形环腔内垂直上升的传热特性

在西安交通大学超临界传热试验台上研究了超临界压力下水在方形环腔中垂直上升的传热特性.试验压力23～25 MPa;质量流速500～1 200kg/(m~2·s);热流密度200～800 kW/m~2;工质进口温度300～400℃.试验结果表明:带绕丝固定的方形环腔结构在高质量流速低热负荷的情况下,在拟临界区域传热会得到强化,而在低质量流速高热负荷的情况下,会发生传热恶化现象;较低的超临界压力下会有更加突出的传热强化表现,但是传热恶化会提前发生,并且更加剧烈,因此较高的超临界压力意味着安全性更高.     

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.     

Combined Forced and Free Convective Heat Transfer Characteristics in a Narrow Vertical Rectangular Channel with 2.5 mm in Gap Heated from Both Sides

Heat transfer characteristics of mainly combined forced and free convective flow in a vertical rectangular flow channel with a gap of 2 5 mm, which was quite narrow compared with those investigated in previous experiments, were studied experimentally for water. As a result, the following heat transfer characteristics were made clear, using a non-dimensional parameter Gr _x /Re ²¹ _x /⁸ Pr ^1/2 similarly to the case for the 18 mm gap which was already reported by the authors.

(1) When the Gr _x /Re ²¹ _x /⁸ Pr ^1/2 is less than 10^?4, both upward flow and downward flow show the nature of forced convective heat transfer.

(2) When the Gr _x /Re ²¹ _x /⁸ Pr ^1/2 is between 10^?4 and 10^?2, heat transfer coefficients for both upward flow and downward flow are higher than any of those predicted by the previous correlations for turbulent forced convection along a flat plate and turbulent free convection along a vertical flat plate. This is, differently from the case of 18 mm gap, due to the effect of the acceleration of main flow induced by the development of the boundary layer along the channel.

(3) When the Gr _x /Re ²¹ _x /⁸ Pr ^1/2 is larger than 10^?2, the upward flow shows the nature of free convective flow even with the gap as narrow as 2.5 mm in the vertical rectangular flow channel. Heat transfer correlations which have been developed for the 18 mm gap channel, are also available for the described-above regions of 2.5 mm gap channel.     

Analysis of the conjugation effect during convective flow film boiling conditions in an eccentric annular test section

The objective of this investigation was to present a technique for estimating the conjugation effect on post-dryout heat transfer. To this aim, an experimental study was conducted to measure the conjugation effect on the post-dryout heat transfer using an internally heated eccentric annular test section (outside diameter (O.D.), 13.51 mm; inside diameter (I.D.), 11.54 mm). The experiments were carried out at 7 and 9 MPa with the mass flux varying from 2.0 to 5.0 Mg m⁻² s⁻¹ and the inlet vapour quality from 0.01 to 0.20. Five different minimum gap sizes between 0.06 and 1.92 mm were tested. The two-dimensional well temperature distribution on the inside surface of the heated tube was measured using a unique sliding thermocouple technique. A data reduction method was developed to determine the radial heat flux variation on the boiling surface from the temperature measurements. A numerical smoothing procedure was used to minimize the effect of the random temperature measurement error of the sliding thermocouple on the estimated radial heat flux variation. The results show that the conjugation effect can cause the local radial heat flux to deviate by as much as 17% from the average value.     

Primary research of heat flux deposition pattern on the surface of HT-7 movable limiter

This paper presents a method of characterizing the heat flux deposition pattern on HT-7 movable limiter, a new component in the spring experimental campaign of year 2006. A new modeling establishment combines with heat transfer simulation with ANSYS code, and a shot of long pulse discharge is chosen with small radius 265 mm of movable limiter, which is 5 mm smaller than that of the main toroidal limiters. Both parallel and perpendicular heat flux are taken into account and different ratios of them are also tested in simulations. The simulation temperature values match well with the results of the IR-camera. Temperature distribution shows that ?_||/?_⊥ = 5 is quite suitable in HT-7 device, and the maximum heat flux is about 10 MW/m² and average heat flux is about 5 MW/m² on the movable limiter. This method can be also used in other tokamak devices with limiter configuration and circular cross-section shape.     

Subcooled boiling heat transfer in a short vertical SUS304-tube at liquid Reynolds number range 5.19 × 10 to 7.43 × 10

The subcooled boiling heat transfer and the steady-state critical heat fluxes (CHFs) in a short vertical SUS304-tube for the flow velocities (u = 17.28-40.20 m/s), the inlet liquid temperatures (T_in = 293.30-362.49 K), the inlet pressures (P_in = 842.90-1467.93 kPa) and the exponentially increasing heat input (Q = Q₀ exp(t/τ), τ = 8.5 s) are systematically measured by the experimental water loop comprised of a multistage canned-type circulation pump with high pump head. The SUS304 test tubes of inner diameters (d = 3 and 6 mm), heated lengths (L  =  33 and 59.5 mm), effective lengths (L_eff = 23.3 and 49.1 mm), L/d (=11 and 9.92), L_eff/d (=7.77 and 8.18), and wall thickness (δ = 0.5 mm) with average surface roughness (Ra = 3.18 μm) are used in this work. The inner surface temperature and the heat flux from non-boiling to CHF are clarified. The subcooled boiling heat transfer for SUS304 test tube is compared with our Platinum test tube data and the values calculated by other workers’ correlations for the subcooled boiling heat transfer. The influence of flow velocity on the subcooled boiling heat transfer and the CHF is investigated into details and the widely and precisely predictable correlation of the subcooled boiling heat transfer for turbulent flow of water in a short vertical SUS304-tube is given based on the experimental data. The correlation can describe the subcooled boiling heat transfer obtained in this work within 15% difference. Nucleate boiling surface superheats for the SUS304 test tube become very high. Those at the high flow velocity are close to the lower limit of Heterogeneous Spontaneous Nucleation Temperature. The dominant mechanisms of the flow boiling CHF in a short vertical SUS304-tube are discussed.     

同心环形管内沸腾两相流动与传热实验研究

对窄缝为2．1mm的同心环形管，试验研究了外管加热条件下水的沸腾两相流动阻力与传热特性，得到了以下结果：窄缝环形管内两相流动的阻力较普通圆管内大，沸腾换热得到了较明显的强化，换热系数弓压力、热平衡干度、工质流量、加热负荷均有关系，且与缝隙宽度和加热方式有关；提出了环形管强化传热的微液膜蒸发机理与汽泡扰动机理的物理解释；得到了环形管内流动摩擦阻力系数与传热系数的实验关联式。     

Heat transfer to a supercritical pressure fluid flowing in sub-bundle channels

Supercritical pressure water cooled reactor (SCWR) has been regarded as an innovative nuclear reactor. For the design and development of the SCWR, heat transfer performance under supercritical pressure is one of the most important indicators. In this paper, experimental data are presented on the heat transfer to a supercritical pressure fluid flowing vertically upward and downward in a small diameter heated tube and two sub-bundle channels with three heater rods and seven heater rods, using HCFC22 as the test fluid. Downstream of grid spacer for the sub-bundles, heat transfer enhancement was observed in the upward flow, but not in the downward flow. The enhancement was remarkable especially when the heat transfer deterioration occurs in the fully developed region unaffected by the spacer. The heat transfer correlation for the downstream region of the spacer was developed in the normal heat transfer of sub-bundles. In the fully developed region for the sub-bundle, occurrence of the heat transfer deterioration was suppressed or degree of the deterioration was moderated. At high mass velocity for downward flow in the seven rod sub-bundle, oscillation of heat transfer was observed in the region of the enthalpy over the pseudocritical point.