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Takashi Hibiki - One of the best experts on this subject based on the ideXlab platform.
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experimental study on local interfacial parameters in upward air water bubbly flow in a vertical 6 6 Rod Bundle
International Journal of Heat and Mass Transfer, 2019Co-Authors: Xu Han, Xiuzhong Shen, Toshihiro Yamamoto, Ken Nakajima, Haomin Sun, Takashi HibikiAbstract:Abstract This paper presents a complete database of the local two-phase flow parameters for upward adiabatic air-water two-phase flows in a vertical 6 × 6 Rod Bundle flow channel. A total of 16 flow conditions were selected and investigated at 16 measuring points in the octant triangular measuring region of a cross section at the position with height-to-diameter ratio of 149 (z/DH = 149) in the Rod Bundle experiment. The local void fraction, interfacial area concentration (IAC), bubble diameter and bubble velocity vector were measured by using the four-sensor optical probe measuring technique. In the measured cross-sectional distributions of local void fraction and IAC, both core-peaking and wall-peaking distribution shapes were observed and the distribution pattern trended to change from the core-peaking shape to the wall-peaking shape with the increase of superficial liquid velocity ( 〈 j f 〉 ) and the decrease of superficial gas velocity ( 〈 j g 〉 ). The radial distributions of local bubble diameter are nearly flat, which explains the similarity of the radial local void fraction and IAC distributions. The measured local bubble diameters increase with the increase of 〈 j g 〉 and the decrease of 〈 j f 〉 . The bubbles move along the Rod Bundle flow channel with their velocity component in the main flow direction ( v gz ) showing a typical radial core-peaking profile and their velocity components in the cross section showing a significant movement from the central region to the channel box wall region especially under high superficial liquid velocities. The cross-sectional area-averaged results of the local parameters were obtained by a measuring-point-based graphical integration. The obtained area-averaged void fraction and IAC are compared with the predictions of the drift-flux model with its available 2 frequently-used distribution parameter and drift velocity correlations and the 2 recently-developed IAC correlations respectively. The applicability of the 2 drift-flux correlations and the 2 IAC correlations to the Rod Bundle flow channel is evaluated and analyzed.
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Local measurements of upward air-water two-phase flows in a vertical 6×6 Rod Bundle
Experimental and Computational Multiphase Flow, 2019Co-Authors: Xiuzhong Shen, Shuichiro Miwa, Yigeng Xiao, Takashi HibikiAbstract:A 3 m high vertical Rod Bundle flow channel consisting of 6×6 Rods with the Rod diameter of 10 mm and the pitch of 16.7 mm and a channel box with the cross section of 100 mm × 100 mm was used to study the local two-phase flow characteristics under influences of the 36 Rods and the channel box. The air and water were selected as the two-phase working fluids in the present experiments. A double-sensor probe installed at the axial position of z / D _H = 149 and six evenly-installed differential pressure (DP) gauges measured the upward-moving two-phase flow in the Rod Bundle flow channel under the atmospheric condition. Since the local parameters of void fraction, interfacial area concentration (IAC), bubble diameter, and gas velocity are essential to know the internal structures of the two-phase flow, their data at 16 points within an octant symmetric triangular area of the flow channel cross section were collected by the double-sensor probe under various flow conditions in the experiments. The local measurements of double-sensor probe were found to agree well with the void fractions from the DP gauges and the superficial gas velocity from a gas flowmeter. Both the measured void fractions and IACs displayed a transition from radial wall-peaking profile to radial core-peaking profile in the low superficial liquid velocity flow conditions and pure radial wall-peaking profiles in the high superficial liquid velocity flow conditions. The measured Sauter mean diameters showed their radial wall-peaking profiles with the peaking degree decreasing with the increasing superficial gas velocity in both the low and high superficial liquid velocity flow conditions. The measured gas velocities in the main flow direction showed a transition from a radial nearly-flat profile or a radial mid-peaking profile to radial core-peaking profiles in the low and high superficial liquid velocity flow conditions. The area-averaged void fractions and IACs integrated from their measured local values were respectively compared with the predictions of existing drift-flux and IAC correlations. The comparison results showed that the drift-flux correlation of Ozaki et al. (2013) and the IAC correlation of Shen and Deng (2016) give the reasonable prediction and can be recommended for the predictions of the void fraction and the IAC respectively in the bubbly flows in the Rod Bundle flow channel.
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constitutive equations for vertical upward two phase flow in Rod Bundle
International Journal of Heat and Mass Transfer, 2018Co-Authors: Takashi Hibiki, Tetsuhiro Ozaki, Ikuo Kinoshita, Xiuzhong Shen, Shuichiro Miwa, Tatsuya Hazuku, Somboon RassameAbstract:Abstract In view of the quality assurance of two-phase flow simulations, CSAU (Code Scalability, Applicability, and Uncertainty) methodology and code V & V (Verification and Validation) have been proposed. The estimation of simulation uncertainty is indispensable in using best-estimate computational codes. A key of successful two-phase flow simulations is to use the state-of-the-art constitutive equations to close the mathematical system used in two-phase flow analyses. The advanced constitutive equations should be developed based on “physics” behind phenomena and should consider scaling parameters which enable their application beyond test conditions used for a code validation. Two-phase flow simulations in a Rod Bundle is important in various industrial apparatuses such as heat exchangers and nuclear reactors. Constitutive equations for two-phase flows in a vertical Rod Bundles have been advanced in recent five years. In view of this, this paper provides a comprehensive review of most advanced constitutive equations for two-phase flow analyses in a vertical Rod Bundle. The constitutive equations of two-phase flow parameters reviewed in this paper are flow regime map, void fraction, void fraction covariance and relative velocity covariance, interfacial area concentration and wall friction. In addition, an exact formulation of one-dimensional momentum equation in two-fluid model considering void fraction distribution is discussed.
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prediction of Rod film thickness of vertical upward co current adiabatic flow in Rod Bundle
Annals of Nuclear Energy, 2018Co-Authors: Yang Liu, Mamoru Ishii, Takashi HibikiAbstract:Abstract Prediction of Rod film thickness and its distribution in annular flow in Rod Bundle is of essential importance to nuclear power industry. This is attributed to the possible film dryout on fuel Rods under certain conditions, which will result in a drastically increasing cladding temperature and possible fuel damage. Though there exist a lot of models or correlations to predict film thickness for annular flow in pipes, there are few models to predict Rod film thickness inside Rod Bundle. The one-dimensional models or correlations for pipe flow cannot be used for Rod Bundle directly due to different geometry and flow dynamics. In relation to this, a new approach to predict the Rod film thickness inside Rod Bundle has been developed. This approach first divides the Rod Bundle cross-section into three regions according to their characteristic length scales. Then, the gas and liquid flow distribution inside Rod Bundle has been given with reasonable assumptions. Based on the flow distribution profile, the sub-channel area-averaged gas and liquid volumetric fluxes have been given. Finally, the correlation developed for pipe flow is adopted for each sub-channel based on the local two-phase parameters in the Rod Bundle. Compared to experimental data obtained in-house and found in the literature, the newly proposed approach can predict Rod film thickness with satisfactory accuracy.
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average liquid film thickness of annular air water two phase flow in 8 8 Rod Bundle
International Journal of Heat and Fluid Flow, 2018Co-Authors: Peng Ju, Xiaohong Yang, Joshua P Schlegel, Takashi Hibiki, Mamoru IshiiAbstract:Abstract In order to develop detailed model of the annular two-phase flow the average liquid film thickness is an important parameter. It can significantly affect the occurrence of dryout and post-dryout phenomenon on heater surfaces. Most research on film thickness has been focused on pipe flows. Data in Rod Bundle geometry are very limited. However data in Rod Bundle geometry is much more valuable for reactor safety modeling and calculations. Because of this, an experiment to measure liquid film thickness has been performed for the air-water annular flow in a 8 × 8 BWR Rod Bundle. Film thickness data were obtained both on the Rods and on the channel wall. Data were recorded at four axial locations within the Rod Bundle. This included locations just before and just after a spacer grid. This allowed the spacer grids effect on the film thickness to be evaluated. The minimum film thickness was on the center Rod. Also, the presence of the spacer grid resulted in reduced film thickness.
Xu Cheng - One of the best experts on this subject based on the ideXlab platform.
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experimental study of cross flow and lateral pressure drop in a 5 5 Rod Bundle with mixing vane spacer grid
Nuclear Engineering and Design, 2019Co-Authors: Zefeng Wang, Jinbiao Xiong, Xu ChengAbstract:Abstract Cross flow and lateral pressure drop downstream a split-type mixing vane spacer grid in a 5 × 5 Rod Bundle were experimentally measured by particle image velocimetry (PIV) and differential pressure transmitter at the sub-channel Reynolds number (Re) of 13200. The geometrical parameters of the 5 × 5 Rod Bundle is as the same diameter (D = 9.5 mm) and pitch (P = 12.6 mm) as the real fuel Rods of a typical pressurized water reactor (PWR), with sub-channel hydraulic diameter (Dh) 11.78 mm. With the aid of matched index of refraction (MIR) technique of transparent fluorinated ethylene propylene (FEP) tubes and water, the PIV measurements of cross flow covering all sixteen inner-sub-channels were conducted for cross sections at 1Dh, 4Dh, 8Dh and 16Dh downstream the spacer grid. With detailed error and uncertainty analysis, the uncertainty of the turbulent statistics is mainly determined by the out of plane error and sampling number error. The maximum standard uncertainty of statistical quantities normalized by the bulk velocity is less than 1%. The characteristics of cross flow were mainly determined by the mixing vanes of spacer grid in Rod Bundle. The mixing vanes generate strong cross flow and drastic turbulence flow downstream the spacer grid, enhancing the cross flow mixing and turbulent mixing between sub-channels. The lateral pressure drop was measured at 1Dh, 2Dh, 3Dh, 4Dh, 5Dh, 6Dh, 8Dh, 10Dh, 12Dh, 14Dh, 16Dh, 18Dh and 20Dh. The relative uncertainty level of lateral pressure drop is between 0.7% and 7.0%. Downstream the spacer grid, the lateral pressure drop increases from 1Dh to 2Dh. After 3Dh, the lateral pressure drop decays in exponential way with distance from spacer grid. After 16Dh, the lateral pressure drop almost keeps constant.
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piv measurement of turbulent flow downstream of mixing vane spacer grid in 5 5 Rod Bundle
Annals of Nuclear Energy, 2019Co-Authors: Jinbiao Xiong, Shilong Chen, Zhifang Qiu, Jian Deng, Xu ChengAbstract:Abstract Axial and cross flow downstream a split-type mixing vane spacer grid in a 5×5 Rod Bundle are measured with particle image velocimetry (PIV) at sub-channel Reynolds number (Resc) of 6600, 13,200, 26,400 and 39,600. The cross flow in the sixteen inner sub-channels are obtained at five elevations, i.e. 3Dh, 5Dh, 10Dh, 15Dh and 20Dh downstream of the spacer grid. The axial flow measurement is conducted in a plane parallel to the Bundle axis. The satisfactory convergence of turbulent statistics is achieved with the sampling number of 2000. The maximum standard uncertainty of statistical quantities normalized by the bulk velocity for all the test conditions is less than 0.6%. The flow characteristics and structures were determined by the mixing vanes located on the upper edge of support grids in Rod Bundle, and the flow downstream the spacer grid can be divided into the advection zone, the reorganization zone and the dumping zone. And the maximum lateral velocity decreases from 44% of the bulk velocity (Wb) at 3Dh to 16%Wb at 20Dh. The mixing vanes generate drastic turbulence flow downstream the spacer grid, and the lateral root mean square (RMS) fluctuating velocity decays in exponential way. The maximum lateral RMS fluctuating velocity decays rapidly from 20%Wb at 3Dh to 9%Wb at 20Dh. The maximum lateral Reynolds stress decreases from 0.6% W b 2 at 3Dh to 0.3% W b 2 at 20Dh. The normalized mean velocity and turbulent statistics show similarity for the Reynolds above 13,200, while the low Reynolds number effect at Re = 6600 is observed for the axial mean velocity and axial RMS fluctuating velocity.
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high fidelity piv measurement of cross flow in 5 5 Rod Bundle with mixing vane grids
Nuclear Engineering and Design, 2019Co-Authors: Jinbiao Xiong, Shilong Chen, Xu ChengAbstract:Abstract In order to achieve high-fidelity PIV measurement for cross flow in Rod Bundles, the measurement approach with the two-dimensional (2D) PIV has been optimized systematically and the systematic and random errors are analyzed. The optimizations, covering the design of test section, the setting of optical system, the PIV parameter configuration and the PIV correlation algorithm, were established based on massive sensitivity analysis. Utilizing the optimized PIV configuration, cross flow measurements were conducted in a 5 × 5 Rod Bundle with the Rods 9.5 mm in diameter and the Rod pitch of 12.6 mm, which is the typical configuration of fuel assembly in pressurized water reactors (PWRs). The FEP tubes filled with deionized water are employed as the matched index of refraction (MIR) Rods to realize uniform illumination across the Rod Bundle. On each investigated cross section the cross flow was measured in the 16 inner subchannels. In order to maximize the magnification factor, four subchannels are measured in a single test run. Several substantial error sources were identified and accounted for in the error and uncertainty analysis. The systematic error consists mainly of the perspective error and displacement variation error, which were quantified and used to correct the measured results. The uncertainty of RMS velocity is the combination of random error and error due to limited sampling number, while the uncertainty of mean velocity equals to the error resulted from limited sampling number assuming that the systematic error is perfectly compensated. In the experiment the relative standard uncertainty of RMS velocity (to the bulk velocity) is within 0.6%, and the relative standard uncertainty of mean velocity (to the bulk velocity) is less than 0.4%. The measurements were carried out with the subchannel Reynolds number (Re) of 39600 at five cross sections, i.e. 3Dh, 5Dh, 10Dh, 15Dh and 20Dh downstream of a spacer grid. The experimental results showed the detailed cross flow features, including the evolution of vortices, in the Rod Bundle. The measured lateral flow was substantially dominated by the mixing vanes. Drastic turbulence generation by mixing vanes was observed downstream of a spacer grid. The generated turbulence decayed rapidly within 10Dh downstream of the spacer grid and then damped gradually.
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cfd simulation of swirling flow induced by twist vanes in a Rod Bundle
Nuclear Engineering and Design, 2018Co-Authors: Jinbiao Xiong, Ruiqi Cheng, Xiang Chai, Chuan Lu, Xu ChengAbstract:Abstract Swirling flow induced by the twist vanes can enhance the heat transfer in the Rod-Bundle fuel assembly. In order to investigate accuracy of the k-e models in predicting the swirling flow in a Rod Bundle subchannel, the CFD simulations with the standard and realizable k-e model are benchmarked with In et al.’s experiment. The nonlinear closure models are utilized with the standard k-e model to account for the curved and rotational flow, while the curvature correction model is employed in the realizable k-e model. Comparison with the experiment indicates that the CFD simulations predict the relatively large core region of swirling flow. Utilization of the nonlinear closure models in the standard k-e model leads to the larger initial angular momentum. The curvature correction in the realizable k-e model reduces the initial angular momentum, but it does not affect the decay rate, which indicates that the curvature correction is effective near the twist vanes where the flow curvature is strong. The decay rate of swirling flow is generally overestimated by all the models. The cross-flow velocity profile in the gap is affected by prediction of swirling flow separation from Rod surface. The cubic and quadratic closure model can improve turbulence modelling in the core of swirling flow, especially near the spacer, where the linear closure models significantly underestimate the pRoduction of turbulence. However, in the further downstream all the models underestimate the turbulence intensity in the swirling core, as a consequence of low-frequency large-scale flow structure in the core region. Underestimation of turbulence in the annular and wall region will result in the underestimation of wall heat transfer and inter-subchannel mixing.
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piv measurement of cross flow in a Rod Bundle assisted by telecentric optics and matched index of refraction
Annals of Nuclear Energy, 2018Co-Authors: Jinbiao Xiong, Xu ChengAbstract:Abstract Spacer grids induce local flow disturbance in fuel assembly including lateral cross flow and elevated turbulence fluctuation. In order to investigate the cross flow downstream of spacer grids and to provide the high-fidelity data for the CFD validation of Rod Bundle flow, a flow measurement method has been proposed to measure the flow in the full cross section of the 5 × 5 Rod Bundle. A two-dimensional (2D) particle image velocimetry (PIV) system is utilized for the flow measurement. A telecentric lens is employed to minimize the perspective angle when the CCD camera is located distant from the measured cross section. The matched refractive index technique is also intRoduced to eliminate the non-uniform laser illumination behind the Rods which is resulted from the laser refraction on the convex Rod surface. Such configuration is examined for the cross flow measurement of a 5 × 5 Rod Bundle in a small scale test facility. It has been shown that the cross flow pattern can be well depicted in the entire cross section when the measuring plane is within 10 Dh downstream of a spacer grid. The obtained distribution of sampled flow velocity appears in good agreement with the normal distribution.
Suizheng Qiu - One of the best experts on this subject based on the ideXlab platform.
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large eddy simulation on the turbulent mixing phenomena in 3 3 bare tight lattice Rod Bundle using spectral element method
Nuclear Engineering and Technology, 2020Co-Authors: Mingjun Wang, Wenxi Tian, Yingjie Wang, Minfu Zhao, Tiancai Liu, Suizheng QiuAbstract:Abstract Subchannel code is one of the effective simulation tools for thermal-hydraulic analysis in nuclear reactor core. In order to reduce the computational cost and improve the calculation efficiency, empirical correlation of turbulent mixing coefficient is employed to calculate the lateral mixing velocity between adjacent subchannels. However, correlations utilized currently are often fitted from data achieved in central channel of fuel assembly, which would simply neglect the wall effects. In this paper, the CFD approach based on spectral element method is employed to predict turbulent mixing phenomena through gaps in 3 × 3 bare tight lattice Rod Bundle and investigate the flow pulsation through gaps in different positions. Re = 5000,10000,20500 and P/D = 1.03 and 1.06 have been covered in the simulation cases. With a well verified mesh, lateral velocities at gap center between corner channel and wall channel (W–Co), wall channel and wall channel (W–W), wall channel and center channel (W–C) as well as center channel and center channel (C–C) are collected and compared with each other. The obvious turbulent mixing distributions are presented in the different channels of Rod Bundle. The peak frequency values at W–Co channel could have about 40%–50% reduction comparing with the C–C channel value and the turbulent mixing coefficient β could decrease around 25%. corrections for β should be performed in subchannel code at wall channel and corner channel for a reasonable prediction result. A preliminary analysis on fluctuation at channel gap has also performed. Eddy cascade should be considered carefully in detailed analysis for fluctuating in Rod Bundle.
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an experiment study of pressure drop and flow distribution in subchannels of a 37 pin wire wrapped Rod Bundle
Applied Thermal Engineering, 2020Co-Authors: Yu Liang, Kui Zhang, Dalin Zhang, Yutong Chen, Wenxi Tian, Suizheng QiuAbstract:Abstract The experiments of pressure drop and flow distribution in subchannels of a 37-pin wire-wrapped Rod Bundle have been performed to evaluate the reliability of existing friction factor correlations and provide available experimental data to validate the accuracy of the thermal hydraulic codes for the design of the lead-based reactor. The isokinetic sampling technique was applied to measure the flow rate of each subchannel at the Rod Bundle outlet. The experiments were carried out within the flow rate range of 0.6–5.0 kg/s and the inlet temperature range of 20–80 °C. The Reynolds number of these experiments covers the range of 1100–22,000. The CFD pre-analysis revealed that the number of Rods has little effect on the subchannel friction factor. The pressure drop measuring section belongs to the fully developed region. The friction factors and flow split factors measured were compared with the existing friction factor correlations. The prediction results of the Chiu-Rohsenow-Todreas (CRT) model agree well with the experimental results of subchannels. The upgraded detailed Cheng and Todreas (UCTD) model is the best fit wire-wrapped Rod Bundle friction factor correlation and the maximum error is 15% between Reynolds numbers 1100 and 22,000. When it comes to the interior subchannel, the experimental data falls within 50% of the UCRD model between interior subchannel Reynolds numbers 900 and 13,000. The updated correlations for predicting Bundle transition Reynolds numbers are presented in this paper. The subchannel friction factor and the flow split factor are coupled with each other. The derivation of the subchannel transition points proves that the pressure drop experiment and the flow distribution experiment are self-consistent.
Da Liu - One of the best experts on this subject based on the ideXlab platform.
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experimental study on onset of nucleate boiling and flow boiling heat transfer in a 5 5 Rod Bundle at low flow rate
International Journal of Heat and Mass Transfer, 2019Co-Authors: Shuo Chen, Da Liu, Yao XiaoAbstract:Abstract Onset of nucleate boiling and flow boiling heat transfer are experimentally investigated in a 5 × 5 Rod Bundle at low flow rate under the pressure of 6 MPa. The experimental mass flux ranges from 25 to 200 kg/m2 s, heat flux varies from 25 to 75 kW/m2, the maximum vapour quality is 0.45. Effects of parameters on ONB and flow boiling are examined. The wall superheating at ONB point increases with the rise of heat flux while decreases with the increase of mass flux. The effects of mass flux and quality on flow boiling heat transfer coefficients are not distinct. For both subcooled and saturated flow boiling in the 5 × 5 Rod Bundle, the effect of heat flux on heat transfer is more remarkable at high flow rate than at low flow rate. Comparisons between predictions by existing empirical correlations and the experimental data are conducted. Basu’s correlation could predict the wall superheating at ONB point within a deviation less than 30% under conditions of higher mass flux (>150 kg/m2 s). As for the comparisons in flow boiling regions, predictions by Chen’s correlation agree well with the experimental data in both subcooled and saturated regions, except under conditions of low mass flux with low heat flux. Most of the correlations for boiling heat transfer and ONB are derived from the experimental data obtained at high flow rate, however, current experiments indicate that these correlations cannot be directly used under low flow rate conditions.
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experimental investigation on heat transfer behavior in a tight 19 Rod Bundle cooled with supercritical r134a
Annals of Nuclear Energy, 2018Co-Authors: Jiayue Chen, Zhenqin Xiong, Da LiuAbstract:Abstract The strong non-uniform distributions of circumferential wall temperature and heat transfer coefficient will exist in tight-lattice Rod Bundles. However, there is still a lack of understanding and only a few experimental data are available. In this paper, experimental investigations are performed on circumferential heat transfer behavior in a tight hexagonal 19 Rod Bundle using supercritical R134a as working fluid. The Rod surface temperatures at various circumferential positions along the axial flow direction are measured. A defined local hydraulic diameter is intRoduced to represent the heterogeneity of channels in the tight Bundle. The circumferentially non-uniform distribution of wall temperature is found to be strongly related to the variation of the local hydraulic diameter. In normal heat transfer pattern, the circumferential temperature gradient is large especially in the region far away the pseudo critical point, but becomes small near the pseudo critical point due to a strong enhancement of heat transfer coefficient. Parametric effects of pressure, mass flux and heat flux on circumferential non-uniformity of wall temperature inside the Rod Bundle are also analyzed.
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experimental study on heat transfer of supercritical water flowing upward and downward in 2 2 Rod Bundle with wrapped wire
Annals of Nuclear Energy, 2018Co-Authors: Jiaqi Tao, Da LiuAbstract:Abstract Heat transfer to supercritical water flowing upward and downward in a 2 × 2 Rod Bundle with wrapped wire was experimentally investigated. The Rod Bundle consists of four heating Rods within out-diameter of 10 mm. The Rods are arranged in a 2 × 2 square array and the pitch-to-diameter ratio is 1.18. The Bundle is inserted into a square assembly box with round corners. The water as coolant flows upward or downward in the channel between the Rod Bundle and the assembly box. The pressure ranges from 23 to 26 MPa, mass flux ranges from 450 to 1200 kg/(m 2 s), heat flux ranges from 200 to 1000 kW/m 2 , and the fluid temperature ranges from 200 to 450 °C. The effects of system parameters including mass flux, heat flux and pressure on the heat transfer are presented, respectively. The heat transfer deterioration occurs at a low G/q ratio in the upward flow, but does not appear in the downward flow. Watts-Chou correlation and Bishop correlation give the best predictions when evaluated against the experimental data both in the upward and downward flow. The experimental data are compared with the bare Rod Bundle. The comparison indicates that the swirl flow of wrapped wire promotes the heat transfer in the Rod Bundle.
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experimental study on heat transfer to supercritical water in 2 2 Rod Bundle with wire wraps
Experimental Thermal and Fluid Science, 2016Co-Authors: Da Liu, Meng Zhao, Xu ChengAbstract:Abstract Experimental studies on heat transfer to supercritical water in 2 × 2 Rod Bundle with wire wraps are performed at Shanghai Jiaotong University. The test section consists of two channels separated by a square steel assembly box with round corners. Water flows downward in the first channel and then turns upward in the second channel to cool the 2 × 2 Rod Bundle with wire wraps installed inside the assembly box. The Bundle consists of four heater Rods of 10 mm in O.D. and 1.18 in. pitch-to-diameter ratio. Experimental parameter ranges cover pressure from 23.0 MPa to 26.0 MPa, mass flux from 400 kg/m 2 s to 1400 kg/m 2 s, heat flux from 300 kW/m 2 to 1000 kW/m 2 and bulk temperature from 280 °C to 500 °C. The fluid enthalpy in the first channel increases due to the heat transfer through the assembly box when flowing downward. Heat transfer deterioration is observed in the Bundle. The degree of the deterioration is moderated in the developed region due to the mixing by the wire wraps. Significant non-uniformity of circumferential wall-temperature distribution around the heater Rods is observed. Effects of various parameters on heat transfer behavior inside the 2 × 2 Rod Bundle are similar to those observed in a tube. The heat transfer enhancement due to the wire wraps becomes remarkable under a high mass flux condition. The Jackson and Fewster correlation and Bishop et al. correlation give the best predictions when evaluated against the experimental data.
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heat transfer to supercritical water in a 2 2 Rod Bundle
Annals of Nuclear Energy, 2015Co-Authors: Da Liu, Meng ZhaoAbstract:Abstract An experimental study on heat transfer to supercritical water flowing in a 2 × 2 Rod Bundle test section with two channels is carried out at Shanghai Jiao Tong University. The Bundle consists of four heated Rods with an OD of 10 mm and a pitch-to-diameter ratio of 1.18. The 2 × 2 Rod Bundle is inserted into a square assembly box with rounded corners by which the test section is separated into two channels. Water flows downward in the first channel between the pressure tube and the assembly box and then turns upward in the second channel inside the assembly box to cool the four heated Rods, which are directly heated by DC power. The experimental conditions are as follows: pressure ranging from 23 to 26 MPa, mass flux ranging from 400 to 1200 kg/m 2 s, heat flux between 300 and 1000 kW/m 2 , and bulk fluid temperatures ranging from 200 to 480 °C. The heat transfer through the assembly box is strongly affected by the temperature difference between the two channels. The minimum fluid enthalpy increase in the first channel appears near the pseudo-critical temperature region. Non-uniformity of circumferential wall-temperature distribution around the heated Rod is observed. The effects of various parameters on heat transfer behavior inside the 2 × 2 Rod Bundle are similar to effects observed in tubes. Six developed correlations are compared with the test data. The correlations of Jackson and Fewster (1975) and Bishop et al. (1964) provide the best predictions of the test data among the selected correlations.
Mamoru Ishii - One of the best experts on this subject based on the ideXlab platform.
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prediction of Rod film thickness of vertical upward co current adiabatic flow in Rod Bundle
Annals of Nuclear Energy, 2018Co-Authors: Yang Liu, Mamoru Ishii, Takashi HibikiAbstract:Abstract Prediction of Rod film thickness and its distribution in annular flow in Rod Bundle is of essential importance to nuclear power industry. This is attributed to the possible film dryout on fuel Rods under certain conditions, which will result in a drastically increasing cladding temperature and possible fuel damage. Though there exist a lot of models or correlations to predict film thickness for annular flow in pipes, there are few models to predict Rod film thickness inside Rod Bundle. The one-dimensional models or correlations for pipe flow cannot be used for Rod Bundle directly due to different geometry and flow dynamics. In relation to this, a new approach to predict the Rod film thickness inside Rod Bundle has been developed. This approach first divides the Rod Bundle cross-section into three regions according to their characteristic length scales. Then, the gas and liquid flow distribution inside Rod Bundle has been given with reasonable assumptions. Based on the flow distribution profile, the sub-channel area-averaged gas and liquid volumetric fluxes have been given. Finally, the correlation developed for pipe flow is adopted for each sub-channel based on the local two-phase parameters in the Rod Bundle. Compared to experimental data obtained in-house and found in the literature, the newly proposed approach can predict Rod film thickness with satisfactory accuracy.
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average liquid film thickness of annular air water two phase flow in 8 8 Rod Bundle
International Journal of Heat and Fluid Flow, 2018Co-Authors: Peng Ju, Xiaohong Yang, Joshua P Schlegel, Takashi Hibiki, Mamoru IshiiAbstract:Abstract In order to develop detailed model of the annular two-phase flow the average liquid film thickness is an important parameter. It can significantly affect the occurrence of dryout and post-dryout phenomenon on heater surfaces. Most research on film thickness has been focused on pipe flows. Data in Rod Bundle geometry are very limited. However data in Rod Bundle geometry is much more valuable for reactor safety modeling and calculations. Because of this, an experiment to measure liquid film thickness has been performed for the air-water annular flow in a 8 × 8 BWR Rod Bundle. Film thickness data were obtained both on the Rods and on the channel wall. Data were recorded at four axial locations within the Rod Bundle. This included locations just before and just after a spacer grid. This allowed the spacer grids effect on the film thickness to be evaluated. The minimum film thickness was on the center Rod. Also, the presence of the spacer grid resulted in reduced film thickness.
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drift flux correlation for Rod Bundle geometries
International Journal of Heat and Fluid Flow, 2014Co-Authors: Collin Clark, Takashi Hibiki, Mamoru Ishii, Matthew Griffiths, Shaowen Chen, Tetsuhiro Ozaki, Ikuo Kinoshita, Yoshitaka YoshidaAbstract:Abstract A new drift-flux correlation has been developed to predict void fraction over a wide range of two-phase flow conditions in Rod Bundle geometries. An experimental database that represents low liquid flow and low pressure conditions in a scaled 8 × 8 Rod Bundle test facility is emphasized for this work. At these conditions, recirculating flow patterns may affect two-phase flow characteristics. Such effects may not be appropriately considered in earlier Rod Bundle correlations for drift velocity and distribution parameter. In the current approach, an existing drift-flux correlation that accounts for the effect of recirculating flow as two-phase flow regimes transition from bubbly to cap-bubbly flow is incorporated to determine distribution parameter. A performance analysis demonstrates that the proposed correlation improves upon existing correlations with an average relative error of ±4.5% when predicting the database utilized for in this work. The new correlation is also demonstrated to scale appropriately to prototypic plant conditions.
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experimental study of void fraction in an 8 8 Rod Bundle at low pressure and low liquid flow conditions
International Journal of Multiphase Flow, 2014Co-Authors: Collin Clark, Takashi Hibiki, Mamoru Ishii, Matthew Griffiths, Shaowen Chen, Ikuo Kinoshita, Yoshitaka YoshidaAbstract:Abstract An experiment has been performed to measure void fraction at stagnant to low liquid flow conditions in a Rod Bundle. An 8 × 8 Rod Bundle test facility scaled from a boiling water reactor design was utilized at atmospheric pressure with air and water as working fluids. Superficial liquid velocity ranged from 0 to 1.0 m/s and superficial gas velocity ranged from 0.03 to 10.0 m/s. Area-averaged measurements of superficial liquid velocity, superficial gas velocity, and void fraction are used to calculate distribution parameter from a kinematic constitutive equation of the drift-flux model. Results indicate a significant increase in distribution parameter when mixture volumetric flux is relatively low. This observation may be attributed to recirculating flow patterns. An investigation is conducted for existing Rod Bundle drift-flux correlations because they may not appropriately consider these mechanisms at low liquid flow and low pressure conditions. Results suggest that improvements may be made if the effects of recirculating flow are taken into consideration.
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experimental study of interfacial area transport in air water two phase flow in a scaled 8 8 bwr Rod Bundle
International Journal of Multiphase Flow, 2013Co-Authors: Xiaohong Yang, Joshua P Schlegel, Takashi Hibiki, Yang Liu, Sidharth Paranjape, Mamoru IshiiAbstract:Abstract In order to accurately predict nuclear reactor behavior, the ability to predict the transfer of mass, momentum and energy between the phases in two-phase flows, whether in the Reactor Pressure Vessel (RPV) or steam generator, is essential. A significant component of this prediction is the area available for transfer per unit volume, called the interfacial area concentration. Current thermal-hydraulic system analysis code predictions use empirical models to predict the interfacial area concentration; however accuracy and reliability can be improved through the use of an Interfacial Area Transport Equation (IATE). The IATE requires rigorously developed models for sources and sinks due to bubble interactions or phase change and an extensive database to validate those models. To provide this database, experiments using electrical conductivity probes to measure the interfacial area concentration at several axial positions have been performed in an 8 × 8 Rod Bundle which was carefully scaled from an actual BWR Rod Bundle.
