Heated Rod Bundle Research Articles

Numerical investigation of membrane oxygenation using sub-channel analysis

Purpose Better membrane oxygenators need to be developed to enable efficient gas exchange between venous blood and air. Design/methodology/approach Optimal design and analysis of such devices are achieved through mathematical modeling tools such as computational fluid dynamics (CFD). In this study, a control volume-based one-dimensional (1D) sub-channel analysis code is developed to analyze the gas exchange between the hollow fiber bundle and the venous blood. DIANA computer code, which is popular with the thermal hydraulic analysis of sub-channels in nuclear reactors, was suitably modified to solve the conservation equations for the blood oxygenators. The gas exchange between the tube-side fluid and the shell-side venous blood is modeled by solving mass, momentum and species conservation equations. Findings Simulations using sub-channel analysis are performed for the first time. As the DIANA-based approach is well known in rod bundle heat transfer, it is applied to membrane oxygenators. After detailed validations, the artificial membrane oxygenator is analyzed for different bundle sizes (L/W) and bundle porosity (epsilon) values, and oxygen saturation levels are predicted along the bundle. The present sub-channel analysis is found to be reasonably accurate and computationally efficient when compared to conventional CFD calculations. Research limitations/implications This approach is promising and has far-reaching ramifications to connect and extend a well-known rod bundle heat transfer algorithm to a membrane oxygenator community. As a variety of devices need to be analyzed, simplified approaches will be attractive. Although the 1D nature of the simulations facilitates handling complexity, it cannot easily compete with expensive and detailed CFD calculations. Practical implications This work has high practical value and impacts the design community directly. Detailed numerical simulations can be validated and benchmarked for future membrane oxygenator designs. Social implications Future membrane oxygenators can be designed and analyzed easily and efficiently. Originality/value The DIANA algorithm is popularly used in sub-channel analysis codes in rod bundle heat transfer. This efficient approach is being implemented into membrane oxygenator community for the first time.

Rod Bundle Heat Transfer Thermal-Hydraulic Program

The U.S. Nuclear Regulatory Commission has been conducting thermal-hydraulic research using the Rod Bundle Heat Transfer (RBHT) facility at the Pennsylvania State University since 2001. The facility has been used for five individual test programs: forced reflood, steam cooling, mixture level swell, dispersed droplet injection, and oscillatory reflood test series. While rod bundle thermal hydraulics has been extensively studied in the past, the RBHT data have provided new insights into rod bundle phenomena especially on the effects of spacer grids. This paper provides a summary of the RBHT test program and discusses some of the major findings from this research with the emphasis on reflood thermal hydraulics and the effect of spacer grids. Of particular interest are data that enable model and correlation development. Recent efforts have focused on the evaluation of RBHT data and development of improved models and correlations suitable for systems thermal-hydraulic codes such as TRACE and RELAP. Because of detailed instrumentation on and about spacer grids, RBHT data have enabled improved models for convective heat transfer enhancement and droplet breakup. New correlations for the inverted annular and the inverted slug film boiling regimes have also been developed as an initial step toward an improved model for dispersed droplet film boiling.

Investigation of the thermal-hydraulic non-equilibrium in a 7 × 7 rod bundle during reflood

The current research aims at investigating the thermal–hydraulic non-equilibrium encountered in the dispersed flow film boiling (DFFB) regime under reflood conditions. A series of constant reflood tests have been performed at the 7 × 7 Rod Bundle Heat Transfer (RBHT) test facility and the experimental results have been analyzed in detail. The data not only is useful for the development of theoretical models for two-phase interfacial mass and heat transfer, void fraction, pressure drop, and rod bundle cooling prediction, but also can be used as a benchmark data base for evaluating the performance of various thermal–hydraulic analysis codes. In this paper, a complete data reduction methodology is employed to reveal and interpret important two-phase flow phenomena during reflood transients. The results show that for the region far away from the quench front location, significant thermal non-equilibrium may exist during reflood and thus poor heat removal capability of the flow mixture is expected. On the other hand, for the region immediately downstream of the quench front, the two-phase flow is found to be close to the thermal equilibrium state, indicating that both the interfacial heat transfer and the heat transfer between the rod surface and fluid mixture are very efficient. In addition, based on the liquid droplet measurement and force balance calculations, it is found that the slip ratio between the two separate phases is a strong function of the droplet size and reflood time. With a larger droplet size and a shorter distance from the quench location, the slip ratio is larger. The liquid carryover fraction at the exit of the test section is also determined as a function of the reflood time. Results indicate that only a small portion of the injected coolant mass is stored within the bundle. A significant amount of liquid mass into the test section is either evaporated or entrained in the vapor flow and eventually left the bundle.

Characterization of salt deposit layer growth and prediction of cladding temperature of heated rod bundles under long-term seawater injection and pool boiling conditions

Seawater has been recognized as an alternative water source in severe accident countermeasures even before the TEPCO Fukushima Dai-ichi NPS accident. In this article, influences of deposit layers of salt precipitates on the early core degradation were investigated. A heated part of a fuel bundle was regarded as a major deposit site and numerical models for predicting growing geometry of deposit layers and resultant temperature responses of cladding were developed. These models have been validated based on precipitation tests employing 5 × 5 short length mockup bundles submerged in solutions of simple seawater and a mixture of seawater and boric acid. Growth and geometry of salt deposit layers were expressed as a product of the shape function and Matsuoka's kinetic law (Matsuoka, 1991) where the growth factor is expressed by the Arrhenius law. Model parameters of the kinetic law and the shape function were quantified based on digitized image data of deposit layers acquired in the abovementioned precipitation tests based on the X-Ray CT scan. Combining with the axial onedimensional two-phase flow model, the effective heat transfer coefficient of the cladding surface was evaluated taking into account growth of deposit layers and two-phase flow regimes. Finally, cladding temperature responses were evaluated based on the energy balance. Final geometry of salt deposit layers was digitized under a wide range of pool boiling conditions based on image data acquired in precipitation tests employing 5 × 5 short length mockup bundles. Based on these data, the shape function of each test case was reduced. It was confirmed that Matsuoka's kinetic law could predict a total mass increase of deposit layers with a good accuracy by specifying a constant in the Arrhenius law as nearly equivalent values for all test cases. From a comparison of measured and predicted surface temperatures at the exit of a heated length of the center rod, it can be interpreted that temperature increased gradually after the salt concentration reached the saturation point, then temperature began to increase steeply when a gap formed by neighboring rods was closed and isolated chimney flow paths surrounding the center heater rod were formed. Another important finding is that there was a certain margin of time until chimneys were formed after reaching the saturated salt concentration, which indicates feasibility of a practical fast running evaluation model based on decay heat, average salt concentration and boundary conditions that can be estimated to a certain degree of confidence level even under accident conditions.

Experimental investigation on the distribution of spray water in a spent fuel-assembly simulator

Abstract The spent fuel pool cooling system in a nuclear power plant, which is comprised mainly by the cooling pumps and heat exchangers, ensures the safety of the spent fuel assemblies and the integrity of the fuel rods during the period of storage. With the development of the passive cooling technique, a spray cooling system for the spent fuels based on the gravity was designed to further enhance the safety of the spent fuel pool in case of accident conditions. This paper presents an experimental investigation of the validity of the spray-cooling system using two types of tight rod bundles, namely a 5 × 5 heated rod bundle and a 17 × 17 isothermal rod bundle. Results shows that the rod bundle heated with a lower power can be effectively cooled only by air without any spray water. With the increase of the heated power, the rod surface temperature increases gradually and the spray cooling has to be implemented to maintain the wall temperature at a certain level. The effect of flow rate on wall temperature was investigated. For the isothermal rod bundle, main interests were focused on the distribution of the spray water after it flowed along the rods.

CFD simulation of heated tight-lattice rod bundles for an IAEA benchmark

The International Atomic Energy Agency (IAEA) organizes Coordinated Research Projects (CRP) to facilitate the development and validation of computer codes for design and safety analysis of nuclear power plants. Canadian Nuclear Laboratories (CNL) participated in such a CRP that included Computational Fluid Dynamics (CFD) benchmark exercises for heated rod bundles. This paper presents the CNL benchmark results for the 4 × 4 tight-lattice rod-bundle experiment (pitch-to-diameter ratio P/D = 1.08) performed at Korea Atomic Energy Research Institute (KAERI). The KAERI experiment comprised of upward flows of water through the rod bundle configuration that included twisted-vane spacers and eight support grids along the length of the heated rod bundle. The commercial CFD code STAR-CCM + v 9.02.007 was used to simulate the flow and heat transfer in the entire bundle geometry including the supporting grids and vane spacers. Turbulence was accounted for using a k-ω turbulence model and conjugated heat transfer was included in the analysis to account for the effect of heat conduction on the fuel-rod wall temperature. The CFD predictions were assessed against the experimental data for the circumferential variation of the rod wall temperature at four different axial positions relative to the twisted-vane spacers. It was observed that the CFD model tends to overpredict the wall temperature, and the extent of overprediction is larger at lower measurement positions than at higher measurement positions.

Heat transfer experiment in a partially (internally) blocked 19-rod bundle with wire spacers cooled by LBE

Internal blockages in fuel rod bundles can be rather small and may have a negligible global effect on the flow rate in a fuel assembly, and yet they can still produce significant local hot spots potentially leading to clad failure. Within the European project MAXSIMA the thermal-hydraulic effects of internal flow blockages in a heated rod bundle cooled by lead-bismuth eutectic (LBE) are investigated in prototypical experiments. This work builds up on previous experiments in an unblocked 19-rod bundle with wire spacers within the project SEARCH.In all cases, operating conditions representative of the scenarios expected in the MYRRHA reactor, developed by SCK• CEN in Belgium, are imposed. Similarly the characteristics of the blockage scenarios are selected according to the worst-case postulated events, in a conservative approach. Only solid blockages are installed, while a certain porosity can be expected if they are formed by accumulation of particles. Moreover, they are constructed using a filling material of low thermal conductivity, instead of being metallic. All blockage elements have the same length (54.6 mm, i.e. one sixth of the wire pitch) but different position and size. A first scenario includes two small blockages covering one central and edge sub-channel each. In the second scenario, a large blockage covering six sub-channels surrounding the central rod is investigated.The experimental results are presented following two approaches. First, selected reference cases are described in detail, allowing for a direct comparison with numerical simulations. Second, the results for all cases are correlated in dimensionless terms. An empirical model named Karlsruhe Blockage Equation (KBE) is proposed for representing the heat transfer resistance at the hottest spots in the blocked region for each element. Extrapolating these results to full-power conditions, the temperature increase produced by solid blockages with low thermal conductivity can be acceptable if they are small and excessive if they are larger. Further results include the wall and fluid temperature in the wake region and differential pressure in blocked and unblocked sections of the bundle.

Experimental study of droplet sizes across a spacer grid location under various reflood conditions

Droplet size and distribution across a spacer grid in a heated rod bundle during reflood stage of a postulated design basis accident loss of coolant accident (LOCA) are studied experimentally using the Rod Bundle Heat Transfer (RBHT) test facility. Effects of spacer grid conditions, quench front location during reflood, and inlet water subcooling on droplet field are investigated. Experimental results show that the droplet sizes decrease when they pass through a dry spacer grid due to cutting and shattering effects at the lower edges of the spacer grid. On the other hand, for a wet grid, the droplet sizes downstream of the grid could be larger than those at the upstream locations, as the droplet generation mechanism, for a wet grid, is due to the formation of liquid ligaments at the trailing edge of the spacer grid and the subsequent aerodynamic breakup of these ligaments into relatively larger droplets induced by instabilities. It is found that, as the quench front propagates upward, sizes of incoming droplets increase correspondingly as a result of the droplet-vapor thermal-hydraulic interactions along the flow channels. The experimental results also indicate that larger droplet diameters occur when the inlet water subcooling is higher. Results of the present study can be utilized to develop models for droplet field behavior under accident scenarios. These models can be incorporated into nuclear reactor thermal-hydraulic safety analysis codes such as COBRA-TF and TRACE.

Uncertainty analysis on droplet size measurement in dispersed flow film boiling regime during reflood using image processing technique

In the dispersed flow film boiling (DFFB) regime during the reflood stage of a postulated loss of coolant accident (LOCA), liquid droplets are entrained by a continuous vapor phase moving upward within the rod bundle. The liquid droplets may interact strongly with the vapor flow and the resulting thermal–hydraulic behavior of the two-phase mixture could significantly affect the evolution of peak cladding temperature (PCT). Therefore, a clear and comprehensive understanding of liquid droplet behavior is crucial for reactor safety analysis. As an attempt to visualize and quantify the droplet behavior during reflood, an advanced image processing technique is used in the present study to capture the distributions of droplet size and velocity at the RBHT test facility. In order to ensure that the experimental droplet data obtained is reliable as well as to better utilize the data for subsequent thermal–hydraulic analyses, the image processing based measurement uncertainties have to be investigated and analyzed. In this paper, various statistical methods are explored to determine the minimum droplet sample size required as a function of standard deviation at a given confidence level. Accordingly, an appropriate droplet count lower limit is selected. Based on this lower limit, the 95% confidence intervals are determined for two typical reflood tests. In addition, the repeatability in droplet measurement of Rod Bundle Heat Transfer (RBHT) reflood tests is carefully examined by comparing data sets from two identical tests. Results show that a lower required level of precision and a smaller scattering in the measured droplet data generally require a smaller sample size. For typical RBHT reflood tests, it is found that the droplet size measured downstream of a spacer grid near the peak power location always has better accuracy compared with that at the upstream location. For the tests investigated, the maximum relative error in the liquid droplet Sauter mean diameter is found to be generally smaller than 0.15 (15%) when the droplet count lower limit is 30. Comparison of the results from two identical tests indicates that the RBHT reflood test results are highly repeatable.

Comparative study of turbulence model predictions of upward supercritical fluid flow in vertical rod bundle subchannels

Computations of three-dimensional turbulent vertical up-flow of supercritical fluid in the subchannel of a heated rod bundle were made using the Computational Fluid Dynamics codes ANSYS CFX and ANSYS Fluent. Results for a total of six cases from three different sets of experiments are presented. For all six cases, steady-state predictions of fluid velocity, pressure, and temperature were made using six versions of the two-equation Reynolds-averaged Navier-Stokes turbulence models with accompanying wall treatments. A conjugate heat transfer model was used that also predicted the temperature distribution in an adjacent solid region representing a heater. In the model of one experiment, the solid region also included cladding and insulation.The k-epsilon turbulence model, implemented using CFX and Scalable Wall Functions, provided the numerical results that have the smallest overall deviation from experimental results for three of the six cases, and predicted the experimental data of the remaining four cases reasonably well, unlike other turbulence models that sometimes severely over-predict the experimental data for wall surface temperature.

Thermal-hydraulic comparisons of 19-pin rod bundles with four circular and trapezoid shaped wire wraps

Coolant in wire-wrapped fuel assembly flows not only in axial direction but also in a transverse direction. This special flow is critical for momentum exchange and heat transfer in fuel rod bundle. To obtain a better performance wire-wrapped fuel assembly for advanced reactor cores, in this paper, the characteristics of turbulent flow and heat transfer inside 19-pin heated rod bundles where each pin is wrapped with four circular shaped and four trapezoid shaped wires are numerically studied, respectively. Considering the geometric complexity, the advanced hybrid-grid technique that combines the structured hexahedron grids and unstructured pentahedral grids is adopted to divide the modeling assembly. The validity of both RSM and SST k-ω model is assessed by comparing calculated pressure loss with the available experimental data to determine a proper turbulence closure. Special attention has been paid to understand the effect of secondary flows created by the wire wraps on the reduction of coolant temperature variation. Additionally, in order to assess the effectiveness of the different shaped wires, comparisons on both global heat transfer and associated penalty in pressure loss are evaluated. Results indicate that the frictional pressure loss of the rod bundle with circular shaped wires is smaller than that with trapezoid shaped wires, while the global heat transfer is equivalent. However, the trapezoid shaped wires give a lower maximum temperature as well as a uniform temperature distribution within the subassembly in comparison with circular shaped geometry.

Heat transfer correlation for film boiling in vertical upward flow

There is no heat transfer correlation available in the Inverted Slug Film Boiling (ISFB) regime. In this study, a new correlation for film boiling in a vertical channel is proposed using a semi-empirical model. This correlation is applicable in the Inverted Annular Film Boiling (IAFB) heat transfer regime as well as ISFB heat transfer regime. A semi-empirical model is developed to establish the functional dependence of the Nusselt number on various controlling parameters. Test data obtained from the transient reflood tests carried out in a full length 7×7 vertical Rod Bundle Heat Transfer (RBHT) facility are used to develop the correlation. The predicted Nusselt number agrees within 15% with the RBHT data set and other data set available in the literature. This heat transfer correlation is applicable up to a void fraction of 90%, i.e., both in the IAFB and ISFB regimes. The proposed correlation is also applicable to a tube geometry using a modified coefficient in the correlation with an error less than 10%.

Experimental study of laminar mixed convection in a rod bundle with mixing vane spacer grids

Heat transfer by mixed convection in a rod bundle occurs when convection is affected by both the buoyancy and inertial forces. Mixed convection can be assumed when the Richardson number (Ri=Gr/Re2) is on the order of unity, indicating that both forced and natural convection are important contributors to heat transfer. In the present study, data obtained from the Rod Bundle Heat Transfer (RBHT) facility was used to determine the heat transfer coefficient in the mixed convection regime, which was found to be significantly larger than those expected assuming purely forced convection based on the inlet flow rate. The inlet Reynolds (Re) number for the tests ranged from 500 to 1300, while the Grashof (Gr) number varied from 1.5×105 to 3.8×106 yielding 0.25<Ri<4.3. Using results from RBHT test along with the correlation from the FLECHT-SEASET test program for laminar forced convection, a new correlation ​is proposed for mixed convection in a rod bundle. The new correlation accounts for the enhancement of heat transfer relative to laminar forced convection.

Subchannel analysis and correlation of the Rod Bundle Heat Transfer (RBHT) steam cooling experimental data

Abstract A subchannel analysis of the steam cooling data obtained in the Rod Bundle Heat Transfer (RBHT) test facility has been performed in this study to capture the effect of spacer grids on heat transfer. The RBHT test facility has a 7 × 7 rod bundle with heater rods and with seven spacer grids equally spaced along the length of the rods. A method based on the concept of momentum and heat transport analogy has been developed for calculating the subchannel bulk mean temperature from the measured steam temperatures. Over the range of inlet Reynolds number, the local Nusselt number was found to exhibit a minimum value between the upstream and downstream spacer grids. The presence of a spacer grid not only affects the local Nusselt number downstream of the grid but also affects the local Nusselt number upstream of the next grid. A new correlation capturing the effect of Reynolds number on the local flow restructuring downstream as well as upstream of the spacer grids was proposed for the minimum Nusselt number. In addition, a new enhancement factor accounting for the effects of the upstream as well as downstream spacer grids was developed from the RBHT data. The new enhancement factor was found to compare well with the data from the ACHILLLES test facility.

Experimental Studies of Spacer Grid Thermal Hydraulics in the Dispersed Flow Film Boiling Regime

Spacer grids have been found to enhance downstream convective heat transfer and to strongly influence droplet size distributions through early spacer grid rewet and droplet breakup. Existing models for enhancement of heat transfer and droplet breakup, however, do not appear to accurately account for these interactions between the coolant and the spacer grid. Data from two series of rod bundle heat transfer tests, low injection rate forced reflood tests, and droplet injection tests are presented in this paper to describe the effects of the spacer grids during dispersed flow film boiling. Heat transfer downstream of the spacer grids is clearly enhanced by the presence of the droplets, while the downstream droplet size was found to depend on the condition of the spacer grid: dry or wetted. Results of this study demonstrate the need to adequately account for the separate modes of dry and wet spacer grid heat transfer enhancement in predicting the thermal-hydraulic behavior during reflood transients.

Multi-dimensional void fraction measurement of transient boiling two-phase flow in a heated rod bundle

Multi-dimensional void fraction measurement of transient boiling two-phase flow in a heated rod bundle

Numerical simulation of flow through an eccentric annulus with heat transfer

Purpose – The purpose of this paper is to investigate the effect of narrow gap on the fluid flow and heat transfer through an eccentric annular region is numerically. Flow through an eccentric annular geometry is a model problem of practical interest. Design/methodology/approach – The approach involves standard finite volume-based SIMPLE scheme. The numerical simulations cover the practically relevant Reynolds number range of 104-106. Findings – In the narrow gap region, temperature shoot up was observed due to flow maldistribution with an attendant reduction in the heat removal from the wall surfaces. CFD analysis is presented with the aid of, streamlines, isotherms, axial velocity contours, etc. The engineering parameters of interest such as, Nusselt number, wall shear stress, etc., is presented to study the effect of eccentricity and radius ratio. Research limitations/implications – The present investigation is a simplified model for the rod bundle heat transfer studies. However, the detailed study of sectorial mass flux distribution is a useful precursor to the thermal hydraulics of rod bundles. Practical implications – For nuclear reactor fuel rods, the effect of eccentricity is going to be detrimental and might lead to the condition of critical heat flux. A thorough sub-channel analysis is very useful. Social implications – Nuclear safety standards require answers to a wide a range of what-if type hypothetical scenarios to enable preparedness. This study is a highly simplified model and a first step in that direction. Originality/value – The narrow gap region has been systematically investigated for the first time. A detailed sectorial analysis reveals that, flow maldistribution and the attendant temperature shoot up in the narrow gap region is detrimental to the safe operation.

Numerical analysis of RBHT reflood experiments using MARS 1D and 3D modules

The Rod Bundle Heat Transfer (RBHT) program was performed experimentally to analyze the reflood heat transfer phenomena under the conditions of reflood phase following a hypothesized loss of coolant accident (LOCA) by the team of Penn State University. In order to verify the experimental data using a numerical analysis, the Multi-dimensional Analysis of Reactor Safety (MARS) assessment of the RBHT experimental data was carried out for the flooding rates of 0.0254 and 0.1524 m/sec with the upper plenum pressure of 276 kPa. The RBHT experimental data of Tests 1285 and 1383 were compared with the calculation results of the MARS 1D and 3D modules. The MARS code shows a good agreement in the general trend of the peak cladding temperatures although there are limitations in predicting accurate quenching time for both modules. However, in comparison to the MARS 1D module simulation, the MARS 3D module shows the improved calculation capability in that the code can capture local enhanced heat transfer with implication of spacer grids. Moreover, the temperature profiles simulated by the 3D module show the accurate prediction at which the local peak temperatures occur. For more enhanced simulations, local flow parameters such as cross flow and vortex flow need to be analyzed for a more accurate prediction of quenching behavior.

Development of Measurement Method of Void Fraction Distribution for Boiling Water Flow in Heated Rod Bundle

SubChannel Void Sensor (SCVS) consisting of electrodes with 6-wire by 6-wire and 5-rod by 5-rod electrodes has been developed to acquire boiling two-phase flow dynamics in 5×5 rod bundle geometry. The SCVS can acquire a local void fraction in 32 points (=6×6-4) of central subchannel regions and 100 points (=4×25) of near rod surface regions. The temporal resolution is up to 5000 frames (cross sections) per second. The devised sensors are installed every 500mm in the heated rod bundle and exhibits the quasi three-dimensional boiling two-phase flow structures, i.e. void fraction distribution.

Effect of Flow Blockage on the Coolability during Reflood in a 2×2 Rod Bundle

During the reflood phase of a large-break loss-of-coolant accident (LBLOCA) in a pressurized-water reactor (PWR), the fuel rods can be ballooned or rearranged owing to an increase in the temperature and internal pressure of the fuel rods. In this study, an experimental study was performed to understand the thermal behavior and effect of the ballooned region on the coolability using a 2 × 2 rod bundle test facility. The electrically heated rod bundle was used and the ballooning shape of the rods was simulated by superimposing hollow sleeves, which have a 90% blockage ratio. Forced reflood tests were performed to examine the transient two-phase heat transfer behavior for different reflood rates and rod powers. The droplet behaviors were also investigated by measuring the velocity and size of droplets near the blockage region. The results showed that the heat transfer was enhanced in the downstream of the blockage region, owing to the reduced flow area of the subchannel, intensification of turbulence, and deposition of the droplet.