Rod Bundle Configuration Research Articles

Modeling of void fraction covariance and relative velocity covariance for upward boiling flows in subchannels of a vertical rod bundle

Accurate simulation of boiling two-phase flows in a rod bundle is indispensable for the robust, economical, and safe design of various heat transfer systems using the rod bundle configuration. Subchannel analysis codes are used for this purpose. Interfacial drag force modeling significantly affects the prediction accuracy of the void fraction. The void fraction and relative velocity covariances constitute the interfacial drag force. However, the covariances are currently not considered in existing subchannel codes due to a lack of reliable constitutive equations to calculate the void fraction and relative velocity covariances. This study aims to model the subchannel-average void fraction and relative velocity covariances for subcooled and saturated boiling flows in three types of subchannels in a rod bundle. The considered subchannels are interior, edge, and corner subchannels. The subchannel-average void fraction and relative velocity covariances for saturated boiling flow are modeled by the data obtained from local void fraction data collected for saturated boiling flow in an 8 × 8 rod bundle under pressures from 1.0 to 8.6 MPa. The subchannel-average void fraction and relative velocity covariances for subcooled boiling flow are modeled based on the bubble-layer thickness model. The modeled subchannel-average void fraction and relative velocity covariances are well validated with the experimental data. The modeled subchannel-average void fraction and relative velocity covariances are expected to be implemented in subchannel analysis codes to improve the void fraction prediction accuracy in each subchannel type.

Towards the accurate prediction of axial flow and heat transfer in a tightly spaced bare rod bundle configuration

The understanding of flow behaviour and temperature distribution in rod bundles during operation is of great importance for the safety and economical design of existing nuclear reactors and as well as new nuclear technologies. In order to study the thermal-hydraulics phenomena in the fuel assembly geometries, researchers all over the world are increasingly using Computational Fluid Dynamics (CFD) as a reliable research tool. Nevertheless, the prediction capabilities of the most commonly used Reynolds Average Navier-Stokes (RANS) models must be assessed. In this regard, in the present research work, an extensive effort has been put forward and is described in three main steps. As a first step, a wide range of RANS and unsteady RANS computations have been performed to design a numerical experiment for a closely-spaced bare rod bundle in order to perform a direct numerical simulation (DNS), which will serve as a reference for the validation purpose. As a second step, the DNS of this bare rod bundle has been performed for three different Prandtl number (Pr) fluids, i.e. Pr = 2, 1 and 0.025. This DNS has been performed using a higher-order spectral element code and consists of 660 million grid points. Accordingly, an extensive database has been generated for validation purposes. Finally, this database is used to assess the prediction capabilities of different linear and non-linear RANS models. This article will provide a good overview of all three steps and will also highlight the main lessons learned from this research.

Experimental Investigation of Rod Pitch Effect on Void Fraction in 5 × 5 Rod‐bundle under Elevated Pressure Conditions

In the present study, void fraction distributions in 5 × 5 rod-bundle test sections were investigated. Two different rod-bundle test sections with different rod pitch lengths were constructed. Experiments were carried out using unheated rod-bundle test sections equipped with 10 mm rod diameter with rod-pitch lengths of 13 and 12 mm, respectively. The HUSTLE (Hitachi Utility Steam Test Leading Facility) facility supplied steam and water with an inlet mass flux (G) ranging from 75 to 500 kg/m2/s at 5.0 to 7.5 MPa system pressure. These inlet and boundary conditions simulate the prototypic BWR at natural circulation conditions. For the measurement of void fraction, the X-ray CT system was newly developed for the three-dimensional time-averaged void fraction measurement, and 3D void fraction distributions were successfully obtained for various pressure conditions. The volume-averaged void fraction was also measured using differential pressure transmitters. Based on the newly developed void fraction database, the assessment was carried out using state-of-the-art drift-flux correlations applicable for rod-bundle. Existing correlations tend to show good void fraction prediction at high mass flux conditions for both geometries, poorer performance was shown for the narrow rod test sections with low mass flux conditions. Furthermore, the effect of the rod-pitch lengths on void fraction distribution in rod-bundle flow field was assessed. It was observed that narrower rod bundle configuration tends to have higher void fraction at given mass flux. Narrowed pitch-length enhances wall-friction against two-phase flow, which likely suppress the central phase fraction peak tendency. This promotes lower distribution parameter values, results in increased area-averaged void fraction for narrower pitch test section. However, the present study revealed that no existing correlations could clearly capture the void fraction trend for the narrow rod test section.

A collaborative effort towards the accurate prediction of turbulent flow and heat transfer in low-Prandtl number fluids

This article reports the experimental and DNS database that has been generated, within the framework of the EU SESAME and MYRTE projects, for various low-Prandtl flow configurations in different flow regimes. This includes three experiments: confined and unconfined backward facing steps with low-Prandtl fluids, and a forced convection planar jet case with two different Prandtl fluids. In terms of numerical data, seven different flow configurations are considered: a wall-bounded mixed convection flow at low-Prandtl number with varying Richardson number (Ri) values; a wall-bounded mixed and forced convection flow in a bare rod bundle configuration; a forced convection confined backward facing step (BFS) with conjugate heat transfer; a forced convection impinging jet for three different Prandtl fluids corresponding to two different Reynolds numbers of the fully developed planar turbulent jet; a mixed-convection cold-hot–cold triple jet configuration corresponding to Ri = 0.25; an unconfined free shear layer for three different Prandtl fluids; and a forced convection infinite wire-wrapped fuel assembly. This wide range of reference data is used to evaluate, validate and/or further develop different turbulent heat flux modelling approaches, namely simple gradient diffusion hypothesis (SGDH) based on constant and variable turbulent Prandtl number; explicit and implicit algebraic heat flux models; and a second order turbulent heat flux model. Lastly, this article will highlight the current challenges and perspectives of the available turbulence models, in different codes, for the accurate prediction of flow and heat transfer in low-Prandtl fluids.

Dispersion of Surrogate LWR Fuel Experiments Under LOCA Conditions

In this study a series of experiments were performed subjecting surrogate nuclear fuel rods to high-pressure transients to induce fuel dispersion representative of the expected conditions of a fuel rod during a hypothetical loss-of-coolant accident. Experiments were conducted on like-for-like pressurized water reactor geometries in both a single-rod and rod-bundle configuration. In the rod-bundle configuration, a matched index of refraction techniques was employed to provide optical access to the bundle internals and to view the surrogate fuel dispersion event. Both configurations used small lead pellets as a surrogate fuel and were observed with a high-speed camera to capture the transient on a resolved timescale. For the single-rod experiments, the test rod was subjected to pressure transients at 4.0, 8.0, and 12.0 MPa multiple times, and for the rod-bundle experiments, the rod was subjected to 8.0 MPa transients in order to compare mechanical behavior against the single-rod test at 8.0 MPa. For both configurations, the results showed highly variable behavior in both the quantity of fuel dispersed and the mean displacement relative to the burst rod origin, likely due to statistical variations in the internal fuel stack orientation. Measurements of the rod plenum internal pressure showed no discernible difference in depressurization rates at a given pressure, indicating the likelihood that the mass flow rate is limited by the valve orifice in the current experimental configuration. The bundle tests also showed that a 5 × 5 array appears to be too small to capture the full spatial distribution of dispersed fuel, thus future tests will employ a larger bundle size and particle collection technique.

Reference numerical database for turbulent flow and heat transfer in liquid metals

Turbulent heat transfer is a complex phenomenon that has challenged turbulence modellers over various decades. In this regard, in the recent past, several attempts have been made for the assessment and further development/calibration of the available turbulent heat flux modelling approaches. One of the main hampering factors with respect to the further assessment of these modelling approaches is the lack of reference data. In the framework of the EU SESAME and MYRTE projects, an extensive effort has been put forward to generate a wide range of reference data, both experimental and numerical, to fill this gap. In that context, this article reports the numerical database that has been generated within these projects for various liquid metal flow configurations in different flow regimes. These high fidelity numerical data include seven different flow configurations: a wall-bounded mixed convection flow at low Prandtl number with varying Richardson number (Ri) values; a wall-bounded mixed and forced convection flow in a bare rod bundle configuration; a forced convection confined backward facing step (BFS) with conjugate heat transfer; a forced convection impinging jet for three different Prandtl fluids corresponding to two different Reynolds numbers of the fully developed planar turbulent jets; a mixed-convection cold-hot-cold triple jet configuration corresponding to Ri=0.25; an unconfined free shear layer for three different Prandtl fluids; and a forced convection infinite wire-wrapped fuel assembly. These high-fidelity numerical databases will serve the further development of turbulent heat transfer models by providing unique, new and detailed data for the thermal-hydraulic behaviour of liquid metals in various flow configurations.

Towards the accurate prediction of the turbulent flow and heat transfer in low-Prandtl fluids

In the numerical simulation of turbulent heat transfer, a reliable representation of the turbulent flow field is a prerequisite in order to obtain an accurate prediction of the thermal field. In the framework of RANS modelling approach, classical two-equation eddy-viscosity models present intrinsic limitations for applications characterised by complex flow fields where significant turbulence anisotropy can be expected. In this respect, differential Reynolds stress models represent the natural choice for dealing with such highly anisotropic flows. On the other hand, the modelling of the turbulent heat flux term almost universally relies on the so-called Reynolds analogy, due to its simplicity and robustness. Nevertheless, this approach presents several well-known limitations, especially when dealing with low-Prandtl fluids. In this context, algebraic heat flux models are regarded as a promising approach to overcome the shortcomings of the Reynolds analogy. In particular, an implicit algebraic heat flux closure, called as AHFM-NRG, has been proposed for applications involving low-Prandtl fluids. In its original formulation, the algebraic turbulent heat flux closure was coupled with a low-Reynolds linear k-ε model for the closure of the Reynolds stress tensor. In the present work, the applicability of the AHFM-NRG model is extended to an advanced low-Reynolds differential Reynolds stress model. This new version of the model is validated against different planar channel flows at unity and low-Prandtl number; successively, the model is applied to two relatively complex flows, i.e. a planar impinging jet and a bare rod bundle configuration. It is shown that the coupling of the AHFM-NRG with an advanced closure for the Reynolds stresses gives encouraging results for these cases, where an accurate prediction of the anisotropy in the flow field results in a noticeable improvement in the prediction of the turbulent heat transfer.

Numerical prediction of flow and heat transfer in a loosely spaced bare rod bundle

A wall-resolved Large Eddy Simulation (LES) is performed for a loosely packed bare rod bundle with Reynolds number of 54,650. The simulation includes heat transfer in a liquid metal with the Prandtl number of 0.016. The rod bundle configuration corresponds to the ALFRED lead-cooled reactor design. The ALFRED design employs nuclear fuel assemblies with triangular arrangement of the fuel rods. In the present study, two important thermal-hydraulics aspects of the ALFRED rod bundle configuration have been analysed. Firstly, the proposed design should not lead to any undesirable vibration issues and, secondly, the heat transport should assure sufficient cooling with no hot spots. The obtained flow and thermal fields are analysed and compared with available data. A detailed investigation of the flow field has revealed the existence of a relatively weak gap vortex street in the loosely packed rod bundle. Hence, this considered configuration is unlikely to cause any vibration issues. In addition, the observed thermal field has shown a specific logarithmic trend, representing the heat transfer in wall bounded flow configurations for a wide range of Péclet numbers. In general, no hot-spots are observed for the considered design parameters.

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.

Space and energy-based turbulent scale-resolving simulations of flow in a 5 × 5 nuclear reactor core fuel assembly with a spacer grid

The partially averaged Navier–Stokes turbulence model, which is a bridging model between the Reynolds averaged Navier–Stokes model and direct numerical simulation, is capable of directly resolving the turbulent scales of interest by filtering the Navier–Stokes equation using turbulent kinetic energy and dissipation based filters. In this study, comparison is made between the partially averaged Navier–Stokes and large eddy simulation models, with emphasis on the mean flow characteristics and spatio-temporal turbulent flow structures. The results show that the partially averaged Navier–Stokes modeling produces results comparable to those of large eddy simulation when the appropriate cut-off energy filter is chosen. The test case involves simulating flows in a 5 × 5 fuel rod bundle configuration with a spacer grid at a Reynolds number of 14,000. The simulation results are compared with particle image velocimetry data obtained from the scientific literature.

Accurate prediction of the wall shear stress in rod bundles with the spectral element method at high Reynolds numbers

Resolving flow near walls is critical to reproducing the high rates of shear that generate turbulence in high Reynolds number, wall-bounded flows. In the present study, we examine the resolution requirements for correctly reproducing mean flow quantities and wall shear stress distribution in a large eddy simulation using the spectral element method. In this method, derivatives are only guaranteed in a weak sense, and the same is true of quantities composed of derivatives, such as the wall shear stress. We are interested in what is required to resolve the wall shear stress in problems that lack homogeneity in at least one direction. The problem of interest is that of parallel flow through a rod bundle configuration. Several meshes for this problem are systematically compared. In addition, we conduct a study of channel flow in order to examine the issues in a canonical flow that contains spanwise homogeneity missing in rod bundle flow. In the case of channel flow, we compare several meshes and subgrid scale models. We find that typical measures of accuracy, such as the law of the wall, are not sufficient for determining the resolution of quantities that vary along the wall. Spanwise variation of wall shear stress in underresolved flows is characterized by spikes—physical points without well-defined derivatives of the velocity—found at element boundaries. These spikes are not particular to any subgrid scale model and are the unavoidable consequence of underresolution. Accurately reproducing the wall shear stress distribution, while minimizing the computational costs, requires increasing the number of elements along the wall (local h-refinement) and using very high order (N=19) basis functions (p-refinement). We suggest that while these requirements are not easily generalized to grid spacing guidelines, one can apply a general process: construct a mesh that progressively increases elements along any walls, and increase the order of basis functions until the distribution of wall shear stress or any other quantity of interest is smooth.

Numerical Analysis of Flow and Heat transfer in Sub-channel of Supercritical Water Reactor

Investigation on the thermal-hydraulic behaviour in the High performance Light Water Reactor (HPLWR) fuel assembly has obtained a significant attention in GENERATION-IV. In this paper, CFD analysis is carried out to study the heat transfer behaviour of supercritical water in sub-channels of squared annular assembly. Effect of various parameters, e.g. thermal boundary conditions and pitch-to-diameter ratio on the thermal–hydraulic behaviour is investigated. Boundary conditions, i.e., constant heat flux at the outer surface of cladding are applied. The effect of pitch-to-diameter ratio (P/D) on the distributions of surface temperature and heat transfer coefficient (HTC) was analyzed .For typical (gap between two rod) channel analysis, it is found that HTC increases with p/d first and then decreases significantly when p/d is <1.2. If P/D is larger than 1.2 heat transfer deterioration (HTD) occurs as main stream enthalpy is quite small. One/eight channel of 40 pin fuel assembly show that the structure of the secondary flow mainly depends on the rod bundle configuration as well as the pitch-to-diameter ratio, whereas, the amplitude of the secondary flow is affected by the thermal boundary conditions, as well. Furthermore, the HTD under wire-wrapped spacer were evaluated.

CFD analysis of thermal–hydraulic behavior in SCWR typical flow channels

Investigations on thermal–hydraulic behavior in SCWR fuel assembly have obtained a significant attention in the international SCWR community. However, there is still a lack of understanding and ability to predict the heat transfer behavior of supercritical water. In this paper, CFD analysis is carried out to study the flow and heat transfer behavior of supercritical water in sub-channels of both square and triangular rod bundles. Effect of various parameters, e.g. thermal boundary conditions and pitch-to-diameter ratio on the thermal–hydraulic behavior is investigated. Two boundary conditions, i.e., constant heat flux at the outer surface of cladding and constant heat density in the fuel pin are applied. The results show that the structure of the secondary flow mainly depends on the rod bundle configuration as well as the pitch-to-diameter ratio, whereas, the amplitude of the secondary flow is affected by the thermal boundary conditions, as well. The secondary flow is much stronger in a square lattice than that in a triangular lattice. The turbulence behavior is similar in both square and triangular lattices. The dependence of the amplitude of the turbulent velocity fluctuation across the gap on Reynolds number becomes prominent in both lattices as the pitch-to-diameter ratio increases. The effect of thermal boundary conditions on turbulent velocity fluctuation is negligibly small. For both lattices with small pitch-to-diameter ratios ( P/ D < 1.3), the mixing coefficient is about 0.022. Both secondary flow and turbulent mixing show unusual behavior in the vicinity of the pseudo-critical point. Further investigation is needed. A strong circumferential non-uniformity of wall temperature and heat transfer is observed in tight lattices at constant heat flux boundary conditions, especially in square lattices. In the case with constant heat density of fuel pin, the circumferential conductive heat transfer significantly reduces the non-uniformity of circumferential distribution of wall temperature and heat transfer, which is favorable for the design of SCWR fuel assemblies.

Applicability of Core Thermal-Hydraulic Models in REFLA Code to 17*17 Type Fuel Assembly of PWR.

Abstract The applicability of core thermal-hydraulic models in REFLA code was evaluated to a 17×17 type fuel assembly of PWR, the models which have been developed for the reflood phase of LOCA of PWR with a 15×15 type fuel assembly. The two assemblies are different each other in respect of (1) the rod bundle configuration (fuel rod diameter and pitch) and (2) the grid spacer type (existence of mixing vane and interval between spacers). The effects of (1) and (2) were investigated experimentally and the applicability was assessed with the data. The film boiling heat transfer and void fraction models in REFLA code could apply to the 17×17 type fuel assembly within the error band of each model (±30%). The difference of grid spacer type did not affect the turnaround temperature which is the maximum clad temperature at each elevation but quench velocity. The quench velocity became lower in the case of grid spacer type of 17×17 type fuel assembly and the increase of heat transfer coefficient was delayed. From the analyses by the REFLA code with adjusted quench velocity correlation, the delay of increase of heat transfer coefficient was found to give no significant effects on the peak clad temperature even under the condition calculated by evaluation model code of reactor safety analysis.