Tight Lattice Rod Bundle Research Articles

Two-group drift-flux model in tight lattice subchannel

Tight lattice rod bundles can increase equipment compactness and improve heat exchange efficiency, and they are widely adopted in heat exchange engineering. Understanding the flow characteristics of gas-liquid two-phase flow in tight lattice subchannels is essential for developing heat exchangers and fuel assemblies with these tight bundles. The drift-flux model (DFM) is a crucial two-phase flow model extensively used in subchannel analysis codes. Investigating the DFM in tight lattice subchannels benefits the advancement of these codes. For two-phase flow interfacial structures, significant two-group characteristics exist, with bubbles divided into small (group-one) and large (group-two) bubbles. The substantial differences in flow characteristics between these two groups provide a solid foundation for developing a two-group DFM. This study examined the two-group characteristics of two-phase flow in a tight lattice interior subchannel and developed a corresponding two-group DFM. The two-group correlations for the distribution parameters and drift velocities at the tight lattice interior subchannel level were proposed and verified using experimental data. The accuracy of the developed two-group DFM in predicting group-wise void fraction and gas velocity was also verified. The standard relative deviation between the model predictions and experimental data was 8.11 % and 13.1 % for group-one and group-two void fractions and 8.33 % and 11.7 % for group-one and group-two gas velocities, respectively. This indicates satisfactory accuracy for the newly developed two-group DFM in a tight lattice interior subchannel.

Two-phase flow evolution and interfacial area transport downstream of the mixing-vane spacer grid in rod bundle channels

This study investigated the axial two-phase flow evolution and interfacial area transport characteristics downstream of the mixing vane spacer grid (MVSG) in a tight lattice rod bundle channel with a subchannel hydraulic diameter of Dh = 17.27 mm. The experiment was conducted under the air-water two-phase flow at room temperature and atmospheric pressure. The phase distributions of 6 positions downstream of MVSG (Z/Dh = 9.15, 14.94, 20.73, 26.52, 32.31, and 61.26) were measured utilizing a self-developed double-layer wire-mesh sensor (WMS). 42 cases were obtained, involving the bubbly flow, cap-bubbly flow, and slug flow. Void fraction, bubble velocity, interfacial area concentration (IAC), and bubble size distribution (BSD) databases were built. Under the group-1 (G-1) flow, due to the swirling flow generated by MVSG, the G-1 bubbles were drawn from the bundle surface and the gap region into the subchannel center to form a spindle core-peak distribution. BSD results demonstrate that the swirling flow promotes bubbles’ coalescence. The one-dimensional (1-D) void fraction and IAC downstream of the MVSG decrease first and then recover, which contrasts with the non-mixing vane spacer grid (N-MVSG) effect. It was attributed to the increment of the bubbles’ velocity after MVSG due to the bubbles’ migration to the channel center and coalescence. Under the group-2 (G-2) flow, MVSG intensifies the core peak distribution of void fraction, and the void fraction profile is also spindle-shaped. The obvious bubbles’ break-up occurs at Z/Dh = 14.94 attributed to the swirling decay. The 1-D void fraction and IAC transport indicate, that with the liquid-velocity increment, the MVSG effect is different from the N-MVSG effect, which is mainly achieved by affecting the G-1 bubbles’ dynamic behaviors. The model evaluation utilizing the interfacial area transport equation (IATE) closing bubble interaction source/sink terms indicates that the advection effect dominates the IAC evolution and the contribution of the pressure effect is weak. The remarkable bubbles’ coalescence occurs under the high liquid velocity. In future studies, the additional bubble break-up and coalescence source/sink term model considering the MVSG effect should be developed based on these cases to improve the IATE prediction ability.

The influence of turbulators on the flow and heat transfer characteristic of helium-xenon mixture inside the tight lattice rod bundles

The gas-cooled nuclear reactor combined with the closed Brayton cycle is the ideal choice for future high-power space missions. The helium-xenon mixture is a suitable working medium for the reactor due to its good compressibility and heat transfer capacity. For the further development and application of a more lightweight compact reactor, the influence mechanism of turbulator structures with different angles and layouts on the heat transfer enhancement of a helium-xenon mixture in the sub-channel of a gas-cooled nuclear reactor is numerically studied. The results show that the heat transfer is effectively enhanced due to the vortex and secondary flow induced by the turbulator, leading to periodic fluctuations in the Nusselt number. When comparing the results at different angles, it is observed that the heat transfer enhancement from the turbulator at a 90° angle is relatively small, while the pressure drop is significant. Due to the secondary flow in a narrow triangular space induced by a turbulator with a 45° angle, the mixing between high and low-temperature fluids is effectively enhanced, leading to a more significant improvement in heat transfer. The heat transfer performance and thermal efficiency index of different turbulator layouts are summarized and compared, providing a reference for the heat transfer design of gas-cooled nuclear reactors.

Experimental investigation on interfacial parameters of two-phase flow in a tight lattice rod bundle with optimized data processing method

Experimental investigation on interfacial parameters of two-phase flow in a tight lattice rod bundle with optimized data processing method

Drift-flux modeling of void fraction for boiling two-phase flow in a tight-lattice rod bundle

The concept of a resource-renewal boiling water reactor (RBWR) that uses spent nuclear fuel as a nuclear fuel has been proposed to reduce the period required for the radiotoxicity of spent nuclear fuel decaying to the same level as natural uranium ore from about 100,000 years to about 300 years. The RBWR core is designed based on the tight-lattice core concept to reduce the ratio of a moderator to fuel. Since the void fraction is one of the most critical design parameters for the RBWR, this study developed the drift-flux correlation applicable to two-phase flow in the tight-lattice rod bundle. The prediction error of the newly developed drift-flux correlation is ±11.7 %. The validated rod bundle geometrical conditions are the number of rods from 4 to 37, the rod spacing from 1.0 to 2.0 mm, the rod diameter from 9.0 to 13.7 mm, the mixture volumetric flux from 0.11 to 211 m/s, and the pressure from 0.1 to 7.2 MPa. The comparison in the distribution parameter between the tight-lattice rod bundle and the conventional BWR rod bundle indicates that the distribution parameter for the tight-lattice rod bundle is larger than that for the conventional BWR rod bundle. The increased distribution parameter in the tight-lattice core may be due to the small rod spacing and the triangular array of the rods. Both of the small rod spacing and the triangular array of the rods tend to increase the distribution parameter.

Large eddy simulation on the turbulent mixing phenomena in 3×3 bare tight lattice rod bundle using spectral element method

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.

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.

Experimental investigation on heat transfer behavior in a tight 19 rod bundle cooled with supercritical R134a

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.

Augmentation of single-phase forced convection heat transfer in tightly arrayed rod bundle with twist-vane spacer grid

The investigation on the augmentation of convection heat transfer of water rod bundle flow was performed experimentally. The 4×4 tightly arrayed rod bundle with narrow gap was tested and its P/D (Pitch-to-Diameter ratio) was ∼1.08. As the spacer grid, the twist-vane spacer grid was used. The axial mean bundle-flow velocity was U∼1.0m/s, and the corresponding Reynolds number was ∼10,400. Circumferential and axial distributions of rod-wall temperatures were measured. Upstream of twist-vane spacer grid, the rod-wall temperature at the subchannel center was measured to be lower than that at the rod-gap center. Downstream of twist-vane spacer grid, the rod-wall temperature toward the tip of twist-vane was lowest, which might be due to the deflected flow caused by twist-vane. The twist-vane spacer grid enhanced significantly the convection heat transfer in the tight-lattice rod bundle, and its maximum enhancement value was ∼77% near the downstream of spacer grid. Based on the present experimental data, the previous correlations of Nusselt number enhancement were compared and assessed. Then, the enhancement correlation of convection heat transfer for the twist-vane spacer grid in the tightly arrayed rod bundle was proposed.

Proper orthogonal decomposition of the flow in a tight lattice rod-bundle

Axial coolant flow inside a tightly packed rod bundle presents a complex behavior; experimental analysis had clearly shown that when reducing the pitch-to-diameter ratio ( P/ D) the turbulence field in rod bundles deviates significantly from that in a circular tube. Moreover for extremely tight configurations the existence of large-scale periodic “flow oscillations” has been shown, which is responsible for the high inter-sub-channel heat and momentum exchange. A complete understanding of these oscillations has still to be achieved; the evidence shown to this point suggests that the oscillations are connected to interactions between coherent structures in adjacent sub-channels. Moreover, the coherent structures show a truly three-dimensional pattern (i.e., structures in different gaps tend to interact) that has not been fully investigated up to this point. A fully transient simulation of turbulence has been performed for an infinite tight triangular lattice. In this case it has been performed with Large Eddy Simulation (LES) at Re = 6400 and P/ D = 1.05. A database of snapshots of the flow field has then been collected and proper orthogonal decomposition (POD), a powerful statistical technique, has been applied in order to obtain the most energetic modes of turbulence. The results obtained highlight the presence of several travelling waves propagating in the streamwise direction. The spatial modulation of the travelling waves offers a phenomenological explanation for observed de-coherence effects between the velocity signal in different gaps. It also provides additional insight into the three dimensional structure of the flow oscillations.

Numerical Simulation of Boiling Two-Phase Flow in Tight-Lattice Rod Bundle by 3-Dimensional Two-Fluid Model Code ACE-3D

Two-fluid model can simulate two-phase flow by computational cost less than detailed two-phase flow simulation method such as interface tracking method or particle interaction method. Therefore, two-fluid model is useful for thermal hydraulic analysis in large-scale domain such as a rod bundle. Japan Atomic Energy Agency (JAEA) develops three dimensional two-fluid model analysis code ACE-3D that adopts boundary fitted coordinate system in order to simulate complex shape flow channel. In this paper, boiling two-phase flow analysis in a tight-lattice rod bundle was performed by ACE-3D code. The parallel computation using 126 CPUs was applied to this analysis. In the results, the void fraction, which distributes in outermost region of rod bundle, is lower than that in center region of rod bundle. The tendency of void fraction distribution agreed with the measurement results by neutron radiography qualitatively. To evaluate effects of two-phase flow model used in ACE-3D code, numerical simulation of boiling two-phase in tight-lattice rod bundle with no lift force model was also performed. From the comparison of calculated results, it was concluded that the effects of lift force model were not so large for overall void fraction distribution of tight-lattice rod bundle. However, the lift force model is important for local void fraction distribution of fuel bundles.

D112 稠密燃料集合体内圧力損失に対する揚力モデルの影響の評価(OS10 混相流動)

D112 稠密燃料集合体内圧力損失に対する揚力モデルの影響の評価(OS10 混相流動)

10505 三次元二流体モデル解析コードACE-3Dによる稠密燃料集合体内圧力損失の評価(産学における混相研究(2),OS.13 産学における混相研究)

10505 三次元二流体モデル解析コードACE-3Dによる稠密燃料集合体内圧力損失の評価(産学における混相研究(2),OS.13 産学における混相研究)

Modeling Pressure Fluctuation With Cross Flow in a Tight-Lattice Rod Bundle

To explore the mechanism of differential pressure fluctuation inducing cross flow between subchannels in the tight-lattice rod bundle, an evaluation method is presented, which permits the prediction in detail of the unsteady differential pressure fluctuation behavior between subchannels. The instantaneous fluctuation of differential pressure between two subchannels in gas-liquid slug flow regime is deemed as a result of the intermittent nature of slug flow in each subchannel. The method is based on the detailed numerical simulation result of two-phase flow that pressure drop occurs mainly in the liquid slug region and it is, however, negligibly small in the bubble region. The instantaneous fluctuation of differential pressure between two subchannels is associated with pressure gradient in the liquid slug for each channel. In addition to a hydrostatic gradient, acceleration and frictional gradients are taken into account to predict pressure gradient in the liquid slug. This method used in conjunction with the numerical simulation code works satisfactorily to reproduce numerical simulation results for instantaneous fluctuation of differential pressure between two modeled subchannels. It is shown that the static head, acceleration, and frictional pressure drops in the liquid slug are main contributions to the fluctuation of differential pressure between subchannels.

Effects of Surface Tension on Two-Phase Void Drift Between Triangle Tight Lattice Subchannels

In this study, void drift phenomena, which are one of three components of the intersubchannel fluid transfer, have been investigated experimentally and analytically. In the experiments, data on flow and void redistributions were obtained for hydraulically nonequilibrium flows in a multiple channel consisting of two subchannels simplifying a triangle tight lattice rod bundle. In order to know the effects of the reduced surface tension on the void drift, water and water with a surfactant were used as test liquids. In addition, data on the void diffusion coefficient, D̃, needed in a void drift model, have been obtained from the redistribution data. In the analysis, the flow and the void redistributions were predicted by a subchannel analysis code based on a one-dimensional two-fluid model. From a comparison between the experiment and the code prediction, the present analysis code was found to be valid against the present data if newly developed constitutive equations of wall and interfacial friction were incorporated in to the model to account for the reduced surface tension effects.

Development of Analytical Procedures of Two-Phase Flow in Tight-Lattice Fuel Bundles for Innovative Water Reactor for Flexible Fuel Cycle

A research and development project to investigate thermal-hydraulic performance in the tight-lattice rod bundles of the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been in progress at Japan Atomic Energy Agency in collaboration with power companies, reactor vendors, and universities since 2002. The FLWR can realize favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burnup, and long operation cycle, based on matured light water reactor technologies. Mixed-oxide fuel assemblies with tight lattice arrangement are used because they increase the conversion ratio by reducing the moderation of neutrons. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. Information about the effects of the gap width and grid spacer configuration on the flow characteristics in the FLWR core is still insufficient. Thus, we are developing procedures for qualitative analysis of thermal-hydraulic performance of the FLWR core using an advanced numerical simulation technology. In this study, an advanced two-fluid model is developed to economize on the computing resources. In the model, interface structures larger than computational cells (such as liquid film) are simulated by the interface tracking method, and small bubbles and droplets are estimated by the two-fluid model. In this paper, we describe the outline of this model and the numerical simulations we performed to validate the model performance qualitatively.

Numerical Investigation of Cross Flow Phenomena in a Tight-Lattice Rod Bundle Using Advanced Interface Tracking Method

In relation to the design of an innovative FLexible-fuel-cycle Water Reactor (FLWR), investigation of thermal-hydraulic performance in tight-lattice rod bundles of the FLWR is being carried out at Japan Atomic Energy Agency (JAEA). The FLWR core adopts a tight triangular lattice arrangement with about 1 mm gap clearance between adjacent fuel rods. In view of importance of accurate prediction of cross flow between subchannels in the evaluation of the boiling transition (BT) in the FLWR core, this study presents a statistical evaluation of numerical simulation results obtained by a detailed two-phase flow simulation code, TPFIT, which employs an advanced interface tracking method. In order to clarify mechanisms of cross flow in such tight lattice rod bundles, the TPFIT is applied to simulate water-steam two-phase flow in two modeled subchannels. Attention is focused on instantaneous fluctuation characteristics of cross flow. With the calculation of correlation coefficients between differential pressure and gas/liquid mixing coefficients, time scales of cross flow are evaluated, and effects of mixing section length, flow pattern and gap spacing on correlation coefficients are investigated. Differences in mechanism between gas and liquid cross flows are pointed out.

Development of Design Technology on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundle: IV Large Paralleled Simulation by the Advanced Two-fluid Model Code

In Japan Atomic Energy Agency (JAEA), the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been developed. For thermal design of FLWR, it is necessary to develop analytical method to predict boiling transition of FLWR. Japan Atomic Energy Agency (JAEA) has been developing three-dimensional two-fluid model analysis code ACE-3D, which adopts boundary fitted coordinate system to simulate complex shape channel flow. In this paper, as a part of development of ACE-3D to apply to rod bundle analysis, introduction of parallelization to ACE-3D and assessments of ACE-3D are shown. In analysis of large-scale domain such as a rod bundle, even two-fluid model requires large number of computational cost, which exceeds upper limit of memory amount of 1 CPU. Therefore, parallelization was introduced to ACE-3D to divide data amount for analysis of large-scale domain among large number of CPUs, and it is confirmed that analysis of large-scale domain such as a rod bundle can be performed by parallel computation with keeping parallel computation performance even using large number of CPUs. ACE-3D adopts two-phase flow models, some of which are dependent upon channel geometry. Therefore, analyses in the domains, which simulate individual subchannel and 37 rod bundle, are performed, and compared with experiments. It is confirmed that the results obtained by both analyses using ACE-3D show agreement with past experimental result qualitatively.

Development of Design Technology on Thermal-hydraulic Performance in Tight-lattice Rod Bundles: V-Estimation of Void Fraction

Development of Design Technology on Thermal-hydraulic Performance in Tight-lattice Rod Bundles: V-Estimation of Void Fraction

Development of Design Technology on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles: I-Master Plan and Executive Summary

R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Innovative Water Reactor for Flexible Fuel Cycle has been progressed at Japan Atomic Energy Agency in collaboration with power utilities, reactor vendors and universities since 2002. In this series-study, we will summarize the R&D achievements using large-scale test facility (37-rod bundle with full-height and full-pressure), model experiments and advanced numerical simulation technology. This first paper described the master plan for the development of design technology and showed an executive summary for this project up to FY2005. The thermal-hydraulic characteristics in the tight-lattice configuration were investigated and the feasibility was confirmed based on the experiments. We have developed the design technology including 3-D numerical simulation one to evaluate the effects of geometry/scale on the thermal-hydraulic behaviors.