Fuel Rod Bundle Research Articles

Development of a flow sweeping mixing model for helical fuel rod bundles

The helical fuel rod is a kind of axially twisted fuel element with a three or four-petaled profile. The helical fuel rod assembly assembled by helical fuel rods against one another in an array is self-supporting and free of spacer grids. This study analyses the flow field in a four-petaled helical fuel rod bundle from a subchannel method perspective, and found that the mixing effects are dominated by the flow sweeping but turbulence mixing. The crossflow is two-way at interior gaps, while it is one-way at side gaps. A basic form of flow sweeping mixing model is proposed and implemented in CTF code. The coefficients are determined based on reported mixing experiments, and the model predicts the twist angle-dependent mixing rates well. Although the proposed model coefficients are geometry dependent, the derivation methodology and basic form of the model could be applied in other helical fuel rod assemblies effectively.

The numerical simulation and analysis of turbulent flow behavior in 5 × 5 fuel rod bundle with split-type mixing vane

In this research, the numerical simulation and analysis of flow behavior in 5 × 5 fuel rod bundle with consideration of spacer grid has been conducted. ANSYS FLUENT 19.0 with the consideration of the transition SST (Shear-Stress Transport) with SAS (Scale-Adaptive Simulation) turbulence model is used for the numerical simulation. The obtained simulation results are reasonable compared with the experimental data based on the comparison in the aspect of velocity distribution. The effect of the spacer grid on the flow behavior in the fuel rod bundle has also been investigated. The mixing vanes in the spacer grid can cause various strong flow branches; these flow branches develop and cause the complex flow behavior in the fuel rod bundle. Additionally, the secondary flow, the swirling factor and the mixing factor were investigated for the quantitative analysis of the spacer grid effect on the flow mixing behavior in fuel rod bundle.

Coupled porous media approaches in sub-channel CFD

In this paper, the porous media method is introduced into a recently developed Computational Fluid Dynamics (CFD)-based sub-channel analysis tool, Sub-Channel CFD (SubChCFD) through direct coupling to improve its capability in modelling complex structures, such as spacers or distorted fuel rod bundles. Two coupling methods have been developed, namely, embedding and interfacing coupling. The former is suitable for rod bundles with attached fine structures which do not change the basic geometry of the sub-channels, so a SubChCFD baseline model can be created based on the bare bundle configuration. Porous media sub-models can then be embedded in this model at locations where fine structures are installed to account for their influence on the flow and pressure drop. The interfacing method is more general and flexible, and used in a wider range of applications. In this method, separate meshes are used for the sub-domains where the porous media method is applied. These sub-domains are then interfaced with the domain covered by SubChCFD using a meshing joining technique and they are simulated together using a single set of governing equations. The methods developed are tested and validated with simulations of two rod bundle configurations, a 7 × 7 rod bundle with locally ballooned fuel rods and a 5 × 5 rod bundle equipped with simple support grids. It has been shown that a suitably selected method of the coupling can greatly simplify the modelling of complex structures using SubChCFD, thus providing additional flexibility and functionality to this newly developed CFD-based sub-channel framework.

CFD simulation of subcooled boiling flow in PWR 5 ⨯ 5 rod bundle

Abstract This paper describes the computational fluid dynamics (CFD) methodology to simulate the boiling flow in a typical Pressurized Water Reactor (PWR) 5 ⨯ 5 rod bundle. The method includes the Eulerian-Eulerian two-fluid model coupled with the improved wall heat partitioning model. The NUPEC PWR Subchannel and Bundle Test (PSBT) International Benchmark are used for validation. The simulated surface averaged void fraction agree well with the experimental data, which indicate the promising application of the present method for modeling the boiling flow in the fuel rod bundle. The main emphasis of current research has been given to the analysis of the phase distribution around and downstream the spacer grid, the effect of the spacer grid structure, including the mixing vanes, the springs and the dimples on the void fraction distribution is investigated. The findings can contribute to a better understanding of three dimensional flow boiling characteristics and can be used to assist in optimizing the spacer grid.

Precipitation profile and dryout concentration of sea-water pool-boiling in 5 × 5 full-height BWR bundle

Sea water shall be injected into water-cooled nuclear reactors during severe accidents, which are located along coastal side to flood the nuclear fuel, which is heated by residual heat. Precipitation growth to narrower flow path area is a key to gain the confidence of accident mitigation procedure to cool down the reactor core during accidental conditions. A pool boiling experiment was conducted with a simulated 5 × 5 full-height BWR fuel-rod bundle with condensed (two and half times higher concentration) sea water. The temperature on the center rod surface in the top spacer rose rapidly, since the flow area inside the top spacer was filled with the precipitated salt. Dryout below the top spacer escalated temperatures of the heater surface. On the other hand, the heater above the top spacer was cooled stably by pool boiling. An example calculation estimates that the dryout due to salt precipitation may occur 19 h after sea water injection for an ABWR, which had operated at 3.926 GWt for 13 months on the basis of critical dryout concentration of 50 wt%.

Effects of flow-induced vibration on forced convection heat transfer from two tandem circular cylinders in laminar flow

Flow-induced vibrations (FIVs) and forced convection heat transfers of two tandem circular cylinders in laminar flow with Re=150 and Pr=0.7 are investigated using fluid-structure interaction (FSI) simulation. The cylinder is heated, elastically supported and only allowed to vibrate in the transverse direction, with the mass ratio, damping ratio, centre-to-centre distance fixed respectively at m*=2, ζ=0, L/D=4 and the reduced velocity varied in Ur=1~20. The two-dimensional incompressible Navier-Stokes equations and energy equation are coupled with the vibrating equation of the spring-mass-damper model to describe FIV of two tandem cylinders involving forced convection heat transfer process. An FSI solver is proposed by combing the modified characteristic-based split scheme, dual-time stepping method, segment spring analogy method and loosely coupled partitioned approach, the accuracy and stability of which are validated using two benchmark models including the forced convection heat transfer from a fixed circular cylinder and FIV of an elastically supported circular cylinder. Using FSI simulation, characteristics of FIVs and their effects on forced convection heat transfers of two tandem circular cylinders are analyzed in detail. The obtained numerical results are insightful to the understanding of the heat transfer features of a group of cylindrical structures under FIVs, such as the tube bundles in heat exchangers and power plants, fuel rod bundles in nuclear reactors and riser system in offshore oil and gas platforms.

Hydrodynamics of coolant in VVER reactor core with fuel assemblies of various designs

The article presents the results from experimental investigations into local coolant flow fluid dynamics that were carried out on the fragmental model of a VVER-type pressurized water reactor’s mixed core consisting of two types of fuel assemblies (FAs): TVSA-T (one segment) and TVSA-T.mod.2 (two segments). The coolant flow processes in the fuel rod bundle were simulated on an aerodynamic test bench. The pressure field in the flow was measured using a five-channel pneumometric probe. The obtained pressure field in the flow was recalculated for the coolant velocity vector according to the dependences derived during calibration. For drawing up a detailed flow motion pattern, a characteristic model cross section area was separated, which included the space between the assemblies and four fuel rod rows in each TVSA-type fuel assembly. The spatial distribution of coolant flow velocity projections was analyzed, as a result of which it became possible to reveal the regularities associated with the streamlining of spacer, mixing-, and combined-spacer grids in TVSA assemblies by coolant; to determine coolant cross flows resulting from streamlining of hydraulically nonidentical grids and to determine their localization in the experimental model longitudinal and cross sections; and to reveal the effect of accumulating flow hydrodynamic disturbances in the model longitudinal and cross sections resulting from the staggered arrangement of hydraulically nonidentical grids. The results obtained from investigations of interassembly interaction of coolant between the neighboring TVSA-T and TVSA-T.mod.2 assemblies have been adopted for practical use at JSC “Afrikantov OKBM” in estimating the thermal reliability of VVER-type reactor cores and have been included in the database for verification of computation fluid dynamics computer programs (CFD codes) and detailed cell-wise numerical analysis of VVER reactors’ core.

Numerical investigation on thermal-hydraulic performance in 7×7 rod bundle with spacer grid and guide tubes

In this paper, the thermal-hydraulic performance in the downstream of the spacer grid in a 7 × 7 rod bundle is numerically studied. The guide tubes effects are revealed by the comparisons of the flow and heat transfer behavior in two types of rod bundles, one without guide tubes and the other with guide tubes. It is found that the cross flow diagonals are fully developed at Z = 2 Dh and unable to sustain at Z = 8 Dh downstream of the spacer grid. The guide tubes discontinue the cross flow diagonals and would decrease the secondary flow intensity. The downstream temperature distribution is influenced by the evolution of the secondary flow as well as the guide tubes. The non-uniformity of the temperature distribution has the lowest value at Z = 2 Dh and the guide tubes would make the temperature distribution more uneven. Furthermore, the normalized Nusselt number in both fuel rod bundles reaches the peak at Z = 2 Dh and then decreases consistently. However, the heat transfer can be enhanced by 15% at least because the secondary flow still exists. The normalized Nusselt numbers downstream of the spacer grid in the rod bundle with guide tubes are smaller than those without guide tubes. Overall, the guide tubes would inhibit the secondary flow and weaken the heat transfer capability downstream of the spacer grid.

CONV-3D Code Aided Calculation of the Flow Parameters of Fuel-Rod Bundles

The results of modeling of incompressible fluid flows in fuel rod bundles performed using the CONV-3D code are reported. The modeling results obtained for different Reynolds numbers and distances between the fuel rods in comparison with experiment are presented. It is shown that the developed code satisfactorily models the flow of an incompressible fluid in fuel rod bundles and can be used to analyze such flows in different applications.

Study on the effects of flow-driven rotational mixing vanes by flow pattern tracking

In pressurized water reactor (PWR), spacer grid is installed to support the fuel rod bundles, located between the fuel rod bundles. The mixing vane is installed on top of the spacer grid to generate swirl and cross flow. The swirl and cross flow enhance the heat transfer and can promote a critical heat flux (CHF) in the PWR. The PWR safety margin can be estimated using the heat-transfer performance and the CHF. In this research, flow-driven rotational mixing vanes were installed. Fan, impeller, and wind-turbine mixing vanes were adopted, which could rotate without any artificial force. The fluid flow and rotational motion simultaneously influenced each other.The effects of the flow-driven rotational mixing vanes were presented as flow patterns. A flow-pattern tracking method demonstrated the effects of various mixing parameters. The rotational speed was validated using a high-speed camera experiment and a computational analysis using a general moving-object method. The lateral velocity reflected the cross-flow magnitude with respect to the axial velocity. The vorticity was considered as a performance measure for the rotational mixing vanes. The bubble behavior demonstrated the mixing effect of the fluid in the subchannel. The pressure drop was gauged to verify the flow blockage. Thus, using flow-pattern tracking, the mixing effect of the parameters can be used to quantify the power-uprate enhancement and the safety margin for a PWR.

Combined Numerical and Experimental Investigations into the Local Fluid Dynamics of Coolant Flow in the Mixed Core of a VVER Reactor

The article presents the results from experimental investigations into local coolant flow fluid dynamics that were carried out on the fragmental model of a VVER-type pressurized water reactor’s mixed core consisting of two types of fuel assemblies (FAs): TVSA-T (one segment) and TVSA-T.mod.2 (two segments). The coolant flow processes in the fuel rod bundle were simulated on an aerodynamic test bench. The pressure field in the flow was measured using a five-channel pneumometric probe. The obtained pressure field in the flow was recalculated for the coolant velocity vector according to the dependences derived during calibration. For drawing up a detailed flow motion pattern, a characteristic model cross section area was separated, which included the space between the assemblies and four fuel rod rows in each TVSA-type fuel assembly. The spatial distribution of coolant flow velocity projections was analyzed, as a result of which it became possible to reveal the regularities associated with the streamlining of spacer-, mixing-, and combined-spacer grids in TVSA assemblies by coolant; to determine coolant cross flows resulting from streamlining of hydraulically nonidentical grids and determine their localization in the experimental model longitudinal and cross sections; and to reveal the effect of accumulating flow hydrodynamic disturbances in the model longitudinal and cross sections resulting from the staggered arrangement of hydraulically nonidentical grids. The results obtained from investigations of inter-assembly interaction of coolant between the neighboring TVSA-T and TVSA-T.mod.2 assemblies have been adopted for practical use at AO Afrikantov OKBM in estimating the thermal reliability of VVER-type reactor cores and have been included in the database for verification of computation fluid dynamics computer programs (CFD codes) and detailed cell-wise numerical analysis of the core of VVER reactors.

Investigation of Coolant Local Hydrodynamics in the Mixed Core of the VVER Reactor

The article presents the results of experimental studies of the local hydrodynamics of the coolant flow in the mixed core of the VVER reactor, consisting of the TVSA-T and TVSA-T mod.2 fuel assemblies. Modeling of the flow of the coolant flow in the fuel rod bundle was carried out on an aerodynamic test stand. The research was carried out on a model of a fragment of a mixed core of a VVER reactor consisting of one TVSA-T segment and two segments of the TVSA-T.mod2. The flow pressure fields were measured with a five-channel pneumometric probe. The flow pressure field was converted to the direction and value of the coolant velocity vector according to the dependencies obtained during calibration. To obtain a detailed data of the flow, a characteristic cross-section area of the model was selected, including the space cross flow between fuel assemblies and four rows of fuel rods of each of the TVSA fuel assemblies. In the framework of this study the analysis of the spatial distribution of the projections of the velocity of the coolant flow was fulfilled that has made it possible to pinpoint regularities that are intrinsic to the coolant flowing around spacing, mixing and combined spacing grates of the TVSA. Also, the values of the transverse flow of the coolant caused by the flow along hydraulically nonidentical grates were determined and their localization in the longitudinal and cross sections of the experimental model was revealed. Besides, the effect of accumulation of hydrodynamic flow disturbances in the longitudinal and cross sections of the model caused by the staggered arrangement of hydraulically non-identical grates was determined. The results of the study of the coolant cross flow between fuel assemblies interaction, i.e. between the adjacent TVSA-T and TVSA-T mod.2 fuel assemblies were adopted for practical use in the JSC of “Afrikantov OKB Mechanical Engineering” for assessing the heat engineering reliability of VVER reactor cores; also, they were included in the database for verification of computational hydrodynamics programs (CFD codes) and for detailed cell-based calculation of the reactor core.

Flexural Vibration Analysis of Nuclear Fuel Rod Bundles Interacting with Surrounding Fluid Subjected to Pressure Wave

The structural behavior of the nuclear rod bundles that consisted of cylindrical beams was predicted using the spectral element method (SEM) while considering the interaction with the surrounding fluid. Viscous fluid behavior was utilized in order to calculate the forces acting on the nuclear rod bundles from the incident pressure waves. The added mass and fluid coupling on the nuclear rod bundles were determined for the position patterns and gaps of each of the cylindrical beams. The pressure field from propagating waves in the surrounding fluid was calculated with respect to the boundary conditions of the surface of the vibrating structures. With the increasing number of nuclear rods and decreasing pressure wavelengths, the structural vibration of the nuclear rod bundles that were induced by the propagating forces affected the scattering events of the pressure field. The frequency response of the nuclear rod bundles from the pressure waves in the water exhibited smaller damping, because the incident pressure wave travels without fluid coupling due to the longer wavelength when compared with distance between rods. The proposed numerical method can be utilized for the detailed design for effective parameters of a supporting system to reduce the vibration of nuclear fuel rod bundles for safety control.

Visualization of three-dimensional boiling two-phase flow in a fuel rod bundle

流れ現象は日常生活から工業プラントまで身近に存在するが，それを定量的なものとして把握するために可視化技術が発展してきた。原子炉燃料集合体はその複雑な構造内部を水が流れながら除熱や中性子減速材の役割を果たす。本稿では，当所が研究開発を進めている燃料集合体内の沸騰二相流計測技術に焦点をあてる。これまでに開発した計測技術の測定原理や計測機器の概略を示し，模擬燃料集合体を用いた計測事例を紹介する。

Hydraulic and statistical study of metastable phenomena in PWR rod bundles

The analysis of fuel rod bundle flows constitutes a key element of Pressurized-Water Reactors (PWR) safety studies. The present work aims at improving our understanding of nefarious reorganisation phenomena observed by numerous studies in the flow large-scale structures. 3D simulations allowed identifying two distinct reorganisations consisting in a sign change for either a transverse velocity in rod-to-rod gaps or for a subchannel vortex. A Taylor “frozen turbulence” hypothesis was adopted to model the evolution of large-scale 3D structures as transported-2D. A statistical method was applied to the 2D field to determine its thermodynamically stable states through an optimization problem. Similarities were obtained between the PWR coherent structures and the stable states in a simplified 2D geometry. Further, 2D simulations allowed identifying two possible flow bifurcations, each related to one of the reorganisations observed in 3D simulations, laying the foundations for a physical explanation of this phenomenon.

CFD simulation of flow and heat transfer characteristics in a 5×5 fuel rod bundles with spacer grids of advanced PWR

High fidelity nuclear reactor fuel assembly simulation using CFD method is an effective way for the structure design and optimization. The validated models and user practice guidelines play critical roles in achieving reliable results in CFD simulations. In this paper, the international benchmark MATiS-H is studied carefully and the best user practice guideline is achieved for the rod bundles simulation. Then a 5 × 5 rod bundles model in the advanced pressurized water reactor (PWR) is established and the detailed three-dimensional thermal-hydraulic characteristics are investigated. The influence of spacer grids and mixing vanes on the flow and hear transfer in rod bundles is revealed. As the coolant flows through the spacer grids and mixing vanes in the rod bundles, the drastic lateral flow would be induced and the pressure drop increases significantly. In addition, the heat transfer is enhanced remarkably due to the strong mixing effects. The calculation results could provide meaningful guidelines for the design of advanced PWR fuel assembly.

Fretting Wear and Fatigue Life Analysis of Fuel Bundles Subjected to Turbulent Axial Flow in CEFR

In a fast spectrum reactor, the fuel rod bundle is mainly positioned radially by the wire which can make contact with the adjacent fuel rods, and then it is inevitable that flow-induced vibration (FIV) will cause fretting wear and vibration fatigue of the fuel cladding at the contact position. Therefore, the computational model of fretting wear and fatigue life about the fuel rod bundle caused by FIV will be studied in this paper. Based on the random vibration model of the fuel rod bundle, Hertz contact theory, and Archard wear theory, the fretting wear life computational model and the fatigue life computational model of the wire-to-adjacent fuel rod (WAFR) contact have been established. Finally, taking CEFR design parameters as an example, the fretting wear life and vibration fatigue life of the cladding are calculated, and it is found that fatigue affects the service life of the fuel rod more seriously than fretting wear. The calculation model and method lay a foundation for further accurate prediction and analysis of the fuel rod life.

Void-fraction measurement with high spatial resolution in a 5×5 rod bundle by linear-accelerator-driven X-ray computed tomography over a wide pressure range

Void fraction (i.e., the volume fraction occupied by gas) is a key parameter for determining the coolability and neutron-moderating performance of a water-cooled nuclear reactor. To develop computational multi-fluid dynamics models for determining the void-fraction distribution, experimental data of comparable quality are required. We have developed a high-energy X-ray computed tomography (CT) system to acquire three-dimensional void-fraction distributions. The CT system comprises a linear-accelerator-driven high-energy X-ray source and a linear detector array. We quantified a boiling two-phase flow in a 5 × 5 heated rod bundle at high pressure, simulating a fuel-rod bundle in a boiling water reactor (BWR). Because the axial travel of the CT system is 4 m and includes the entire BWR fuel-rod bundle, we optimized the CT imaging conditions and reconstruction method for rod-bundle visualization to reduce uncertainties due to density fluctuations in the boiling flow and imaging artifacts. We conducted a boiling experiment at a low flow rate and low thermal power and acquired three-dimensional distributions of the void fraction over a wide pressure range of 0.1–7.2 MPa. The experiment provided three-dimensional void-fraction distributions with high spatial resolution, especially in subchannel regions surrounded by rods, and the results are suitable for validating three-dimensional thermal-hydraulic analysis codes.

Computational Fluid Dynamics Simulation of Flow-Mixing and Heat Transfer in 4 × 4 Rod Bundle With a Twist-Vane Spacer Grid

The clearance between fuel rods is maintained by spacer grid or helical wire wrap. Thermal-hydraulic characteristics inside fuel rod bundle are strongly influenced by the spacer grid geometry and the bundle pitch-to-diameter (P/D) ratio. This includes the maximum fuel temperature, critical heat flux, as well as pressure drop through the fuel bundle. An understanding of the detailed structure of flow mixing and heat transfer in a fuel rod bundle geometry is therefore an important aspect of reactor core design, both in terms of the reactor's safe and reliable operation, and with regard to optimum power extraction. In this study, computational fluid dynamics (CFD) simulations are performed to investigate isothermal turbulent flow mixing and heat transfer behavior in 4 × 4 rod bundle with twist-vane spacer grid with P/D ratio of 1.35. This work is carried out under International Atomic Energy Agency (IAEA) co-ordinated research project titled as “Application of Computational Fluid Dynamics (CFD) Codes for Nuclear Power Plant Design.” CFD simulations are performed using open source CFD code OpenFOAM. Numerical results are compared with experimental data from Korea Atomic Energy Research Institute (KAERI) and found to be in good agreement.

Verification of the TIGRSP Computer Code as Applied to a 19-Rod Fuel Assembly Using CFD Computations

For assessing the core critical heat flux ratio in VVER-type nuclear reactors, the lattice computation code TIGRSP (steady-state thermal and hydraulic analysis of parameters in fuel rod bundles) is used. Despite the fact that quite a large amount of verification studies has been accomplished and that the code has passed the certification procedure, work on improving the code capabilities is still being continued. One of the modern tendencies is the wide-scale involvement of CFD codes for studying local parameters in fuel assemblies (FAs). Along with comparison with experimental data, it is advisable to carry out cross verification of lattice codes with the results of 3D thermal-hydraulic computations for conditions for which experimental data are either lacking or limited for some or other reasons. This article considers thermal-hydraulic processes during coolant flow in a 19-rod assembly with parameters close to the nominal conditions in the FAs of a VVER-1000 reactor. The 3D model is verified with respect to empirical formulas and experimental data. Based on the results of test and verification computations, we determined the mesh and turbulence model parameters using the ANSYS CFX code by means of which it is possible to obtain reliable results in estimating the hydraulic characteristics of FAs fitted with spacer grids, heat transfer in the single-phase region, and coolant flow distribution over the FA cross section. The CFD computation of coolant flow for the 19-rod assembly is carried out, and the coolant mass velocity and temperature in different sections over the length and corresponding cells of the TIGRSP code are calculated. The TIGRSP code has been verified against experimental data and taking into account the CFD simulation of the 19-rod FA, and a software module for graphically visualizing the lattice code output results has been developed. The presented data can be used for verification of thermal-hydraulic codes and in elaborating the design of FAs for VVER-type reactors.