Wire-wrapped Rod Bundle Research Articles

Validating RANS to predict the flow behavior in wire-wrapped fuel assemblies

Liquid metal fast reactors (LMFRs) represent one of the most promising technologies in terms of sustainability, safety and reduction of nuclear waste and are expected to play a relevant role in the coming future. In liquid metal fast reactors, wire wraps convoluted helically along the fuel pins are often used as a spacer design, in order to avoid pin-to-pin contact and maintain a constant spacing among the fuel rods, with the result of altering noticeably the flow and heat transport, by introducing additional turbulence and establishing a bulk rotation of the fluid.Three different reference cases have been used for validation of the flow behaviour in engineering simulations, an infinite bundle, a 7-pin bundle, and a 61-pin bundle. In addition, the infinite bundle data also provides the opportunity to validate the thermal behaviour.A high fidelity Direct Numerical Simulation (DNS) database has been generated for a wire-wrapped fuel assembly configuration. This database is used for validation of engineering simulation approaches based on Reynolds Averaged Navier Stokes (RANS) modelling. Periodicity is conveniently applied thanks to the opportune definition of the computational domain. The dimensions are selected such that they are representative for the MYRRHA reactor fuel assembly design. Extensive analyses of flow and thermal fields, which in a DNS database are available simultaneously, while are hard to achieve in liquid metal experiments, have been performed and used in the validation process for RANS calculations. Mesh sensitivity as well as a sensitivity analysis on different turbulence models are performed. Both qualitative and quantitative analyses are presented in terms of velocity, temperature and turbulent kinetic energy profiles. Discussion on the different turbulence models adopted is presented.Simulations are compared with experimental measurements as well. A series of experiments have been performed for a 7-pin wire-wrapped rod bundle at varying flow rates. A matched-index-of-refraction technique was used employing water as a working fluid in the rod bundle. RANS calculations reproduced the same experiments and results are compared with the available measurements.Finally, a 61-pin wire-wrapped rod bundle case is simulated using steadystate RANS. The simulations reproduced isothermal flow in exactly the same geometry as the high-fidelity Large Eddy Simulation (LES) performed at Argonne National Laboratory, USA. Results of the LES are used as a reference data for validation of the RANS calculations. Parallel to this work, the LES results are being compared to matched-index-of-refraction experiments performed at the Texas A&M university. Even though this is outside the scope of the work presented in this paper, it will provide added value to the comparison in the future.

Numerical investigation of thermal-hydraulic behaviors in a LBE-cooled 19-pin wire-wrapped rod bundle

Thermal hydraulics of heavy liquid metal (HLMs), such as lead-bismuth eutectic(LBE) which is one of choices for new generation of nuclear power plant designs has drawn a lot of attentions. The wire-wrapped rod bundle has a great influence on the inter-channel mixing and leads to a non-uniform temperature profile over the rod surface. In this paper, a CFD modeling and simulation for a LBE-cooled 19-pin wire-wrapped rod bundle in which pins are separated by helical wire-wrapped spacer was carried out based on RANS method. Three-dimensional flow and temperature distributions of LBE were simulated using commercial CFD software. The simulation results are firstly compared with the experimental data from the literature. A good agreement demonstrated the capability of the employed approach to predict the flow and temperature field in the rod bundle. Consequently, the influence from mass flow rate on the inter-channel mixing caused by the wire-wrapped spacer was investigated. The cross flow between sub-channels through gap has been evaluated as well as its induced temperature variance over the rod surface. Their dependency on mass flow rate is also evaluated based on CFD simulation results and it can be employed in the sub-channel analysis of reactor core in the future.

CFD simulation on the flow and heat transfer characteristics of mist flow in wire-wrapped rod bundle

The heat transfer characteristics of post-CHF (critical heat flux) is very important and complicated, which has aroused wide attention in the topic of nuclear reactor safety. In this paper, a phase change model of liquid droplet during mist flow is proposed to consider the heat transfer of mist flow. Couple with this phase change model, three dimensional CFD simulation has been carried out to study the flow and heat transfer characteristics of mist flow in a wire-wrapped rod bundle. The discussion and analysis focus on the turbulent flow behavior and the heat transfer performance of post-CHF mist flow in the rod bundle. The effect of vapor properties such as density, heat capacity, thermal conductivity and viscosity on heat transfer are considered in the CFD simulation, and those properties are dependent on temperature. According to the CFD simulation result, the pressure distribution, temperature distribution, velocity distribution and phase distribution in the wire-wrapped rod bundle are analyzed in detail.

CFD analysis of flow blockage phenomena in a LBE-cooled 19-pin wire-wrapped rod bundle

Flow blockage accident is a severe accident. When fuel assembly is partially or totally blocked, the heat transfer between coolant and cladding is significant impaired due to a rapid decrease of coolant flow, which leads to a local increase of cladding and coolant temperature and possibly results in a release of fission products. A recirculation region is formed downstream the blockage and a temperature peak can be found in this region. In this paper, CFD modelling and simulation for a LBE-cooled 19-pin wire-wrapped bundle were carried out to consider the influence from the flow blockage on the rod surface temperature and cross flow. Three-dimensional flow and temperature distributions of LBE were obtained using commercial CFD software. The influence from the blockage area and its location was investigated as well as the axial length of blockages. It was found that flow blockages in the center of bundle result in a higher surface temperature while comparing with blockages in the side sub-channels. The cross flow between sub-channels through gaps has been evaluated and their dependency on the blockage area is also evaluated based on CFD simulation results.

CFD Analysis of Scaled Wire-wrapped rod Bundle Cooled by LBE

Lead bismuth eutectic (LBE) is one of the most potential materials for coolant for Lead based reactor and Accelerator Driven Systems (ADS). Thermal-hydraulic behaviour of LBE in fuel assembly is a key issue for development of the systems. To get a deeper understanding on the complex thermal-hydraulic features of wire-wrapped rod bundle cooled by upward LBE, CFD calculation based on RANS methodologies were also performed to support the experimental results analysis. The results concluded that LBE has the similar flow resistance characteristics with traditional fluids. Both the Rehme correlation and CFD showed a good agreement with the experimental results. As for the entrance characteristics, during the fully heating length (exceeding 140 times the hydraulic diameter), the thermal field did not reach a fully developed and stable condition which is contrary to the ducted flows. Based on the experimental results and CFD investigation of heat transfer coefficient showed that the hexagonal shell has a great influence on the heat transfer coefficient in rod bundle geometry. For this reason, the application of empirical correlation should be kept cautious in rod bundle analysis.

Corrigendum to “The upgraded Cheng and Todreas correlation for pressure drop in hexagonal wire-wrapped rod bundles” [Nucl. Eng. Des. 335 (2018) 356–373

Corrigendum to “The upgraded Cheng and Todreas correlation for pressure drop in hexagonal wire-wrapped rod bundles” [Nucl. Eng. Des. 335 (2018) 356–373

Experiment design to assess the inter-wrapper heat transfer in LMFR

Abstract The inter-wrapper flow (IWF) between fuel assemblies in an LMFR plays an important role in the reduction of the peak cladding temperature in abnormal scenarios, such as blockages or flow reductions. This paper focusses on the heat transfer through the IWF in a geometry which is representative for the MYRRHA reactor under design by SCK•CEN in Belgium. To this purpose an LBE experiment is designed in the Karlsruhe Lead LAboratory (KALLA) at KIT in Germany. Three 7-pin wire-wrapped rod bundles, creating a rotating flow representative for a MYRRHA wire-wrapped fuel assembly, and the adjacent inter-wrapper region form the experimental setup. The thickness of the wrapper and the region for the inter-wrapper flow have the same dimensions as selected for MYRRHA. The instrumentation is designed such that the temperatures and heat fluxes can be measured and used to validate Computational Fluid Dynamics (CFD) models for the inter-wrapper flow. This validation will allow CFD models to be applied in full core models for liquid metal cooled reactors. This paper concerns the experiment design and the supporting CFD calculations. The influence of the equipment upstream of the test section is assessed to determine the domain size and boundary conditions required for the code validation simulations. The selected domain and boundary conditions are subsequently applied in simulations to determine the most suitable locations for the instrumentation to measure the most interesting phenomena of the heat transfer through the inter-wrapper region and to be suitable for CFD validation.

The upgraded Cheng and Todreas correlation for pressure drop in hexagonal wire-wrapped rod bundles

The Cheng and Todreas (detailed) correlation (CTD) for wire-wrapped rod bundle friction factor published in 1986 has been identified in Chen et al. (2014) as the most used correlation for design and safety analyses for the Generation IV Sodium Fast Reactor(SFR). Since its publication in 1986, sixteen additional wire wrapped fuel bundle experiments have been performed worldwide providing data to supplement regions of minimal data in the bundle test data base. Based on this valuable new data, the CTD correlation has been refined in the following three aspects and renamed the Upgraded CTD (UCTD) Correlation.1.The laminar to transition boundary Reynolds value has been significantly reduced. As well, a new laminar boundary equation is adopted which gives much lower value for the laminar boundary (RebL) as a function of P/D.2.The transition region equation for bundle friction factor modified by Chen et al. (2013) to make the friction factor trajectory smoother moving through the turbulent boundary now has the exponent λ reassessed to the value of 7.3.The UCTD is now applicable to 7-pin bundle geometry. This has been achieved by adjusting the bundle average friction factor predicted by CTD for most bundle geometries such that the bundle fiction factor increases as the pin number increases. The cause of the previous inverse behavior was identified as the formulation of the wire drag and wire swirl empirical constants. A new process for finding these empirical constants is proposed and the constants set is calibrated by an 80 bundle data set. For Rehme’s set of twenty 7 pin bundle experiments, the mean error prediction is now 9% which is about the measurement uncertainty and improved from 26% with CTD.The new correlation incorporating the above three improvements to the original CTD correlation is designated as the Upgraded CTD (UCTD) correlation. For the available 118 wire-wrapped bundles with pin number greater than 7 bundles, UCTD predicts the data in the transition and turbulent regimes with a mean error of 0.95% and a 90% confidence interval of ±14.8%, while both the CTD and the Rehme correlations (REH) have 1.5% higher values of this index. For the 27 available bundles with laminar data,(none of which are as small as 7 pin bundles) the performance of UCTD is slightly better than CTD with mean and RMS values of −2.0%, and 9.5% versus −3.0%, and 9.6%. (The REH correlation is not applicable to the laminar regime).A web-based numerical code of the UCTD correlation which incorporates these advances has been provided on the web for convenient design and research use.

CFD investigation on thermal-hydraulic behaviors of a wire-wrapped fuel subassembly for sodium-cooled fast reactor

Different from the square-arrayed fuel rod bundles in pressurized water reactors, the fuel subassembly for sodium-cooled fast reactors features with wire spacers helically wrapped on the fuel rod. This intricate geometry consequently imposes a significant challenge in numerical simulation. In the present study, a new marked and separated meshing method is developed to reduce the required mesh and modeling time. Instead of directly generating the complex geometry and mesh, the hexahedral wire wrapper mesh was dealt with locus equations. First, the numerical modeling method was validated by combining the experimental data of a 19-pin wire-wrapped rod bundle and the simplified Cheng and Todreas correlation. Furthermore, a comparative study of different meshing methods was conducted. Then, an investigation on the three-dimensional flow phenomenon and heat transfer characteristics of a 61-pin wire-wrapped fuel subassembly for the China Experimental Fast Reactor was carried out. Finally, complex axial and transverse flow behaviors and temperature field in different types of subchannels were quantitatively analyzed. Results indicate that the axial velocity exhibits periodic oscillation with a maximum normalized value of 1.54. The flow oscillation in the interior subchannels presents a smaller amplitude and higher frequency than that in the corner and edge subchannels. Local hot spots occur in the gaps beneath the spacer wires. Sweeping and cross flow patterns exist in the peripheral and interior subchannels, respectively. In addition, the secondary transverse flow induced by the spacer wires is stronger in the corner subchannel. The cross flow strength reaches a peak value equal to 30% of the average axial velocity, which promotes the thermal mixing.

Computational Fluid Dynamic Simulations of Heat Transfer From a 2 × 2 Wire-Wrapped Fuel Rod Bundle to Supercritical Pressure Water

Within the Generation-IV International Forum, Canadian Nuclear Laboratories (CNL) led the conceptual fuel bundle design effort for the Canadian supercritical water cooled reactor (SCWR). The proposed fuel rod assembly for the Canadian SCWR design comprised of 64-elements with spacing between elements maintained using the wire-wrap spacers. Experimental data and correlations are not available for the fuel-assembly concept of the Canadian SCWR. To analyze the thermalhydraulic performance of the new bundle design, CNL is using computational fluid dynamics (CFD) as well as the subchannel approach. Simulations of wire-wrapped bundles can benefit from the increased fidelity and resolution of a CFD approach due to its ability to resolve the boundary layer phenomena. Prior to the application, the CFD tool has been assessed against experimental heat transfer data obtained with bundle subassemblies to identify the appropriate turbulence model to use in the analyses. In the present paper, assessment of CFD predictions was made with the wire-wrapped bundle experiments performed at Xi'an Jiaotong University (XJTU) in China. A three-dimensional CFD study of the fluid flow and heat transfer at supercritical pressures for the rod-bundle geometries was performed with the key parameter being the fuel rod wall temperature. This investigation used Reynolds-averaged Navier–Stokes turbulence models with wall functions to investigate the behavior of flow through the wire-wrapped fuel rod bundles with water subjected to a supercritical pressure of 25 MPa. Along with the selection of turbulence models, CFD results were found to be dependent on the value of turbulent Prandtl number used in simulating the experimental test conditions for the wire-wrapped fuel rod configuration. It was found that the CFD simulation tends to overpredict the fuel wall temperature, and the predicted location of peak temperature differs from the measurement by up to 65 deg.

Stereoscopic PIV measurements of near-wall flow in a tightly packed rod bundle with wire spacers

In this study, we experimentally investigated the flow field characteristics in the complex geometry of a tightly packed 61-rod bundle with wire spacers by employing matched-index-of-refraction (MIR) and stereoscopic particle image velocimetry (SPIV) techniques. SPIV measurements were performed on a cross-flow plane at Reynolds numbers of 500, 2500, and 6300, and the measurement area covered the interior and edge sub-channels that were located near the enclosure wall. The SPIV results showed that peaks of the velocity magnitude occurred in the edge sub-channel, indicating the wall effects on the flow symmetry in the cross-flow plane. Mean velocities at the central areas of the edge and interior sub-channels showed flat profiles for Re=500 and local peaks around 0.15<y/Drod<0.3, due to the wire’s presence and near-wall location of the edge sub-channel. The rms fluctuating profiles velocities showed that local maximal peaks linearly shifted from y/Drod=-0.44 to −0.36 and −0.28 as Re increased. Taylor’s hypothesis was applied to the SPIV velocity vector fields for reconstructing quasi-instantaneous three-dimensional (3D) vortical structures in the wire-wrapped rod bundle. The visualization of 3D vorticity iso-surfaces showed that vortical structures were generated from shear layers of neighboring rods and traveled in the streamwise direction. Finally, proper orthogonal decomposition (POD) analysis was applied to SPIV velocity snapshots taken at the edge and interior sub-channels to extract the most statistically dominant flow structures. The POD velocity decomposition revealed coherent structures with sizes and shapes comparable to instantaneous vortical structures identified by iso-surfaces. Energy fractions of the first POD modes were found to reduce from 46% and 31% to 26% and 18% when Reynolds numbers increased from Re=500 to 6300, respectively. On the other hand, kinetic energy levels of low-order POD modes (modes 2, 3, 4, etc.) increased when the Reynolds number increased.

Numerical simulations for determination of minimum representative bundle size in wire wrapped tube bundles

When constructing an experiment or simulation for flow through a wire-wrapped fuel rod bundle, scientists may utilize a smaller bundle than design in order to minimize material or computational costs. Small bundles may not capture the required flow physics experienced by larger bundles. This paper compares the flow fields through wire-wrapped rod bundles of 19, 37, 61, and 91 pins and investigates the ability of each to capture the relevant physics of a larger bundle. For model verification, the SST k-ω and elliptic blending k-∊ RANS models were compared against an LES simulation of the 19 pin domain, finding both RANS turbulent models capable for the given geometry. The central subchannel transverse velocity and inter-channel mass exchange for each bundle was compared, revealing a strong dependence of inter-channel mixing on the bundle size. Furthermore, analysis of the inter-channel mass exchange for subchannels at a varying distance from the surrounding shroud revealed than the mass exchange as a function of height is strongly tied to distance from the wall, with a slight adjustment in magnitude for the bundle size. The asymmetrical aspect of the wall effect was observed, revealing that the outside 2 or 3 rings of subchannels experience a large deviation from the characteristic subchannel behavior, but the central subchannels of the 37, 61, and 91 pin bundles are largely isolated from the asymmetric wall effect. Based on these findings the authors recommend a 37 pin bundle as the minimum surrogate for a larger bundle, and 61 pins as the preferred bundle size.

Experimental study of the flow characteristics in an SFR type 61-pin rod bundle using iso-kinetic sampling method

Hydrodynamic experiments were performed to characterize pressure drop and flow distribution in the subchannels of KAERI’s fuel bundle configuration. A 61-pin wire wrapped rod bundle was chosen as a test assembly considering the hydrodynamic similarity of the prototype reactor. The rod diameter, rod length and rod pitch were 8.0mm, 1500mm and 1.14, respectively. Three instrument pins were specially fabricated to measure the pressure drop. Flow measurements for all subchannels including interior, edge and corner subchannels, were performed under iso-kinetic conditions. The nominal operating flow conditions of the test assembly were 15.02kg/s at 60°C with water, which is equivalent to Re∼60,750. Conventional hydrodynamic correlations were assessed using the measured pressure drop and subchannel flow rate data. The experimental data of the pressure drop and flow distribution in the subchannels can be utilized to validate hydrodynamic models and subchannel analysis codes.

On the Safety and Performance Demonstration Tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and Validation and Verification of Computational Codes

Abstract The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V&V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

Heat transfer from a 2 × 2 wire-wrapped rod bundle to supercritical pressure water

An experiment has been completed to study the heat transfer characteristics of supercritical water in a 2×2 wire-wrapped rod bundle. The rod-bundle simulated a fuel assembly with four heated rods and was installed inside a square channel with rounded corners. A wire of 1.2mm outer diameter was helically wrapped around the outer surface of each heated rod at an axial pitch of 200mm. The experimental parameters covered the pressure of 23–28MPa, mass flux of 400–1000kg/m2s and heat flux range on the rod surface of 200–1000kW/m2. Wall temperatures obtained at each heated rod were examined and average heat transfer coefficients were calculated at each measuring location of the rod and compared against those obtained with a 2×2 bare rod bundle. Heat transfer characteristics of supercritical water in the 2×2 rod bundle were analyzed with respect to the variations of heat flux, mass flux and pressure. Eleven heat transfer correlations were assessed against the current set of test data. The results showed that the correlations of Jackson and Krasnoshchekov–Protopopov give the best predictions on average Nusselt number.

Flow Distribution and Pressure Loss in Subchannels of a Wire-Wrapped 37-pin Rod Bundle for a Sodium-Cooled Fast Reactor

A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20e115% of the nominal flow rate and 60℃ (equivalent to Re ~ 37,100) at the inlet of the test rig. The pressure loss data in three measured subchannels were almost identical regardless of the subchannel locations. The flow rate at each type of subchannel was identified and the flow split factors were evaluated from the measured data. The predicted correlations and the computational fluid dynamics results agreed reasonably with the experimental data.

A momentum source model for wire-wrapped rod bundles—Concept, validation, and application

Large uncertainties still exist in the treatment of wire-spacers and drag models for momentum transfer in current lumped parameter models. To improve the hydraulic modeling of wire-wrap spacers in a rod bundle, a three-dimensional momentum source model (MSM) has been developed to model the anisotropic flow without the need to resolve the geometric details of the wire-wraps. The MSM is examined for 7-pin and 37-pin bundles steady-state simulations using the commercial CFD code STAR-CCM+. The calculated steady-state inter-subchannel cross flow velocities match very well in comparisons between bare bundles with the MSM applied and the wire-wrapped bundles with explicit geometry. The validity of the model is further verified by mesh and parameter sensitivity studies. Furthermore, the MSM is applied to a 61-pin EBR-II experimental subassembly for both steady state and PLOF transient simulations. Reasonably accurate predictions of temperature, pressure, and fluid flow velocities have been achieved using the MSM for both steady-state and transient conditions. Significant computing resources are saved with the MSM since it can be used on a much coarser computational mesh.

Static pressure and wall shear stress distributions in air flow in a seven wire-wrapped rod bundle

An experimental investigation is performed in a turbulent flow in a seven wire-wrapped rod bundle, mounted in an open air facility. Static pressure distributions are measured on central and peripheral rods. By using a Preston tube, the wall shear stress profiles are experimentally obtained along the perimeter of the rods. The geometric parameters of the test section are P/D=1.20 and H/D=15. The measuring section is located at L/D=40 from the air inlet. It is observed that the dimensionless static pressure and wall shear stress profiles are nearly independent of the Reynolds number and strongly dependent of the wire-spacer position, with abrupt variations of the parameters in the neighborhood of the wires.

Model for subchannel friction factors and flow redistribution in wire-wrapped rod bundles

The fuel element of LMFBR consists of a bundle of rods wrapped with an helical wire as spacer, surrounded by an hexagonal duct. In the present work, a semi-empirical model is developed to calculate bundle average and subchannel based friction factors and flow redistribution. The obtained results were compared to experimental data and they were considered satisfactory for wide range of geometrical parameters.

Distributed resistance modeling of wire-wrapped rod bundles

Three-dimensional flow hydrodynamic distributed resistance models for rod bundles were developed. The models specifically account for the presence of the wire-wrap spacer and may be used for any lumped parameter thermohydraulic analysis numerical program. Validation studies of the hydrodynamic resistance models were also performed using a subchannel code ASFRE. The models were tested against subchannel velocity and temperature data taken from bundles of triangular rod array configurations with wire-spacers. Overall the models performed satisfactorily predicting the most important qualitative trends for flows in wire-wrapped rod bundles.