Wire wrapped fuel pins, typical of sodium fast reactor designs, introduce a strong secondary rotational flow and enhanced mixing compared to bare pins found in many other nuclear reactor designs. The transverse flow is created by deflection off of the helical wires, exerting a force onto the wires. The resulting pressure field can be more intricate than found for bare fuel pins. There is an increased concern for fluid structure interactions such as twisting, bending, and vibration due to the fluid deflection and increased transverse flow. The present study uses LES to simulate the fluid flow in a wire-wrapped fuel pin bundle. The simulation is performed using Nek5000, a spectral element LES/DNS code. The time-dependent, fluctuating pressure field at the surface of select fuel pins are specifically captured for analysis. Statistical analysis is performed on the pressure data to find areas of persistent force or bending moment, which over large periods of time may result in deformation of the pins. Further, the pressure data is analyzed in search of areas with large fluctuations in pressure, which may result in flow induced vibrations. Frequency spectra of the force fluctuations are analyzed.

The development and validation of mechanistic boiling heat transfer models has been a focal point in the efforts to improve the efficiency and profitability of power generation systems, e.g. nuclear reactors. The primary goal of these models is improving the accuracy of boiling heat transfer simulations and reducing the uncertainty margins that affect both the design and the safety of a system. However, the emergence of these models has also stimulated the need for high-fidelity experiments and experimental data for validation and verification. In this work we present first-of-a-kind data of heat flux partitioning in boiling heat transfer, obtained using cutting-edge diagnostics and post-processing techniques. A HSV camera was used to visualize the boiling surface at 10,000 frames per second with simultaneous front and side views of the two-phase flow. A high-speed IR camera was used to capture the 2-D radiative signal from the boiling surface to visualize bubble nucleation, growth and detachment at a 115 μm/pixel resolution at 2,500 frames per second. A coupled radiation-conduction calibration model was used to calibrate the IR data and extract the full local temperature and heat flux distributions on the boiling surface, which enable a direct measurement of the partitioned heat fluxes. Here we report the results of investigations performed in flow boiling conditions with a mass flux of 500 kg/m 2 /s, at atmospheric pressure and 10 K of subcooling. These data will be leveraged to inform the development and validation of the next generation of mechanistic boiling heat transfer models, to be used in Computational Fluid Dynamics (CFD) codes for the design and the safety analysis of nuclear reactors.

Thermal striping is of particular significance in nuclear reactor applications, primarily in sodium cooled fast reactors. The mixing chamber of the upper plenum of a nuclear reactor can be subjected to thermal striping unless designed such that the coolant is sufficiently mixed prior to reaching the top wall of the upper plenum. In order to conduct a systematic analysis of this phenomenon a simplified experimental set-up was designed and built at Argonne National Laboratory. In a parallel effort a similar simulation was conducted using the spectral-element code Nek5000. The set-up consists of two turbulent jets entering a rectangular tank via two hexagonal inlets, the interesting phenomena being the mixing within the tank. Two different inlet geometries were studied previously, both experimentally and via high-fidelity large-eddy simulations reporting various turbulent statistical quantities. To further assess the flow behavior we hereby perform a Proper Orthogonal Decomposition (POD) to identify the most dominant energetic modes and quantify their impact on the top wall of the upper plenum. The POD analysis of the experimental data in both inlet geometrical configurations is compared with LES and presented to highlight the impact of geometry on the velocity and thermal fields. We find a qualitative coherence between both simulation and experiment, characterized by a strong backflow in the weakly stable geometry, as indicated by the first mode, and the presence of three stagnation points in the strongly stable geometry setup. Also we identify a pairing of modes 1 and 3 with higher frequency than the second mode. This pairing is opposite in the two flow configurations leading to a faster decay of one of the jets in one case and a stable flow in the other.

Microchannel heat sinks (MCHS) are explored as integrated cooling option in diversified applications ranging from microelectronic devices to nuclear reactors. Parallel channel MCHS suffers from the problem of flow maldistribution in channels. The phenomenon of flow maldistribution causes undesired effects such as non-uniform temperature distribution, increase in base maximum temperature and drop in MCHS performance. In the present single phase numerical study; flow maldistribution in vertical supply MCHS has been estimated by taking entire heat sink as computational domain. Three-dimensional governing equations for both fluid flow and energy transfer are solved by finite volume method. A novel heat sink design by splitting the inlet flow in two equal parts has been proposed to effectively reduce flow maldistribution problem in MCHS. Results of velocity distribution, temperature distribution of fluid and overall thermal performance of proposed MCHS are compared with the conventional MCHS design. Proposed scheme helps in reducing flow maldistribution problem significantly (∼55%) as compared to conventional MCHS.

The study of flows over spherical bluff bodies is relevant in engineering systems such as nuclear power reactors. These flows exhibit complex spatial and temporal behavior due to the presence of significant streamwise curvature that leads to boundary layer separation and reattachment, and the turbulence phenomena associated with these effects. In this study, we present results for flow over one and two spheres in free-stream conditions at Reynolds number 1,000, which corresponds to a turbulent-flow regime. The goal of the study was to verify our simulation methodology and develop an initial understanding regarding the spatial and temporal effects of additional spheres in close proximity to the canonical single-sphere case. Based on our results, we find that additional spheres significantly alter the behavior of the flow. The addition of a second sphere suppresses vortex formation, while the temporal correlation between points in the wake is also reduced.

Bubbly flow is quite common in various natural and engineering phenomena. In particular, nuclear engineers are interested in fundamental understanding of the bubbly flow behavior due to its importance in cooling light water reactor cores. Given the extreme conditions and complex support structures in nuclear reactor cores, it is very challenging to study the flow behavior using high-fidelity experiments. Typically validated computational codes are chosen as practical tools for the thermal-hydraulic and safety analyses. As the new generations of nuclear reactors are being developed, more advanced modeling techniques are required to design safe and efficient systems. Different from most simulation approaches, direct numerical simulation (DNS) employs no turbulence closure assumptions, which makes it a promising tool for model development. The major bottleneck of DNS was and remains to be the high computational cost, increasing exponentially with the Reynolds number. However, thanks to the on-going improvements in computer power, these computationally expensive simulations are becoming more and more affordable. Coupled with level-set interface tracking method (ITM), DNS can be used for the high-fidelity studies of two-phase bubbly flows with unprecedented details. Meanwhile, another concern that arises is how one can best take advantage of the ‘big data’ generated from large-scale DNS and translate it into new knowledge. The traditional level-set method utilizes a signed distance field to distinguish different phases while the interface is modeled by the zero level-set. Although level-set method can distinguish gas bubbles from the liquid phase, it cannot recognize and track individual bubbles which hinders the collection of useful bubble information. As a result, the bubble tracking capability has to be developed to improve the data extraction efficiency. In the present work, a marker field is created and advected for bubble distinction and extraction of detailed bubble parameters from the simulations. Each bubble in the flow gets assigned a unique ID, based on which the code will collect the corresponding bubble information. It has been demonstrated that bubble tracking capability can significantly improve the data extraction efficiency for level-set based two-phase flow simulations. Statistical analysis tools are also developed to post-process the recorded information about the bubbles to study the dependencies/correlations of bubble behavior with bubble local conditions. For example, in the pressurized water reactor (PWR) subchannel geometry investigated in this paper, bubbles are observed to experience different relative velocity when presenting at different distance from fuel rod surfaces. With proper grouping criterion, statistical analysis would allow introducing variable drag coefficient for bubbles based on their positions. These new insights are contributing to more accurate modeling of the multiphase computational fluid dynamic (M-CFD) simulations, and better prediction of two-phase flow behavior in engineering systems. Together with the analysis tools, bubble tracking capability will open a new door to study and understand two-phase flows.

Flow-induced vibration (FIV) is a widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occur. Given the importance of reliable components in the nuclear industry, flow-induced vibrations have long been a major concern in the safety and operation of nuclear reactors. In particular, nuclear fuel and steam generators have been known to suffer from flow-induced vibrations and related failures. Over the past five years, the Nuclear Energy Advanced Modeling and Simulation program has developed the integrated multiphysics code suite SHARP. The goal of developing such a tool is to perform multiphysics modeling of the components inside a reactor core, the full reactor core or portions of it, and be able to achieve that with various levels of fidelity. This flexibility allows users to select the appropriate level of fidelity for their computational resources and design constraints. In particular SHARP contains high-fidelity single-physics codes for structural mechanics and fluid mechanics calculations: the structural mechanics implicit code Diablo and the computational fluid dynamics spectral element code Nek5000. Both codes are state-of-the-art. highly scalable (up to millions of processors in the case of Nek5000) tools that have been extensively validated. These tools form a strong basis on which to build an FIV modeling capability. This work discusses in detail the implementation of a fluid-structure interaction methodology in SHARP for simulating flow-induced viration based on the coupling between Diablo and Nek5000. Initial verification and validation efforts are also discussed, with a focus on standard benchmark cases: the flow past a cylinder, the Turek benchmark, and the flow in a Coriolis flow meter.

Interface tracking simulation (ITS) is one of the promising approaches to describe heat transfer of boiling phenomena and their underlying mechanisms. Better understanding and modeling of this process will benefit various engineering systems. In modern nuclear reactors, study on nucleate boiling phenomena is very important for the prediction of the Critical Heat Flux (CHF) phenomena. The presented research will implement and verify the capability of evaporation process modeling by the massively parallel research code, PHASTA. The comparison of the numerical results and the analytical results demonstrates that the overall behavior of the simulation compares well with the analytical solution. A second simulation of the single bubble growth with non-uniform temperature distribution demonstrates both condensation and evaporation modeling. In the third simulation flow boiling capabilities were preliminary tested with laminar flow demonstration case. These results will be applied to a larger scale, multi-bubble simulations and help modeling of nucleate boiling phenomena.

The convective heat transfer of rod bundle flow with spacer grid was investigated preliminarily for nuclear reactor core application. As the test fluid, the water was used. To simulate the nuclear fuel assembly, 4×4 rod bundle with P/D (=pitch between rods/rod diameter) of ∼1.35 was prepared together with a spacer grid with twist-mixing vane. A single heated section with five thermocouples embedded in the surface along the circumferential direction was installed around the center subchannel. The measurements of wall temperatures were carried out upstream and downstream of spacer grid. For the rod bundle flow at the inlet of spacer grid (i.e., upstream of spacer grid), the wall temperatures at the gap and subchannel centers exhibited the higher and lower, respectively, which was because in the subchannel center, the axial flow velocity became higher, as compared with the gap center. On the other hand, downstream of spacer grid, the rod wall toward the tip of twist-mixing vane showed the lowest temperature in the measurements along the circumferential direction of rod wall. Near the twist-mixing vane, the averaged wall temperature was observed to be remarkably low, which implies that the twist-mixing vane is an effective tool to enhance the convective heat transfer performance. However, along the axial flow direction behind the spacer grid, the averaged wall temperatures became to increase, and the enhancement of convective heat transfer performance by mixing vane faded away.

The flow characteristic of fluid at low Prandtl number is of continued interest in the nuclear industry because liquid metals are to be used in the next-generation nuclear power reactors. In this work we performed direct numerical simulation (DNS) for turbulent channel flow with fluid of low Prandtl number. The Prandtl number was set to 0.025, which is representative of the behavior of liquid metals. Constant heat flux was imposed on the walls to study heat transfer behavior, with different boundary conditions for temperature fluctuation. The bulk Reynolds number was set as high as 50,000, with a corresponding friction Reynolds number of 1,200, which is closer to the situation in a reactor or a heat exchanger than used in normally available databases. Budgets for turbulent variables were computed and compared with predictions from several RANS turbulence models. In particular, the Algebraic Heat Flux Model (AHFM) has been the focus of this comparison with DNS data. The comparisons highlight some shortcomings of AHFM along with potential improvements.

The axial flow in rod bundles has been the object of several investigations, since they are relevant for various industrial applications (e.g., heat exchangers, nuclear reactor cores). For tight configurations (pitch-to-diameter ratio smaller than 1.1) the large difference in velocity within the cross section creates the possibility of a Kelvin-Helmholtz instability and the generation of a vortex street in the gap (“gap instability”). While the presence of spacing devices may perturb the flow, simple grid spacers — grids of the type usually encountered in sodium fast reactors — do not prevent this instability. This work seeks to improve our understanding of the gap instability in large rod bundles by using high-fidelity large-eddy simulation (LES). The application of LES methods has been historically limited to single-pin calculations because of the large cost of full-bundle calculations. However, using Nek5000, a massively parallel spectral-element computational fluid dynamics code, we are able to carry out well-resolved LES simulations for this class of geometries. Results are compared with experimental data. For the largest bundle case, over 8 billion collocation points at a polynomial order of 20 are judged necessary to achieve overall excellent accuracy. More than 500,000 processors are used to carry out the simulations (costing approximately 50,000,000 CPU-hours), and 3,000 snapshots of the flow field have been collected to apply several coherent structure recognition techniques, including one of the largest proper orthogonal decompositions carried out to date. Also investigated here is the effect of geometry, including the (presence of spacers, changing pitch-to-diameter ratio, and canister wall. In particular, global linear stability analysis is applied to a series of simplified geometries in order to gain insight into the physics of the gap instability.

Numerical simulation is performed for a quenching process in liquid jet impingement, which is applicable to thermal control in metal manufacturing and emergency cooling of nuclear reactors. The flow and cooling characteristics of jet impingement are investigated by solving the conservation equations of mass, momentum and energy in the liquid and gas phases. The liquid-vapor and liquid-air interfaces are tracked by a sharp-interface level-set method which is modified to include the effect of phase change at the liquid-vapor interface. The temporal and spatial variation of solid temperature is analyzed by solving a conjugate problem with the conduction in the solid as well as the convection in the liquid and gas phases. The numerical results demonstrate that the temporal variations of the temperature and heat flux near the fluid-solid interface are very steep compared to those inside the solid. The heat flux variations at the fluid-solid interface are observed to be much larger in the convection mode than in the film boiling mode. The solid temperatures and heat fluxes obtained from the present study are compared with the experimental data reported in the literature.

Sodium-cooled Fast Reactor (SFR) is one of the generation IV (Gen-IV) nuclear reactors. Prototype Gen-IV SFR (PGSFR) is a SFR being developed in Korea Atomic Energy Research Institute (KAERI). Decay Heat Removal System (DHRS) in the PGSFR has a safety function to make shutdown the reactor under abnormal plant conditions. Single DHRS loop consists of sodium-to-sodium decay heat exchanger (DHX), helical-tube sodium-to-air heat exchanger (AHX) or finned-tube sodium-to-air heat exchanger (FHX), loop piping, and expansion vessel. The DHXs are located in the cold pool and the AHXs and FHXs are installed in the upper region of the reactor building. The DHRS loop is a closed loop and liquid sodium coolant circulates inside the loop by natural circulation head for passive system and by forced circulation head for active system. There are three independent heat transport paths in the DHRS, i.e., the DHX shell-side sodium flow path, the DHRS sodium loop path through the piping, the AHX shell-side air flow path. To design the components of the DHRS and to determine its configuration, key design parameters such as mass flow rates in each path, inlet/outlet temperatures of primary and secondary flow sides of each heat exchanger should be determined reflecting on the coupled heat transfer mechanism over the heat transfer paths. The number of design parameters is larger than that of the governing equations and optimization approach is required for compact design of the DHRS. Therefore, a genetic algorithm has been implemented to decide the optimal design point. The one-dimensional system design code which can predict heat transfer rates and pressure losses through the heat exchangers and piping calculates the objective function and the genetic algorithm code searches a global optimal point. In this paper, we present a design methodology of the DHRS, for which we have developed a system code coupling a one-dimensional system code with a genetic algorithm code. As a design result, the DHRS layouts and the sizing of the heat exchangers have been shown.

Direct Numerical Simulation (DNS) of fully developed velocity and passive scalar temperature fields in two-dimensional turbulent channel flow was coupled with the unsteady conduction in the idealized slab heated with constant volumetric heat source. Similar geometry can be found in some experimental nuclear reactors with fuel in the form of parallel slabs. Beside streamwise and spanwise directions, periodicity of the computational domain was assumed also in the wall-normal direction. Simulations were performed at constant friction Reynolds number 180 and Prandtl number 1, and with various geometrical and material properties of the heated slab. Due to the periodicity, the same Reynolds number and the same flow direction is assumed on both sides of the slab. Results of the simulations predict penetration of the turbulent temperature fluctuations into the solid wall. For thick slab, temperature fluctuations from both sides of the slab do not interfere. As the slab gets thinner, fluctuations from both sides interfere and tend to a finite value as the slab thickness limits toward zero. However, due to the non-coherent turbulent flows on each side of the slab, thermal fluctuations of the zero-thickness slab are actually lower than in the case of the zero-thickness wall heated by the same turbulent flow on one side but cooled by the constant heat flux boundary condition on the other side. Results of the present study can serve as benchmarks for less accurate mathematical models used to predict temperature fluctuations and thermal fatigue in realistic conditions.

The flow in rod bundles is of key importance in a variety of engineering fields. It is representative, for instance, of the flow in a nuclear reactor core and the flow in a tube and shell heat exchanger. For particularly tight rod bundles (pitch to diameter ratio lower than 1.1 in hexagonal arrays) the flow is subject to a Kelvin-Helmholtz instability that is present in both laminar and turbulent flows and is likely to influence transition. Such instability has been proven to exist in bare bundles and even in bundles containing spacing devices. In fact, rod bundles are typically separated by devices designed to keep the rods (or tubes) in place and avoid vibrations. Simple spacer designs, such as honeycomb arrays with dimples, do not remove the instability. The present paper discusses a procedure to investigate the effect of spacing devices have on the most unstable mode of turbulence. A linear stability analysis of the laminar flow for several spacing device configurations has been performed by using adjoint methods. The results show the effect of the spacing on the formation of large scale coherent structures and help lay the ground for a theory of turbulence control in tight rod bundles.

Gas injection into a liquid cross flow from a nozzle causes bubble formations which have potential applications in industry such as chemical plants, waste water treatment and bio- and nuclear-reactors. The purpose of this study is to experimentally investigate the effects of nozzle shape and configuration with respect to the liquid cross-flow direction, on the bubbly flow characteristics such as bubble formation, detached bubble size and frequency at different gas and liquid flow rates. The experiments were conducted in a Plexiglas two-dimensional rig using a high speed camera. High speed imaging and an image processing algorithm were used to track each individual bubble and to quantify the bubble growth as well as the detachment frequency and the bubble velocity. Back light shadowgraphy which utilizes a low intensity diffuse light source to illuminate the background was used to image bubbles. Nozzles were mounted in the test section which was designed such that the flow in this section has a two-dimensional profile. The results showed that the bubble size increases with an increase in GLR (gas to liquid flow rates ratio). Furthermore, the bubble formations and detached bubble size were strongly influenced by the nozzle shape and configuration.

Turbulent air flow through subchannels is simulated to obtain characteristic frequency on large scale flow pulsation in square arrayed 6-rod bundles. The turbulent mixing between the subchannels is dominated by the large scale flow pulsation phenomenon. Therefore, it is crucial to understand flow patterns in the subchannels for the design of reliable nuclear reactor elements. Numerical simulation using a CFD code is performed by solving the unsteady equations with large eddy simulation (LES) model. For the numerical simulation, two kinds of grid including two subchannels are created. Models with two different P/D ratios (P/D = 1.08, 1.35) are simulated. P/D = 1.08 case is our main concern to investigate flows in tight lattice rod buddle. The shortened channel length with periodic boundary condition is employed to reduce the computing time. The geometry of the simulated model is based on the 3-D experimental equipment. The velocity fluctuation data in the stream-wise and cross directions are obtained from the simulated model to find the characteristic frequency. The large scale flow pulsations are explained by time history of velocity, Fast Fourier Transform analysis and cross-correlation analysis. The fluctuation frequency at the center of gap and 20 mm away from that point are obtained. The characteristic frequency is 87.5 Hz through the entire monitoring points. Results demonstrate that the periodic flow pulsations occur with small pitch-to-diameter ratio (P/D = 1.08) and the vortices trains are highly synchronized. There is no large scale pulsation in P/D = 1.35 case.

Parametric study on an outlet plenum of a PBMR type gas cooled nuclear reactor has been carried out using three-dimensional Reynolds-Averaged Navier-Stokes equations. Shear stress transport turbulence model is used for analysis of turbulence. Two geometric parameters are selected, namely, displacement on the horizontal line and angle of rotation about the center of gravity of the roof support block. Pressure drop and uniformity of flow distribution in the outlet plenum have been considered as the performance parameters of an outlet plenum of the PBMR. The results show that the performance parameters are affected considerably by the location of the roof support block.

Echo-PIV system was applied to measure the velocity fields of a pump system inside the nuclear reactor model. The scale-downed model of pump system driven by a motor was built for this study. 2D instantaneous velocity vector fields of the downstream region of impeller were obtained. The flow rate was calculated and compared with the result from the flow meter. From this study, the feasibility of echo-PIV for studying the opaque flow in the fluid machinery was shown.

Accurate analyses of turbulence structure and void fraction distribution are quite important in designing and safety evaluation of various industrial equipments using gas-liquid two-phase flow such as nuclear reactor, etc. Using turbulence model of two-phase flow and models of bubble behaviors in bubble flow and slug flow, systematic analyses of distributions of void fraction, averaged velocity and turbulent velocity were carried out and compared with experimental data. In bubbly flow, diffusion of bubble and lift force are dominant in determining void fraction distribution. On the other hand, in slug flow, large scale turbulence eddies which convey bubbles into the center of flow passage are important in determining void fraction distribution. In turbulence model, one equation turbulence model is used with turbulence generation and turbulence dissipation due to bubbles. Mixing length due to bubble is also modeled. Using these bubble behavior models and turbulence models, systematic predictions were carried out for void distributions and turbulence distributions for wide range of flow conditions of two phase flow including bubbly and slug flow. The results of predictions were compared with experimental data in round straight tube with successful agreement. In particular, concave void distributions in bubbly flow and convex distribution in slug flow were well predicted based on the present model.