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As part of a joint U.S. Department of Energy project, Framatome, Argonne National Laboratory, Terrapower, and Texas A&M University have collaborated to produce experimental data and computational results to characterize the flow behavior of a 61-pin wire-wrapped hexagonal fuel bundle. In this paper, comparisons are made between Texas A&M’s experimental particle image velocimetry velocity field measurements and Argonne National Laboratory’s large-eddy simulation results in an interior subchannel for a Reynolds number of 19,000. It can be concluded that the shape and magnitude of the mean vertical velocity component is in excellent agreement for the selected horizontal line plots. Maximum relative errors are less than 10% until the subchannel walls are approached. The shape and, to a lesser extent, magnitude of the mean horizontal velocity component is in satisfactory agreement between the particle image velocimetry and large-eddy simulation results.

Complex flow structure interactions and heat transfer processes take place in nuclear reactor cores. Given the extreme pressure/temperature and radioactive conditions inside the core, numerical simulations offer an attractive and sometimes more feasible approach to study the related flow and heat transfer phenomena in addition to the experiments. Under the Exascale Computing Project, the full-core simulation of a small modular reactor (SMR) has been pursued coupling Computational Fluid Dynamics (CFD) and neutronics. A key aspect of the modeling of SMR fuel assemblies is the presence of spacer grids and the mixing promoted by mixing vanes or the equivalent. A reduced order methodology is adopted based on momentum sources to mimic the mixing of the vanes. The momentum sources have been carefully calibrated with detailed Large Eddy Simulations (LES) of spacer grids performed with Nek5000. Modeling the spacer grid and mixing vanes (SGMV) effect without body-fitted computational grid avoids the excessive costs in resolving the local geometric details, and thus supports the simulation to be scaled up to the full core. Besides the progress on momentum source modeling, this paper also features the first full-core pin resolved CFD simulation ever performed to the authors’ knowledge. This represents a significant advancement in capability for the CFD of nuclear reactors, which will hopefully serve as an inspiration for further integrating high-fidelity numerical simulations in actual engineering designs.

Numerical simulation has been an intrinsic part of nuclear engineering research since its inception. In recent years a transition is occurring toward predictive, first-principle-based tools such as computational fluid dynamics. Even with the advent of petascale computing, however, such tools still have significant limitations. In the present work some of these issues, and in particular the presence of massive multiscale separation, are discussed, as well as some of the research conducted to mitigate them. Petascale simulations at high fidelity (large eddy simulation/direct numerical simulation) were conducted with the massively parallel spectral element code Nek5000 on a series of representative problems. These simulations shed light on the requirements of several types of simulation: (1) axial flow around fuel rods, with particular attention to wall effects; (2) natural convection in the primary vessel; and (3) flow in a rod bundle in the presence of spacing devices. The focus of the work presented here is on the lessons learned and the requirements to perform these simulations at exascale. Additional physical insight gained from these simulations is also emphasized.

As part of a U.S. Department of Energy International Nuclear Energy Research Initiative (I-NERI), Argonne National Laboratory (Argonne) is collaborating with the Dutch Nuclear Research and consultancy Group (NRG), the Belgian Nuclear Research Centre (SCK·CEN), and Ghent University (UGent) in Belgium to perform and compare a series of fuel-pin-bundle calculations representative of a fast reactor core. A wire-wrapped fuel bundle is a complex configuration for which little data is available for verification and validation of new simulation tools. UGent and NRG performed their simulations with commercially available computational fluid dynamics (CFD) codes. The high-fidelity Argonne large-eddy simulations were performed with Nek5000, used for CFD in the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) suite. SHARP is a versatile tool that is being developed to model the core of a wide variety of reactor types under various scenarios. It is intended both to serve as a surrogate for physical experiments and to provide insight into experimental results. Comparison of the results obtained by the different participants with the reference Nek5000 results shows good agreement, especially for the cross-flow data. The comparison also helps highlight issues with current modeling approaches. The results of the study will be valuable in the design and licensing process of MYRRHA, a flexible fast research reactor under design at SCK·CEN that features wire-wrapped fuel bundles cooled by lead-bismuth eutectic.

The authors develop basic components of a parametric model-order reduction (pMOR) procedure targeting high Rayleigh-number buoyancy-driven flows. The pMOR is based on Galerkin formulation of the governing Boussinesq equations using eigenfunction bases derived from proper orthogonal decomposition. The advantages of pMOR over parametric interpolation are demonstrated for quantities of interest (QOIs) with linear and nonlinear dependencies on the solution in low Rayleigh-number cases. Constraint-base and Leray-type regularizations are shown to offer significant advantages over the standard reduced-order Galerkin formulation in a variety of examples including 3D Rayleigh–Benard flow at Rayleigh number 10 7 and turbulent flow in a half-pipe at Reynolds number 5300.

Numerical simulation is an intrinsic part of nuclear engineering research supporting the design of nuclear power plants. With the advent of Petascale computing (i.e., computers capable of more than 1 PFlop), the simulation of portions of reactor components with turbulence-resolving techniques is now possible. These simulations can provide invaluable insight into the flow dynamics, which is difficult or often impossible to obtain with experiments alone. The spectral element method in particular has emerged as a powerful method to deliver massively parallel calculations at high fidelity by using Large Eddy Simulation or Direct Numerical Simulation. In this work we review the fundamentals of the method and the reasons it is compelling for the simulation of nuclear reactor core flows. We review a series of wall-resolved large eddy simulations of reactor cores applied to three reactor types. In particular we discuss the simulation of the flow in a 17 × 17 rod bundle related to light water reactor (LWR) applications; the flow in a 37-pin wire-wrapped rod bundle related to liquid metal reactor (LMR) applications and the flow through a graphite block with application to molten salt reactors (MSR).

Large eddy simulation (LES) of an experimental benchmark investigating buoyant mixing among two fluids is developed for verification and validation. Flow physics involve a completely closed system where after a valve is opened, two initially stationary fluids with different densities are allowed to mix. The Boussinesq approximation and low-Mach equations are coded into spectral element code Nek5000 to predict the buoyancy forces, and miscibility is accounted for using the advection-diffusion equation for concentration. Due to computational costs of higher order solutions, current analysis consists of initial flow development after the valve is opened. Minor differences between the models and Schmidt number are found within the turbulent spectrum and time integrated quantities. Moreover, solutions for higher polynomial orders are compared to quantify error due to spatial discretization. Error estimates using a posteriori spectral derivation of truncation and quadrature error are also calculated. Validation with particle image velocimetry (PIV) reveals good prediction of both average and fluctuating components of the primary flow within the cold leg but less accurate results in the downcomer. It is concluded that instabilities due to resolving more of the interface using the high molecular Schmidt number is essential for accurately predicting this type of flow.