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Delchini, Marc-Olivier G.

Publications and source records attributed to Delchini, Marc-Olivier G..

10 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

A Large Eddy Simulation Study of Flow Turbulence, Alumina Transport, and Bath Temperature Evolution in Conventional Aluminum-Smelting Cell Using OpenFOAM

In this study, a Large Eddy Simulation (LES) of the aluminum-smelting process is performed using OpenFOAM. To understand the coupled behavior of heat transfer, mass transfer, and flow of the smelting process, a multi-physics computational fluid dynamics (CFD) model based on the Eulerian–Eulerian multi-fluid approach is adopted. The model accounts for CO 2 bubble and magnetohydrodynamics (MHD)-driven flow, along with alumina dissolution, transport, and bath temperature evolution. The simulation predictions show small-scale turbulent vortical structures in the anode–cathode space caused by combined effect of MHD and CO 2 bubble-bath interactions and relatively large-scale asymmetric vortices in the inter-anode space caused by the CO 2 bubble-bath interactions. The vortex formation at the edges of the anodes evidently aids in transporting alumina from the central channel to the bottom of the anodes and prevents accumulation of gas bubbles in the periphery of the anode bottom. Symmetric bath cold spots are observed in the vicinity of the feeder. Cold spots are also observed in the anode–cathode distance space below the anode bottom due to the transport of undissolved solid to this region by the flow. The findings from the work are useful in developing and designing alumina-feeding strategy leading to reduced anode effects and smooth operation of the cell. The work also highlights the important flow structures in conventional aluminum-smelting cell.

36 MATERIALS SCIENCE↗
DOE OSTI2019

High Performance Computing to Enable Next-Generation Low-Temperature Waste Heat Recovery

US manufacturing industry fails to recover an estimated 900 trillion BTUs of low-temperature waste heat from its processes each year. A grand research challenge has been to develop waste heat recovery technology that can be applied to industrial manufacturing processes and vehicle operations. The Oak Ridge National Laboratory (ORNL)/Eaton research team performed computational research and development to design an innovative direct-contact heat exchanger (DCHE) technology to deliver low-cost, compact, longer-lifetime, high-efficiency waste heat recovery that is optimized for a low-temperature organic Rankine cycle. ORNL resources and expertise in high performance computing (HPC) and multiphase flows were utilized to realize this goal while advancing the fundamental understanding of two-phase, two-immiscible-fluid turbulent flow heat transfer for heat exchangers. In the first stage of the project, a computational fluid dynamics (CFD) model of a DCHE was developed. Verification and limited validation of the numerical solution were performed using experimental data from published literature. To support efforts in verification and validation, two-dimensional CFD models were also developed with the objective of investigating different boundary conditions to reach converged, steady-state solutions.Based on benchmarking efforts, a baseline industry-grade design of a DCHE was developed, which consists of two horizontal pipes of different cross-sectional areas – joined by a converging-diverging nozzle in the middle and relies on a three-phase system with co-current flows for operation. Evaporation of the cold liquid (n-pentane or methanol) by the hot liquid water occurs in the first pipe. The first pipe contains two inlets for liquid water and liquid pentane (or liquid methanol). The second pipe with the larger cross-sectional area serves as a gravity-driven phase-separator. It contains two outlets that are staggered to avoid entrainment of the liquid water phase by the gas phase. This baseline design resulted from several trial-and-error evaluations using two-dimensional CFD models. Confirmatory three-dimensional CFD models were also run for different mass flow rates and inlet temperatures (inlet water temperature ranging from 50°C to 90°C), and efficiency plots were produced to fully characterize the proposed industrial-grade design of the DCHE.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2019

Integration of the Nek5000 Computational Fluid Dynamics Code with the NEAMS Workbench

The mission of the US Department of Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program is to develop, apply, deploy, and support state-of-the-art predictive modeling and simulation tools for the design and analysis of current and future nuclear energy systems using computing architectures ranging from laptops to leadership-class facilities. NEAMS is organized into three product lines: the Fuels Product Line (FPL),the Reactors Product Line (RPL), and the Integration Product Line (IPL). The FPL and RPL focus on the development of high-fidelity tools such as the MOOSE-based applications BISON and MARMOT [3, 9] at Idaho National Laboratory, and the PROTEUS neutronics code and the NEK5000 computational fluid dynamics code from the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) framework [10]. The NEAMS IPL will enable users to apply high-fidelity simulations to inform lower order models for the design, analysis, and licensing of advanced nuclear systems.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

High Performance Computational Models of Synthetic Jet Actuators for Step Change Improvements in Freight Efficiency

The project presents results of a computational fluid dynamics (CFD) model developed by the Oak Ridge National Laboratory team for a synthetic jet produced by an actuator. Numerical results were compared against experimental data provided by Actasys, Inc. Three CFD packages were used for this work that are Star-CCM+, Nek5000 and OpenFOAM. A sensitivity analysis was performed with Star-CCM+ and provided more insights on how to optimize the design of the actuator. Nek5000 was used to obtain a high-resolution solution of the synthetic jet. In a exploratory stage, a CFD model of the actuator was also developed in OpenFOAM. All CFD models were able to capture the main flow features.

42 ENGINEERING↗
DOE OSTI2018

An Uncertainty Quantification of the Computational Fluid Dynamics Solution to the Modeling of the Contact Point in a Wire-Wrapped Fuel Assembly

An uncertainty quantification and sensitivity analysis of the turbulent flow in a wire-wrapped pin bundle to the contact-point model and the inlet conditions is presented. Three geometries are considered in this work that are an infinitely large bundle (single subchannel with periodic boundary conditions in the stream-wisedirection), a 7-pin bundle and a 19-pin bundle. The response functions of this study are the friction factor, thevorticity and the mixing parameters. The results show that the friction factor is not sensitive to the contact-point model. The vorticity shows a strong dependency to the inclusion of prismatic layers near the wall boundaries. Results for the mixing parameters are inconclusive. A variance-based sensitivity analysis(Sobol indices) indicates that the influence of the contact-point model on the variance of the friction factor decreases when increasing the size of the bundle. Other response functions (vorticity and mixing parameters)show little sensitivity to the size of the bundle.

42 ENGINEERING↗
DOE OSTI2017

Report on UQ Assessments to support SESAME wire-wrapped bundle experiment

This work assesses the influence of assumptions made when generating a mesh of a wire-wrapped geometry. The contact region between a wire and its adjacent pin is commonly modeled by either embedding the wire to the adjacent pin or trimming the wire so that a gap separates the wire from its adjacent pin. These models are referred to as close-gap and open-gap approaches herein and are applied to two geometries. The first geometry consists of a single pin wire-wrapped subchannel. A polyhedral mesh and a hexahedral mesh are generated. The second and third geometry are a 7-pin and a 19-pinwire-wrapped bundles meshed with polyhedral elements only. Pressure drops are obtained with the STAR-CCM+computational fluid dynamic package. Sensitivity analyses of the mesh density, the mesh type, and the turbulent models are performed. Numerical results show that the best match to the experimental data and to the Cheng-Todreas correlation is obtained with the combination of a hexahedral mesh, the shear stress transport (SST) turbulent model, and the open-gap approach. In the case of the 7-pin geometry, the best results are obtained with the open-gap approach and the SST turbulent model. The 19-pin geometry yields contradictory results to the 7-pin geometry results, and thus will require further investigations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Implementation of a Pressure Correction Term in the CTF Subchannel Code

CTF is a modernized and improved version of the legacy subchannel code COBRA-TF, being jointly developed and maintained by Oak Ridge National Laboratory (ORNL) and North Carolina State University. The code was adopted by ORNL for use in the Consortium for Advanced Simulation of Light Water Reactors (CASL) in 2012 to help address CASL challenge problems. Since that time, activities related to CTF have included implementing software quality assurance measures, implementing new features and models, performing validation and verification testing, establishing and supporting a CTF user group, and developing the code for use in coupled applications. CTF is applicable to single- and two-phase flows in light water reactor geometries at normal and accident operating conditions. The two-phase flow model consists of the six-equation two-phase flow model augmented by equations for the droplet field, and it also includes appropriate source terms. This work focuses on the numerical stability of the six-equation two-phase flow model and is part of a recent effort to improve CTF capabilities for simulating flows in boiling water reactors (BWRs). The six-equation two-phase flow model is not strongly hyperbolic, meaning that some of its eigenvalues are either not real or not distinct. In a Cauchy sense, and for an initial-value problem, strong hyperbolicity was found to be equivalent to well-posedness, which ensures the uniqueness of the solution in the space-time used. This concept is also of importance when designing high-order numerical methods and also when implementing boundary conditions that rely on real eigenvalues such as Riemann solvers. A tremendous amount of research is available in the literature on how to hyperbolize the six-equation two-phase flow model by adding ad hoc terms to the system of equations. Such terms are often referred to as virtual mass or interfacial pressure correction terms. From a mathematical perspective, this approach adds terms to the phasic momentum equation containing first-order partial derivatives of the unknowns to modify the Jacobian matrix, thus the imaginary part of the eigenvalues. For this work, an interfacial pressure correction (IPC) term on the same model as what was used for the CATHARE system code, is implemented in CTF. This allows for recovery of a two-pressure formulation. In the following, the six-equation two-phase flow model is recalled, along with the IPC term and its influence on the eigenvalues. Then, numerical results and convergence studies are presented for a water-faucet problem at a pressure of 10 bar. Finally, conclusions and future work are addressed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Assessment of SFR Wire Wrap Simulation Uncertainties

Predictive modeling and simulation of nuclear reactor performance and fuel are challenging due to the large number of coupled physical phenomena that must be addressed. Models that will be used for design or operational decisions must be analyzed for uncertainty to ascertain impacts to safety or performance. Rigorous, structured uncertainty analyses are performed by characterizing the model’s input uncertainties and then propagating the uncertainties through the model to estimate output uncertainty. This project is part of the ongoing effort to assess modeling uncertainty in Nek5000 simulations of flow configurations relevant to the advanced reactor applications of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. Three geometries are under investigation in these preliminary assessments: a 3-D pipe, a 3-D 7-pin bundle, and a single pin from the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility. Initial efforts have focused on gaining an understanding of Nek5000 modeling options and integrating Nek5000 with Dakota. These tasks are being accomplished by demonstrating the use of Dakota to assess parametric uncertainties in a simple pipe flow problem. This problem is used to optimize performance of the uncertainty quantification strategy and to estimate computational requirements for assessments of complex geometries. A sensitivity analysis to three turbulent models was conducted for a turbulent flow in a single wire wrapped pin (THOR) geometry. Section 2 briefly describes the software tools used in this study and provides appropriate references. Section 3 presents the coupling interface between Dakota and a computational fluid dynamic (CFD) code (Nek5000 or STARCCM+), with details on the workflow, the scripts used for setting up the run, and the scripts used for post-processing the output files. In Section 4, the meshing methods used to generate the THORS and 7-pin bundle meshes are explained. Sections 5, 6 and 7 present numerical results for the 3-D pipe, the single pin THORS mesh, and the 7-pin bundle mesh, respectively.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Dakota Uncertainty Quantification Methods Applied to the CFD code Nek5000

Predictive modeling and simulation of the performance of a nuclear reactor and its fuel is a challenging task because of the large number of coupled physical phenomena that must be addressed. In addition to this intrinsic complexity, model uncertainty must be accounted for in any analysis if the model will be used to facilitate design or operational decisions which may impact safety and performance. Rigorous, structured uncertainty analyses are performed by first characterizing the model's input uncertainties and, by propagating the uncertainty through the model in order to estimate the output's uncertainty. This paper presents the state of advancement of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) project to characterize the uncertainty of the CFD code Nek5000 using the Dakota package. Nek5000 is a high order spectral element computational fluid dynamics (CFD) code developed at Argonne National Laboratory for high resolution spectral-filtered large eddy simulations (LESs) and unsteady Reynolds averaged Navier-Stokes (RANS) simulations. The Dakota package developed at Sandia National Laboratory can be integrated with many scientific and engineering codes to facilitate efficient, effective uncertainty quantification (UQ) and sensitivity analyses. Dakota is designed to interface with external packages/codes such as Nek5000 through a script-driven interface called fork. Such interface provides maximum flexibility to the users for pre- and post-processing tasks preceding and following each function evaluation. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗